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Sample records for 237np 231pa ionizatsionnykh

  1. Chemical aspects of 237 Np Moessbauer spectroscopy

    The 237Np Moessbauer effect has been especially useful in studies of neptunium chemistry, by virtue of its excellent resolution and straightforward experimental techniques. Neptunium can have valences from +3 to +7, and a broad range of compounds can be prepared that are analogous to those of other actinide elements. Studies on neptunium compounds, for example, have a ready application to the analogous compounds of uranium and plutonium. The emphasis in this paper will be on the application of the 237Np Moessbauer effect to problems in neptunium chemistry

  2. Prospection for natural 231Pa in India

    Protactinium-231 (231Pa) occurs in nature as a member of the decay chain of naturally occuring 235U of the 4n+ 3 radioactive series. The expected protactinium concentration in the Jaduguda ore body (with uranium concentration of 0.03-0.06 %) is around 0.2 parts per billion (ppb) and that in monazite ore (uranium concentration 0.3%) is 0.9 ppb. The process at uranium ore processing plant at Jaduguda was studied. 231Pa content in samples from the process streams of the plant was determined. The gamma ray spectrometry method was chosen and standardised in our laboratory to detect and measure 231Pa in parts per billion levels in these samples. A concentrated source of protactinium could not be found among the assessed streams of Jaduguda uranium plant. The Monazite processing plant at IRE, Aluva was then studied. From the known chemistry of protactinium, the possible distribution of the 231Pa was guessed at. Accordingly, the process streams of IRE process plant were selected to prospect for 231Pa and determine the fractionation of protactinium. For analysis of 231Pa, the thorium bearing samples were chemically treated to remove the thorium daughter products, which interfere in gamma spectrometry. This report describes the planning for prospecting, sample selection, the standardisation of the analysis procedure for determination of 231Pa content, and the analysis results. The 231Pa content in various streams of Indian Rare Earths plant was found in the range 0.2 -6.5 ppb. Some of the streams did not carry any protactinium. The fractionation of 231Pa in the various streams of the plant and the selection of source for recovery of protactinium are discussed in detail. (author)

  3. In-beam spectroscopy of 231Pa

    Information on energy levels and on E2 and M1 matrix elements in 231Pa has been obtained using conversion-electron and gamma-ray spectroscopy following the 232Th(p, 2p)231Pa reaction and Coulomb excitation of the radioactive target 231Pa by 4He and 32S ions. The results are analyzed in the framework of the rotational model, applied to the rotational band built on the 1/2-[530] Nilsson state whose 3/2- member forms the ground state of this nucleus. The deviations of the level energies from the rigidrotor values can be described by Coriolis couplings. The analysis of the Coulomb-excitation process shows that a constant set of rotational parameters Q0, gR, gK, and b can fairly well account for the measured line intensities. (orig.)

  4. The radiological significance of 237Np in the environment

    Measurements have been made of the concentration of 237Np in tissues taken from sheep, which have grazed on sea-washed marshland on the west Cumbrian coast. The concentrations of 239Pu + 240Pu and 241Am were also determined in the same samples, to enable a comparison to be made of relative uptake of these radionuclides to livestock. Consequently, an assessment has been made of the relative radiological significance of these radionuclides, in terms of the dose received by the public from consuming sheep meat products. The concentrations of the same radionuclides were also measured in soil samples taken from the marsh on which the sheep grazed, to estimate the 237Np available to the sheep, via plant uptake and soil ingestion. (Author)

  5. Pilot measurements of 237Np in forest litter from Poland

    Described are results and the procedure for a pilot study on 237Np content in forest litter samples from Poland in relation to their plutonium activity. Neptunium was determined by inductively coupled plasma mass spectrometry (ICP-MS) and Pu by alpha spectrometry. Two samples originated from a location with pure global fallout and two others from a place with about 65% of the plutonium from Chernobyl. Plutonium activities were determined twice: at Krakow and in Monaco. The two results were consistent and 239+240Pu activities ranged from about 1 to about 7 Bq/kg dry weight (dw). The chemical recovery for Np was between 27 and 89%. Results for 237Np activity concentrations were between 0.099 ± 0.005 and 2.21 ± 0.076 mBq/kg dw. Observed activity ratios were lower than expected and could be explained by fractionation of Np against Pu in forest litter. (author)

  6. Determination of 237Np in sediments and sea water

    Compared to other transuranic elements such as Pu and Am present environmental levels of Np are very low. Neptunium nuclides are therefore of little radiological significance in the concentrations found in the environment. In the longer term, however, 237Np will become one of the most important radionuclides remaining in high-level radioactive waste. We have further developed a new method for analysis of neptunium in large samples of soil, sediment and seawater. The samples are measured on alpha detectors and by ICPMS (Inductively coupled plasma mass spectrometry)

  7. Study on adsorption of 237Np on bentonite

    The performance of adsorbing 237Np for bentonite as buffer/backfill material was investigated. The adsorption coefficients of 237Np were determined for three kinds of bentonite under atmosphere and anoxic atmosphere. Further, it was studied that Kd values were affected by pH and CO32-. The results are shown as follows: (1) Distribution coefficients Kd under atmosphere: for mixed-bentonite is 47.3 mL/g, for Mg-bentonite 52.0 mL/g and for Ca-bentonite 42.4 mL/g; and the corresponding Kd under anoxic atmosphere: 89.3 mL/g, 38.8 mL/g and 29.0 mL/g, respectively. (2) When pH was lower than 9.2, Kd of the mixed-bentonite increased with the increase of pH and the maxi-mumt Kd value appeared at pH 9.2. (3) Kd of the mixed-bentonite got lower in high carbonate concentration when the total neptunium concentration was smaller than the solubility of NaNpO2CO3. (author)

  8. 237Np: Oxidation state in vivo and chelation by multidentate catecholat and hydroxypyridinonate ligands

    Chemically, 237Np(V) is as toxic as U(VI), and radiologically, about as toxic as 239Pu. Depending on redox conditions in vivo, 237Np exists as weakly complexing Np(V) (NpO2+) or as Np(IV), which forms complexes as stable as those of Pu(IV). Ten multidentate catecholate (CAM) and hydroxypyridinonate (HOPO) ligands with great affinity for Pu(IV) were compared with CaNa3-DTPA for in vivo chelation of 237Np. Mice were injected intravenously with 237NpO2Cl: those in a kinetic study were killed 1 to 2,880 min; in ligand studies, fed mice were injected intraperitoneally with a ligand 5, 60, or 1,440 min after 237Np(V), mice fasted for 16 h were gastrically intubated with a ligand 3 min after 237Np(V), and all were killed 24 h after ligand administration; tissues and excreta were radioanalyzed. Rapid plasma clearance and urinary excretion of 237Np(V) resemble U(VI); deposition and early retention in skeleton and liver resemble Pu(IV). The x-ray absorption near edge structure spectroscopy (XANES) spectra of femora of 237Np(V)-injected mice, compared with spectra of Np(V) and Np(IV) from reference solids, showed predominantly Np(IV). Significant in vivo 237Np chelation was obtained with all of the HOPO and CAM ligands injected at molar ratio 22; the HOPO ligands reduced 237Np in skeleton, liver, and other soft tissue, on average, to 72, 25, and 25% of control, respectively, while CaNa3-DTPA was ineffective

  9. 231Pa systematics in postglacial volcanic rocks from Iceland

    Turner, Simon; Kokfelt, Thomas; Hoernle, Kaj; Lundstrom, Craig; Hauff, Folkmar

    2016-07-01

    Several recent studies have highlighted the potential of combined 238U-230Th and 235U-231Pa systematics to constrain upwelling rates and the role of recycled mafic lithologies in mantle plume-derived basalts. Accordingly, we present measurements of the 231Pa concentrations from 26 mafic volcanic rocks from Iceland, including off-axis basalts from the Snaefellsnes Peninsula, to complement previously published 238U-230Th-226Ra data. 231Pa concentrations vary from 27 to 624 fg/g and (231Pa/235U) ratios from 1.12 to 2.11 with the exception of one anomalous sample from the Southeast Rift which has a 231Pa deficit with (231Pa/235U) = 0.86. An important new result is that basalts from the Southeast Rift and the Snaefellsnes Peninsula define a trend at relatively low (231Pa/235U) for a given (230Th/238U) ratio. Many of the remaining samples fall in or around the global field for ocean island basalts but those from the Mid-Iceland Belt and the Southwest Rift/Reykjanes Peninsula extend to higher (231Pa/235U) ratios at a given (230Th/238U), similar to mid-ocean ridge basalts. In principle, these lavas could result from melting of peridotite at lower pressures. However, there is no reason to suspect that the Mid-Iceland Belt and the Southwest Rift lavas reflect shallower melting than elsewhere in Iceland. In our preferred model, these lavas reflect melting of garnet peridotite whereas those from the Southeast Rift and the Snaefellsnes Peninsula contain a significant contribution (up to 20%) of melt from garnet pyroxenite. This is consistent with incompatible trace element and radiogenic isotope evidence for recycled oceanic crust in these lavas. There is increasing agreement that the displacement of ocean island basalts to lower (231Pa/235U) ratios at a given (230Th/238U), compared to mid-ocean ridge basalts, reflects the role of recycled mafic lithologies such as garnet pyroxenite as well as higher average pressures of melting. It now seems likely that this interpretation may

  10. Measurements of {sup 237}Np secondary neutron spectra

    Kornilov, N.V.

    1997-03-01

    The activities carried out during the first year of the project are summarized. The main problems for Np spectra measurements arise from high intrinsic gamma-ray activity of the sample and admixture of the oxygen and iron nuclei. The inelastically scattered neutrons and the fission neutrons spectra for {sup 237}Np were measured by time-of-flight spectrometer of the IPPE at incident neutron energies {approx_equal}1.5 MeV, and {approx_equal}0.5 MeV. A solid tritium target and a Li-metallic target were used as neutron sources. The neutron scattering on C sample (C(n,n) standard reaction) was measured to normalize the Np data. The experimental data should be simulated by Monte Carlo method to correct the experimental data for oxygen and iron admixture as well as for multiple scattering of the neutrons in the sample. Therefore the response function of the spectrometer, and the neutron energy distribution from the source were investigated in detail. (author)

  11. Pre-concentration studies of 237Np using sulphonic acid based actinide™ resin

    The absorption studies of the standard 237Np have been performed using actinide™ resin to standardise the selective separation. The supernatant solution was checked for alpha activity of 237Np using scintillation counting technique. It was found that more than 95 % of 237Np was absorbed in the actinide™ resin. The absorbed 237Np from the actinide™ resin was leached out using isopropyl alcohol. The leached out activity of 237Np in the isopropyl alcohol was estimated using same alpha scintillation counting technique and was found to be greater than 95 %. The selective absorption of 237Np resin from the other impurities in actinide™ is helpful in the analytical solution during recycle of fuels as well as in the waste management process. This is very much important because 237Np is one of the long-lived minor actinides produced from the 238U(n, 2n) reaction followed by beta decay. Thus it is useful for the conventional reactor based on natU and fast reactor based on U-Pu fuel. (author)

  12. Neutron-induced fission cross-section of 231 Pa

    Beside the importance of 231 Pa for basic fission studies it is also of interest in the field of future reactor design based on the thorium-uranium fuel cycle. The 232 Th/233 U breeder cycle, where the natural resources of the main fuel thorium are estimated to last for hundred thousands of years, is contemplated to provide 'clean' and almost inexhaustible nuclear energy. Among the first priority isotopes the IAEA had pointed out 231 Pa and 233 Pa. Both are of special interest being intermediate nuclei in the formation of the fissile 233 U from the fertile 232 Th. The latter has been investigated in the recent past in great detail. In particular, 231 Pa carry a similar risk as 239 Pu does in the standard uranium-plutonium cycle due to its comparable half-life and radio-toxicity. Despite the wealth of existing experimental data important discrepancies exist, a scenario, which holds for the existing evaluated data files ENDF/B-VI and JENDL-3.3, too. Presently, the neutron-induced fission cross-section of 231 Pa is under investigation at the VdG neutron source at IRMM for incident neutron energies up to 20 MeV. The obtained cross-sections, representing the 3rd and higher chance fission in 233 Pa(n,f) will serve as precise input for the validation of the reaction cross-section calculations performed on 233 Pa up to 20 MeV and the envisaged extension up to 50 MeV. (authors)

  13. Status of the Neutron Capture Measurement on 237Np with the DANCE Array at LANSCE

    Neptunium-237 is a major constituent of spent nuclear fuel. Estimates place the amount of 237Np bound for the Yucca Mountain high-level waste repository at 40 metric tons. The Department of Energy's Advanced Fuel Cycle Initiative program is evaluating methods for transmuting the actinide waste that will be generated by future operation of commercial nuclear power plants. The critical parameter that defines the transmutation efficiency of actinide isotopes is the neutron fission-to-capture ratio for the particular isotope in a given neutron spectrum. The calculation of transmutation efficiency therefore requires accurate fission and capture cross sections. Current 237Np evaluations available for transmuter system studies show significant discrepancies in both the fission and capture cross sections in the energy regions of interest. Herein we report on 237Np (n,γ) measurements using the recently commissioned DANCE array

  14. Photo fission cross-section of 232Th, 238U and 237Np

    In the present work, photo fission cross-section of 232Th, 238U and 237Np evaluated with the help of fission fragment angular distribution measurements by using Bremsstrahlung radiation from 7.4 MeV to 9.0 MeV have been carried out by employing high efficiency SSNTD technique

  15. A study of the multipolar composition of the electrofission cross section of 237Np

    The electrofission cross section for 237Np was measured over the energy range from 0,6 to 60,0 MeV. The multipolar composition of this cross section was investigated using the virtual photons formalism with three different techniques of analysis: unfolding and two versions of multiple parameter regression. (A.C.A.S.)

  16. Reliable determination of 237Np in environmental solid samples using 242Pu as a potential tracer

    Qiao, Jixin; Hou, Xiaolin; Roos, Per;

    2011-01-01

    , thereby demonstrating the usefulness of 242Pu as a non-isotopic tracer for 237Np chemical yield monitoring. The on-column separation procedure fosters rapid analysis as required in emergency situations since each individual sample can be handled within 2.5h, and leads to a significant decrease in labor...

  17. Radiochemical determination of 237NP in soil samples contaminated with weapon grade plutonium

    Antón, M. P.; Espinosa, A.; Aragón, A.

    2006-01-01

    The Palomares terrestrial ecosystem (Spain) constitutes a natural laboratory to study transuranics. This scenario is partially contaminated with weapon-grade plutonium since the burnout and fragmentation of two thermonuclear bombs accidentally dropped in 1966. While performing radiometric measurements in the field, the possible presence of 237Np was observed through its 29 keV gamma emission. To accomplish a detailed characterization of the source term in the contaminated area using the isotopic ratios Pu-Am-Np, the radiochemical isolation and quantification by alpha spectrometry of 237Np was initiated. The selected radiochemical procedure involves separation of Np from Am, U and Pu with ionic resins, given that in soil samples from Palomares 239+240Pu levels are several orders of magnitude higher than 237Np. Then neptunium is isolated using TEVA organic resins. After electrodeposition, quantification is performed by alpha spectrometry. Different tests were done with blank solutions spiked with 236Pu and 237Np, solutions resulting from the total dissolution of radioactive particles and soil samples. Results indicate that the optimal sequential radionuclide separation order is Pu-Np, with decontamination percentages obtained with the ionic resins ranging from 98% to 100%. Also, the addition of NaNO2 has proved to be necessary, acting as a stabilizer of Pu-Np valences.

  18. Effect of DTPA on concentration ratios of 237Np and 244Cm in vegetative parts of bush bean and barley

    Bush beans (Phaseolus vulgaris L.) and barley (Hordeum vulgare L.) were grown in two different soils (noncalcareous and calcareous) in containers in a glasshouse with 530 pCi/g 237Np or 15040 pCi/g 244Cm mixed into separate containers of the soil. The chelating agent DTPA at 100 μg/g soil was added to one half of the containers. The concentration ratio (C.R.) without DTPA was two orders of magnitude higher for 237Np than for 244Cm by two and three orders of magnitude but had little influence on 237Np. For the calcareous soil with DTPA, 244Cm C.R.'s were greater than those for 237Np. In bush beans both 237Np and 244Cm C.R.'s were higher in primary leaves than in trifoliate leaves which were higher than for stems

  19. Measurement of Delayed Neutron Yields from Thermal Neutron Induced Fission of $^{237}$Np

    Gundorin, N A; Pikelner, L B; Revrova, N V; Salamatin, I M; Smirnov, V I; Zhdanova, K V; Zhuchko, V E

    2005-01-01

    This paper reports about the measurement of delayed neutron yields from a thermal neutron induced fission of $^{237}$Np. The method based on periodic irradiation of the sample in pulsed neutron beam with the subsequent registration of neutrons in intervals between pulses is used in the experiment. The method is realized on the "Isomer-M" installation, located on the channel of the IBR-2 pulsed reactor. A description of the installation and a technique of the experiment are presented, a thorough analysis of background processes is performed, results of measurements are shown in this paper. The value of delayed neutron yields from thermal neutron induced fission of $^{237}$Np obtained in the present investigation is $\

  20. Covalency of Neptunium(IV) organometallics from 237Np Moessbauer spectra

    The isomer shifts in 237Np Moessbauer spectra arise from the shielding of neptunium's 6s orbitals by the inner 5f orbitals. In covalent bonding, ligand contributions to the 5f electron density increase the shielding, and the 237Np isomer shift reflects differences in bond character among covalently bonded ligands. The large difference in isomer shift (3.8 cm/sec) between ionic Np(IV) and Np(III) compounds permits a good determination of ligand bonding differences in Np(IV) organometallic compounds. The Moessbauer spectra for about 20 Np(IV) organometallic compounds, principally cyclopentadienyl (Cp) compounds of the general composition Cp/sub x/Np chi/sub 4-x/ (x = 1,2,3; chi = Cl, BH4, /sup n/Bu, Ph, OR, acac), show both the differences in sigma bonding among the chi ligands, as well as the covalent effect of the Cp ligands

  1. Evaluation of 237Np reaction amount by chemical analysis of neptunium sample irradiated at experiment fast reactor 'JOYO'

    The chemical analysis technique was established to determine the nuclide generated in Neptunium (Np) sample with a high accuracy, to contribute to evaluation of transmutation characteristics of 237Np in the fast reactor. (1) Establishment of Chemical Analysis Technique. The chemical analysis technique containing determination technique of fission amount of 237Np, which was consist of Vanadium (V) of capsule material removal and Neodymium (Nd) recovery at high efficiency, was established with optimization of experimental conditions. Four Np samples irradiated in 'JOYO' were analyzed using this technique. Results were as follows. 237Np were determined with high accuracy (relative error was 2.2% at maximum). Errors of fission amount monitoring nuclides 148Nd were half less than that of 137Cs. Small amount of 236Pu was able to determined. (2) Evaluation of 237Np Reaction Amount. The reaction amount of capture, fission and (n, 2n) reactions were evaluated using analyzed values. Transmutation characteristics of 237Np were evaluated using reaction amount. Evaluated results were as follows. The ratio of capture or fission amount to unirradiated 237Np amount were 6.1 - 25.5 at%, 0.7 - 3.6 at%, respectively. The 237Np (n, 2n) 236mNp reaction amount was 7.0 x 10-6 times of 237Np amount at maximum. The dependences of 237Np reaction amount to neutron energy spectrum were revealed from the fact such as linearity of fission to capture reaction amount ratio against fast neutron ratio in same fuel assembly. (author)

  2. Measurement of the Thermal Neutron Capture Cross section of 237Np for the Study of Nuclear Transmutation

    Full text of publication follows: Accurate nuclear data of minor actinides are required for the study of nuclear transmutation of radioactive wastes. The 237Np is one of the most important minor actinides for this study because of its relatively large abundance in irradiated fuels. However, there are apparent discrepancies between the reported neutron capture cross sections of the 237Np for thermal neutrons. History on the measurements of the neutron capture cross section of 237Np for thermal neutrons is briefly presented first. Recent three data measured by a γ ray spectroscopic method are much smaller than those measured by other methods. To deduce the neutron capture cross section by an activation method with γ ray spectroscopy, the relevant γ-ray emission probabilities are used. These decay data could be an origin of the discrepancies on the neutron capture cross section of 237Np. To examine the hypothesis, we measured the relevant γ-ray emission probabilities of 233Pa and 238Np from the ratio of the emission rate to the activity. The obtained emission probabilities are used to correct the thermal neutron capture cross section of 237Np reported previously. The cross section is also independently determined by irradiating 237Np sample in the research reactor of Kyoto University and counting α rays emitted from 237Np and 238Pu with a Si detector. The measured emission probabilities of 233Pa and 238Np, and the neutron capture cross section of 237Np are compared with others from references. The results of the precise decay data explain the discrepancy on the neutron capture cross section of 237Np. Details of the experiments and results will be presented. (authors)

  3. Electronic and structural properties of some ternary neptunium(VII) oxides from 237Np Moessbauer spectroscopy

    The 60 kev Moessbauer resonance of 237Np has been measured in some complex oxides of heptavalent neptunium. The nature of bonding and the molecular symmetry are discussed on the basis of the isomer shift and quadrupole coupling constant data. The molecular character of the compounds is evidenced by the low Debye temperatures and the strong bond covalency. The quadrupole coupling constant is temperature independent; this reveals the absence of any non-bonding states of f electrons. (authors)

  4. (237)Np(n,f) Cross Section: New Data and Present Status

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

    2011-01-01

    In this document, we present the final result obtained at the n_TOF experiment; for the neutron-induced fission cross section of the (237)Np, from the fission threshold up to 1 GeV. The method applied to get tins result is briefly discussed. n_TOF data are compared to the last experimental measurements using other TOF facilities or the surrogate method, reported experiments performed with monoenergetic sources and the FISCAL systematic, including a discussion about the existing discrepancies.

  5. Proton-induced fission on 241Am, 238U and 237Np at intermediate energies

    Deppman, A.; Andrade-II, E.; Guimaraes, V; Karapetyan, G. S.; Balabekyan, A. R.; Demekhina, N. A.

    2013-01-01

    Intermediate energy data of proton-induced fission on 241Am, 238U and 237Np targets were analysed and investigated using the computational simulation code CRISP. Inelastic interactions of protons on heavy nuclei and both symmetric and asymmetric fission are regarded. The fission probabilities are obtained from the CRISP code calculations by means of the Bohr-Wheeler model. The fission cross sections, the fissility and the number of nucleons evaporated by the nuclei, before and after fission, ...

  6. Measurements of neutron capture cross section of 237Np for fast neutrons

    The neutron capture cross section of 237Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product 238Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. 'Representative neutron energy' is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of 237Np corresponding to the representative neutron energy was deduced as 0.80±0.04 b at 214±9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of 237Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008. (author)

  7. New experimental determination of the neutronic resonance parameters of 237Np below 500 eV

    For studies of future nuclear reactors dedicated to nuclear waste transmutation, an improvement of the accuracy of the neutron radiative capture cross section of 237Np appears necessary. In the framework of a collaboration between the Commissariat a l'Energie atomique (CEA) and Institute for Reference Materials and Measurement (IRMM, Geel, Bergium), a new determination of the resonance parameters of 237Np has been performed. Two types of experiments are carried out at GELINA, the IRMM pulsed neutron source, using the time of flight method: a transmission experiment which is related to the neutron total cross section and a capture experiment which gives the neutron radiative capture cross section. The resonance parameters presented in this work are extracted from the transmission data between 0 and 500 eV with the least square code REFIT, using the Reich-Moore formalism. In parallel, the Doppler effect is investigated. The commonly used free gas model appears inadequate below 20 eV for neptunium dioxide at room temperature. By the use of the program DOPUSH, which calculates the Doppler broadening with a harmonic crystal model according to Lamb's theory, we are able to produce abetter fit of the experimental data for the resonances of 237Np in NpO2 at low energy or temperatures. In addition to the resonance parameters, a study of their mean value and distribution is included in this work. (authors)

  8. New radiochemical methods for determination of 237Np a 241Pu using extraction chromatography (Presentation)

    Thesis was focused on the development of a new methodology for the separation of anthropogenic transuranium radionuclides 237Np a 241Pu from different kinds of matrices. The analytical methods used in this study were based on extraction chromatography and were optimized according to the sample type. The proposed radiochemical procedure is a combination of two algorithms, which represent the separation of radionuclides by using extraction chromatographic sorbents TEVA resin and TRU resin supplied by Eichrom Technologies LLC. 239Np a 237Np were selectively captured on sorbent TEVA resin in oxidation state 4+. TRU resin was used for purification of plutonium fraction from interfering americium radionuclide. 242Pu and 239Np radionuclides as tracers have been used to monitor the radiochemical yields of separation. Before every radiochemical separation tracer radionuclide 239Np was obtained by separation from the parent radionuclide 243Am, which is in radioactive equilibrium to 239Np. The average yield of chemical separation was 69,3% for 239Np at 277 keV energy line and 65,9% at 228 keV energy line. The NPL AH-B08069 (2008) samples which consist of the mixture of alpha-radionuclides were used for the modification and optimization of separation method used for separation of Np and Pu in model samples. This method provided high radiochemical yields of 239,240Pu (95,0 ± 3,5)% and 237Np (87,9 ± 3,0)%.. Reliability of the method was verified by applying our modified separation procedures on reference materials IAEA-375 and IAEA-414 supplied by International Atomic Energy Agency. A good agreement between the results is obtained by this procedure and the certified values were found. Samples of contaminated soils from the area of Nuclear power plant A-1 Jaslovske Bohunice which is stored temporarily before disposal were analyzed using developed separation procedure. Specific activity of investigated radionuclides was determined in these samples. (author)

  9. New radiochemical methods for determination of 237Np a 241Pu using extraction chromatography

    Thesis was focused on the development of a new methodology for the separation of anthropogenic transuranium radionuclides 237Np a 241Pu from different kinds of matrices. The analytical methods used in this study were based on extraction chromatography and were optimized according to the sample type. The proposed radiochemical procedure is a combination of two algorithms, which represent the separation of radionuclides by using extraction chromatographic sorbents TEVA resin and TRU resin supplied by Eichrom Technologies LLC. 239Np a 237Np were selectively captured on sorbent TEVA resin in oxidation state 4+. TRU resin was used for purification of plutonium fraction from interfering americium radionuclide. 242Pu and 239Np radionuclides as tracers have been used to monitor the radiochemical yields of separation. Before every radiochemical separation tracer radionuclide 239Np was obtained by separation from the parent radionuclide 243Am, which is in radioactive equilibrium to 239Np. The average yield of chemical separation was 69,3% for 239Np at 277 keV energy line and 65,9% at 228 keV energy line. The NPL AH-B08069 (2008) samples which consist of the mixture of alpha-radionuclides were used for the modification and optimization of separation method used for separation of Np and Pu in model samples. This method provided high radiochemical yields of 239,240Pu (95,0 ± 3,5)% and 237Np (87,9 ± 3,0)%.. Reliability of the method was verified by applying our modified separation procedures on reference materials IAEA-375 and IAEA-414 supplied by International Atomic Energy Agency. A good agreement between the results is obtained by this procedure and the certified values were found. Samples of contaminated soils from the area of Nuclear power plant A-1 Jaslovske Bohunice which is stored temporarily before disposal were analyzed using developed separation procedure. Specific activity of investigated radionuclides was determined in these samples. (author)

  10. RAPID SEPARATION METHOD FOR 237NP AND PU ISOTOPES IN LARGE SOIL SAMPLES

    Maxwell, S.; Culligan, B.; Noyes, G.

    2010-07-26

    A new rapid method for the determination of {sup 237}Np and Pu isotopes in soil and sediment samples has been developed at the Savannah River Site Environmental Lab (Aiken, SC, USA) that can be used for large soil samples. The new soil method utilizes an acid leaching method, iron/titanium hydroxide precipitation, a lanthanum fluoride soil matrix removal step, and a rapid column separation process with TEVA Resin. The large soil matrix is removed easily and rapidly using this two simple precipitations with high chemical recoveries and effective removal of interferences. Vacuum box technology and rapid flow rates are used to reduce analytical time.

  11. Study of 237Np photonuclear reactions near threshold, induced by gamma rays from thermal neutron capture

    The photodisintegration of 237Np has been studied using monochromatic photons produced by thermal neutron capture in several materials. The partial cross sections σ gamma, sub(f) and σ gamma, sub(n) were measured in the energy interval from 5.43 MeV to 10.83 MeV. Analysing the photofission data according to the liquid drop model, the height (E sub(f)) and the curvature ((h/2π)ω) of the simple fission barrier were determined: E sub(f) = (5.9 +- 0.2) MeV and (h/2π)ω = (0.8 +- 0.4) MeV. For the competition between photoneutron emission and fission (GAMMA sub(n) / GAMMA sub(f) a constant value was found (1.28+- 0.15) in the energy range 6.73 - 10.83 MeV. From this result the following nuclear temperatures for 237Np were extracted on bases of some models of levels density: T = 0.84 +- 0.06 MeV (Fujimoto-Yamaguchi model) and T = 0.60 +- 0.04 MeV (Constant Nuclear temperature model). (Author)

  12. Isomeric ratio and cross section of the 237Np(n, 2n) reaction

    It has been shown that on the basis of the Hauser-Feshbach theory, and taking into account the law of the conservation of moment, the isomer ratio for the reaction 237Np (n, 2n) can be calculated. The indeterminateness of the modeling of the 236Np level scheme, to all appearances, has little effect on the energy dependence of the isomeric ratio. The error in the calculated cross section for the (n, 2n) reaction is determined chiefly by the error in the experimental data on the isomeric ratio and on the cross section for the formation of the short-lived state. Obtaining a correct estimate of the error is made difficult by the scarcity of experimental data on the isomeric ratio. The results of this work can be useful in practical activity when combined with an estimate of the cross sections and the creation of a complete system of neutron cross sections for 237Np. Theoretical estimates of the cross sections can to a significant extent compensate for the scarcity and indeterminateness of the experimental data

  13. Study of the excited levels of 233Pa by the 237Np alpha decay

    The excited levels in 233Pa following the 237Np alpha decay have been studied, by performing different experiences to complete available data and supply new information. Thus, two direct alpha spectrum measurement, one alpha-gamma bidimensional coincidence experiment, three gamma-gamma and gamma-X ray coincidences and some other measurements of the gamma spectrum, direct and coincident with alpha-particles have been made. These last experiences have allowed to obviate usual radiochemical separation methods, the 233Pa radioactive descendent interferences being eliminated by means of the coincidence technic. As a result, a primary decay scheme has been elaborated, including 15 new gamma transitions and two new levels, not observed in the most recent works. (Author) 60 refs

  14. Improvement of evaluated neutron nuclear data for 237Np and 241Am

    The nuclear data of 237Np and 241Am that are particularly important among the minor actinides were investigated by comparing JENDL-3.2 with the recent evaluated data and available experimental data. As a result of the study, several defects of JENDL-3.2 data were revealed. They were improved on the basis of experimental data or recent evaluated data. For the both nuclides, main quantities revised in the present work were the resonance parameters, cross sections, angular and energy distributions of secondary neutrons, number of neutrons per fission. The data were given in the neutron energy range from 10-5 eV to 20 MeV, and compiled in the ENDF-6 format. (author)

  15. A comparative study of the carcinogenetic effects of 241Am, 239Pu and 237Np

    In this experiment, 420 wistar rats were used to study the comparative carcinogenetic effects of 241Am, 239Pu and 237Np. These nuclides were injected to animals intravenously, subcutaneously or directly into the lung (Stansen's lung puncture method) in doses of 1.0, 5.0 and 8.5 μCi/kg, respectively. As soluble nitrate, the nuclides were rapidly transfered from the site of injection into the bone and the liver. Osteosarcomas were found in some animals 8 months to one year after intoxication. Diagnosis of osteosarcoma is based on the histopatological examination and X-ray photography. In the Am-poisoned rats the incidence of osteosarcoma is about 31-74%, varied with different doses and different routes of intoxication; in Pu-poisoned rats, the incidence of osteosarcoma is about 55-66%. while in Np-poisoned rats, it is about 36-53%. Primary lung cancers were also found in those animals poisoned by means of Stansen's lung puncture method with the above three nuclides. The incidence of primary lung cancers is about 6% in Am-and Pu-poisoned rats and 13% in Np-poisoned rats. The incidence of metastasis of osteosarcoma in lung is about 25-65% for Am-poisoned rats, 45-55% for Pu-poisoned rats and 41-80% for Np-poisoned rats. The life-span of above poisoned rats was significantly shorter than that of the normal control animals. The chemical weight for 241Am, 239Pu and 237Np in same unit of radioactivity (1.0 μCi) equals to 0.308 μg, 15.9 μg and 1418.7 μg, respectively. For this reason, we have to pay more attention to the chemical mass effect in carcinogenesis of the above three nuclides

  16. Measurement of the fission cross-section ratio for 237Np/235U around 14 MeV neutron energies

    Fission cross-section ratio was determined for 237Np/235U around 14 MeV neutron energies with a back-to-back ionization chamber. Neutrons were produced by a 180 KV accelerator using T(d,n)4He reaction. No significant energy dependence was found in the cross section ratio

  17. DETERMINATION OF 237NP AND PU ISOTOPES IN LARGE SOIL SAMPLES BY INDUCTIVELY COUPLED PLASMA MASS SPECTROMETRY

    Maxwell, S.

    2010-07-26

    A new method for the determination of {sup 237}Np and Pu isotopes in large soil samples has been developed that provides enhanced uranium removal to facilitate assay by inductively coupled plasma mass spectrometry (ICP-MS). This method allows rapid preconcentration and separation of plutonium and neptunium in large soil samples for the measurement of {sup 237}Np and Pu isotopes by ICP-MS. {sup 238}U can interfere with {sup 239}Pu measurement by ICP-MS as {sup 238}UH{sup +} mass overlap and {sup 237}Np via {sup 238}U peak tailing. The method provides enhanced removal of uranium by separating Pu and Np initially on TEVA Resin, then transferring Pu to DGA resin for additional purification. The decontamination factor for removal of uranium from plutonium for this method is greater than 1 x 10{sup 6}. Alpha spectrometry can also be applied so that the shorter-lived {sup 238}Pu isotope can be measured successfully. {sup 239}Pu, {sup 242}Pu and {sup 237}Np were measured by ICP-MS, while {sup 236}Pu and {sup 238}Pu were measured by alpha spectrometry.

  18. Development of ionization technique for measurement of fast neutron induced fission products yields of {sup 237}Np

    Goverdovski, A.A.; Khryachkov, V.A.; Ketlerov, V.V.; Mitrofanov, V.F.; Ostapenko, Yu.B.; Semenova, N.N.; Fomichev, A.N.; Rodina, L.F. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-03-01

    Twin gridded ionization chamber and corresponding software was designed for measurements of masses, kinetic energies and nuclear charges of fission fragments from fast neutron induced fission of {sup 237}Np. The ionization detector design, electronics, data acquisition and processing system and the test results are presented in this paper. (J.P.N.)

  19. 231Pa and 233Pa neutron-induced fission up to 20 MeV

    Consistency of neutron-induced fission cross section data of 231Pa and 233Pa and data extracted from transfer reactions is investigated. Present estimate of 233Pa(n,f) fission cross section is supported by smooth level density parameter systematic, validated in case of 231Pa(n,f) data description up to En=20 MeV. The fission probabilities of Pa, fissioning in 231,233Pa(n,nf) reactions, are defined by fitting (3He,d) or (3He,t) transfer reaction data

  20. Fission mode analysis of the reaction {sup 237}Np(n,f) - possibilities and perspectives

    Siegler, P. [Joint Research Centre, Geel (Belgium). Geel Establishment

    1996-03-01

    Fission fragment properties for the reaction {sup 237}Np(n,f) have been measured at the Van de Graaff Laboratory of the IRMM. Using a double gridded ionization chamber the mass, kinetic energy and the angular distribution for both fission fragments could be determined simultaneously for an incident neutron energy range from E{sub n}=0.3 MeV upto E{sub n}=5.5 MeV. Complete datasets have been acquired for 13 different neutron energies covering sub barrier fission as well as fission in the plateau region. A detailed analysis of the fragment distributions and the respective momenta has been carried out, checking the coherence against the excitation energy of the compound nucleus. The consideration of multi-modal fission offers an improved possibility for the description of the fragment distributions backed up by theoretical calculations on the basis of the multi-model random-neck rupture model of Brosa, Grossmann and Mueller. The changes of the fission fragment properties under investigation are completely described and an interpretation of the findings is presented. (author)

  1. Removal of 230Th and 231Pa from the open ocean

    Concentrations of 230Th and 231Pa were measured in particulate matter collected by sediment traps deployed in the Sargasso Sea (Site S2), the north equatorial Atlantic (site E), and the north equatorial Pacific (Site P) as well as in particles collected by in situ filtration at Site E. Concentrations of dissolved Th and Pa were determined by extraction onto manganese dioxide adsorbers at Site P and at a second site in the Sargasso Sea (site D). Dissolved 230Th/231Pa activity ratios were 3-6 at Sites P and D. In contrast, for all sediment trap samples from greater than 2000 m, unsupported 230Th/231Pa ratios were 22-35 (average 29.7). Ratios were lower in particulate matter sampled at shallower depths. Particles filtered at 3600 m and 5000 m at Site E had ratios of 50 and 40. Results show that suspended particulate matter in the open ocean preferentially scavenges Th relative to Pa. Most of the 230Th produced by decay of 234U in the open ocean is removed by adsorption to settling particulate matter. In contrast, less than 50% of the 231Pa produced by decay of 235U is removed from the water column by this mechanism. Mixing processes transport the remainder to other sinks. (orig.)

  2. New experimental determination of the neutronic resonance parameters of {sup 237}Np below 500 eV; Nouvelle determination experimentale des parametres de resonances neutroniques de {sup 237}Np en dessous de 500 eV

    Gressier, V

    1999-10-01

    For studies of future nuclear reactors dedicated to nuclear waste transmutation, an improvement of the accuracy of the neutron radiative capture cross section of {sup 237}Np appears necessary. In the framework of a collaboration between the Commissariat a l'Energie atomique (CEA) and Institute for Reference Materials and Measurement (IRMM, Geel, Bergium), a new determination of the resonance parameters of {sup 237}Np has been performed. Two types of experiments are carried out at GELINA, the IRMM pulsed neutron source, using the time of flight method: a transmission experiment which is related to the neutron total cross section and a capture experiment which gives the neutron radiative capture cross section. The resonance parameters presented in this work are extracted from the transmission data between 0 and 500 eV with the least square code REFIT, using the Reich-Moore formalism. In parallel, the Doppler effect is investigated. The commonly used free gas model appears inadequate below 20 eV for neptunium dioxide at room temperature. By the use of the program DOPUSH, which calculates the Doppler broadening with a harmonic crystal model according to Lamb's theory, we are able to produce abetter fit of the experimental data for the resonances of {sup 237}Np in NpO{sub 2} at low energy or temperatures. In addition to the resonance parameters, a study of their mean value and distribution is included in this work. (authors)

  3. RAPID DETERMINATION OF 237 NP AND PU ISOTOPES IN WATER BY INDUCTIVELY COUPLED PLASMA MASS SPECTROMETRY AND ALPHA SPECTROMETRY

    Maxwell, S.; Jones, V.; Culligan, B.; Nichols, S.; Noyes, G.

    2010-06-23

    A new method that allows rapid preconcentration and separation of plutonium and neptunium in water samples was developed for the measurement of {sup 237}Np and Pu isotopes by inductively-coupled plasma mass spectrometry (ICP-MS) and alpha spectrometry; a hybrid approach. {sup 238}U can interfere with {sup 239}Pu measurement by ICP-MS as {sup 238}UH{sup +} mass overlap and {sup 237}Np via peak tailing. The method provide enhanced removal of uranium by separating Pu and Np initially on TEVA Resin, then moving Pu to DGA resin for additional removal of uranium. The decontamination factor for uranium from Pu is almost 100,000 and the decontamination factor for U from Np is greater than 10,000. This method uses stacked extraction chromatography cartridges and vacuum box technology to facilitate rapid separations. Preconcentration is performed using a streamlined calcium phosphate precipitation method. Purified solutions are split between ICP-MS and alpha spectrometry so that long and short-lived Pu isotopes can be measured successfully. The method allows for simultaneous extraction of 20 samples (including QC samples) in 4 to 6 hours, and can also be used for emergency response. {sup 239}Pu, {sup 242}Pu and {sup 237}Np were measured by ICP-MS, while {sup 236}Pu, {sup 238}Pu, and {sup 239}Pu were measured by alpha spectrometry.

  4. 231Pa and 233Pa Neutron-Induced Fission Data Analysis

    The 231Pa and 233Pa neutron-induced fission cross-section database is analyzed within the Hauser-Feshbach approach. The consistency of neutron-induced fission cross-section data and data extracted from transfer reactions is investigated. The fission probabilities of Pa, fissioning in 231,233Pa(n,nf) reactions, are defined by fitting (3He,d) or (3He,t) transfer-reaction data. The present estimate of the 233Pa(n,f) fission cross section above the emissive fission threshold is supported by smooth level-density parameter systematics, validated in the case of the 231Pa(n,f) data description up to En =20 MeV

  5. Some spectroscopic properties of fine structures observed near the 231Pa(n,f) fission threshold

    The 231Pa neutron-induced fission cross section from 140 to 400 keV was resolved into finer structures. For some of the fractionated vibrational resonances in this energy region, the assignment of spectroscopic parameters may support evidence for an asymmetrically deformed third minimum in the 232Pa fission barrier. Also, for the first time, narrow fission resonances are observed above 1.3 eV exhibiting an average fission width /sub obs/ = 8meV

  6. Irradiation studies of 231Pa in DHRUVA reactor for preparation of 232U

    In this paper, identification of the potential source of 231Pa, various stages of its recovery and the sample of 231Pa supplied for trial irradiation are described. For producing 232U, irradiation work is proposed in different stages of irradiation. Various trial irradiations and its results are discussed in this paper along with their specific objectives. A computational estimation of conversion efficiency of 231Pa (∼ 10μg) to 232U in DHRUVA reactor, was carried out. The computational result predicted the conversion efficiency and the results of γ-spectroscopy based analysis of the second trial sample matched well. A first level calculation was also carried out to estimate the tolerable quantity of 232Th to keep the concentration of 233U below 5% in the 232U sample. The radioactivity due to the presence of various impurities present in the sample, during the high fluence irradiation, is being estimated. The feasibility of production of the 232U isotope in DHRUVA reactor has thus been established. (author)

  7. Uranium age determination: Separation and analysis of 230Th and 231Pa

    In this work we focused on the age determination of uranium materials of different uranium enrichment. The radioactive decay of the uranium isotopes provides a chronometer that is inherent to the material, in particular the mother/daughter pairs 234U/230Th and 235U/231Pa can be advantageously used. Due to the relatively long half-lives of 234U (2.46 · 105 years) and 235U (7.04 · 108 years) only minute amounts of daughter nuclides are growing in, therefore both separation of Th and Pa from uranium must be of high chemical recovery and must afford large decontamination factors. Analytical methods for the age determination of uranium samples using the parent/daughter relations 234U/230Th and 235U/231Pa is demonstrated. Thorium is separated from bulk uranium using extraction chromatography and subsequently quantified using square-spectrometry, thermal ionisation mass spectrometry (TIMS) and inductively coupled mass spectrometry (ICP-MS). Protactinium is separated by highly selective sorption of protactinium to silica gel followed by square-spectrometric quantification. The methods were tested and validated using uranium reference materials of different uranium enrichment and of known ages. The experimental results obtained with both methods were found to agree with the assumed ages of the reference materials within the combined uncertainty of the measurement. The analysis exploiting the parent/daughter pair 235U/231Pa exhibits a slightly larger combined uncertainty and bias than the thorium method but is found valuable in validating the experimental results by means of a second, independent analysis

  8. The rotational bands in the nuclei 229Pa and 231Pa

    Experimental evidence is presented for the similarity between the rotational spectra built on the 1/2-[530] state in 231Pa, where the 3/2- member of the band forms the ground state, and in 229Pa, where this state lies within 20 keV of the ground state. Our findings are in contrast with earlier work invoking octupole deformations in the ground state to account for the different positions of low-lying Nilsson states in the two isotopes. (author)

  9. NKS-Norcmass reference material for analysis of Pu-isotopes and 237Np by mass spectrometry

    The aim of the reference material in the Norcmass-project was to produce a low-level (239Pu) sample of sufficient amount to allow individual laboratories to perform several tests without risk of using up the material. Although there are several reference materials available (eg IAEA) few have 239Pu/240Pu data and almost none have 237Np/239Pu-data. Those who have (eg IAEA-384) have very high concentrations and are not useful for testing analytical methods designed for low-level measurements where a large sample mass may be required. The reference material consist of the top 10cm of 2mm sieved soil pooled together from 12 different Danish locations collected during 2003. The Soil was blended and sieved through 0.6 and finally through a 0.4 mm sieve. A total amount of 17 kg soil was produced. Several aliquots of the material was subject to analysis by alpha spectrometry and ICP-MS. The material contain 239+240Pu at a concentration of 0.24 ± 0.01 mBq/g and a 240Pu/239Pu atom ratio of 0.19 ± 0.006. The ratio 237Np/239Pu was determined to 0.32 ± 0.01. (au)

  10. Emission probabilities of the KX-rays following the decay of 237 Np in equilibrium with 233 Pa

    Following participation in the international EUROMET project No. 416 and after our recent paper, concerning the measurement of the emission probability values of the main gamma-rays of 237 Np in equilibrium with 233 Pa, a complementary work has been done in the frame of the collaboration LNHB-VNIIM-KRI-IFIN (with the support of 'Ministere des Affaires Etrangeres' of France). The purpose was to determine the photon emission probabilities for the KX-rays following the decay of these two nuclides. Two different analysis methods have been used. At first, the KX-rays region was analyzed by fitting Voigt functions according to a least squares procedure, included in 'COLEGRAM' deconvolution code. In the second case, the analysis was performed by using full response functions. Thus, the work allowed the determination of the photon emission probabilities with a relative uncertainty of about 2%. This accurate set of data is useful in calculations related to the atomic level scheme of 237 Np/233 Pa and in X-ray spectrometry based applications. (authors)

  11. Inductively Coupled Plasma Mass Spectrometry For The Determination Of 237Np In Spent Nuclear Fuel Samples By Isotope Dilution Method Using 239Np As A Spike

    A determination method for 237Np in spent nuclear fuel samples was developed using an isotope dilution method with 239Np as a spike. In this method, inductively coupled plasma mass spectrometry (ICP-MS) was taken for the 237Np instead of the previously used alpha spectrometry. 237Np and 239Np were measured by ICP-MS and gamma spectrometry, respectively. The recovery yield of 237Np in synthetic samples was 95.9±9.7% (1S, n=4). The 237Np contents in the spent fuel samples were 0.15, 0.25, and 1.06 μg/mgU and these values were compared with those from ORIGEN-2 code. A fairly good agreement between the measurements (m) and calculations (c) was obtained, giving ratios (m/c) of 0.93, 1.12 and 1.25 for the three PWR spent fuel samples with burnups of 16.7, 19.0, and 55.9 GWd/MtU, respectively

  12. Rapid determination of 237Np in soil samples by multi-collector inductively-coupled plasma mass spectrometry and gamma spectrometry

    A radiochemical procedure is developed for the determination of 237Np in soil with multi-collector inductively-coupled plasma mass spectrometry (MC-ICP-MS) and gamma-spectrometry. 239Np (milked from 243Am) was used as an isotopic tracer for chemical yield determination. The neptunium in the soil is separated by thenoyl-trifluoracetone extraction from 1 M HNO3 solution after reducing Np to Np(IV) with ferrous sulfamate, and then purified with Dowex 1 x 2 anion exchange resin. 239Np in the resulting solution is measured with gamma-spectrometry for chemical yield determination while the 237Np is measured with MC-ICP-MS. Measurement results for soil samples are presented together with those for two reference samples. By comparing the determined value with the reference value of the 237Np activity concentration, the feasibility of the procedure was validated. (author)

  13. Testing the 231Pa/230Th paleo-circulation proxy: A data versus 2D model comparison

    Variations of the Atlantic Meridional Overturning Circulation (AMOC) are believed to have crucially influenced Earth's climate due to its key role in the inter-hemispheric redistribution of heat and carbon. To assess its past strength, the sedimentary 231Pa/230Th proxy has been developed and improved but also contested due to its sensitivity to other factors beyond ocean circulation. In order to provide a better basis for the understanding of the Atlantic 231Pa/230Th system, and therefore to shed light on the controversy, we compare new measurements of Holocene sediments from the north Brazilian margin to water column data and the output of a two-dimensional scavenging-circulation model, based on modern circulation patterns and reversible scavenging parameters. We show that sedimentary 231Pa/230Th data from one specific area of the Atlantic are in very good agreement with model results suggesting that sedimentary 231Pa/230Th is predominantly driven by the AMOC. Therefore, 231Pa/230Th represents an appropriate method to reconstruct past AMOC at least qualitatively along the western margin. (authors)

  14. 240Pu(n,f), 242Pu(n,f), 237Np(n,f), neutron fission cross sections, Esub(n) = 2.5 MeV

    Measurements of the absolute neutron fission cross section of 240Pu, 242Pu and 237Np have been made at 2.5 MeV using a hybrid detector. The fission events were detected in an ionization chamber (2π) and the neutron flux was determined by a proton recoil telescope and a directional long counter. Our values are compared to previous data

  15. Fission Cross Sections of {209}Bi, {232}Th, {235}U, {238}U and {237}Np for Intermediate Energy Protons and Deuterons

    Yurevich, V I; Yakovlev, R M; Sosnin, A N

    2001-01-01

    Fission cross sections of {209}Bi, {232}Th, {235}U, {238}U and {237}Np have been measured with 1.0-3.7 GeV protons and 1.0 GeV deuterons. The results are compared with other experimental data, available evaluations and predictions of the theoretical model.

  16. Uranium age determination - Separation and analysis of 230Th and 231Pa

    Full text: In recent years several incidents involving illicit trafficking and smuggling of nuclear material, radioactive sources and radioactively contaminated materials have raised growing public concern about criminal acts involving nuclear materials. Consequently, research efforts in nuclear forensic science have been intensified in order to develop and improve methods for the identification of the nature and origin of seized materials. Information obtained from the analysis of unknown nuclear materials is of key importance in order to aide authorities that are in charge of developing fast and appropriate response action. For the identification of nuclear materials various sample characteristics are of relevance, including isotopic composition, the content of chemical impurities, material properties and the date of production. Information on the production date, respectively the 'age' of nuclear materials, will also be of key importance in other fields of nuclear science, i.e. for the verification of a Fissile Materials Cut-Off Treaty (FMCT) in order to distinguish freshly produced materials from 'old' excess weapons materials. The age of nuclear materials may also be of relevance under a strengthened safeguards regime to reveal clandestine production of weapons usable materials, i.e. the separation of plutonium or production of highly enriched uranium (HEU). The age dating of plutonium samples has been described in detail for bulk samples as well as for particles. In this work we focused on the age determination of uranium materials of different uranium enrichment. The radioactive decay of the uranium isotopes provides a chronometer that is inherent to the material, in particular the mother/daughter pairs 234U/230Th and 235U/231Pa can be advantageously used. Due to the relatively long half-lives of 234U (2.46·105 years) and 235U (7.04·108 years) only minute amounts of daughter nuclides are growing in, therefore both separation of Th and Pa from uranium must

  17. Exposure to radioactive aerosols in mining and milling operations: the importance of 227Ac and 231Pa

    'Full-Text:' 227Ac (half-life of 21.8 y) is the daughter or 231Pa; a beta emitter and parent of a subseries with five short-lived alpha-emitters. 231Pa (half-life of 3.27x104 y) is an alpha-emitter of the actinium series, the decay chain of 235U. As daughters of this uranium isotope they are thought to be unimportant as a radiological hazard, despite the fact that their ALI values for ingestion and inhalation are the lowest of any other radionuclide. Both nuclides can be considered as being in secular equilibrium with uranium in most geological media and so the mass concentration of 231Pa is the same of the 226Ra and that of 227Ac is the same of the 210Pa, to mention only two radionuclides of radiological concern. It is shown in this paper that if 231Pa and 227Ac are considered in the evaluations of dose commitments incurred by inhalation of aerosols in mining and milling operations, the results can be 70% higher than those calculated by the methodology of ICRP Publication 47. (author)

  18. Mass and nuclear charge yields for sup 237 Np(2n sub th ,f) at different fission fragment kinetic energies

    Martinez, G.; Barreau, G.; Sicre, A.; Doan, T.P.; Audouard, P.; Leroux, B. (CEA Centre d' Etudes Nucleaires de Bordeaux-Gradignan, 33 - Gradignan (France)); Arafa, W.; Brissot, R.; Bocquet, J.P. (Grenoble-1 Univ., 38 (France). Inst. des Sciences Nucleaires); Faust, H. (Institut Max von Laue - Paul Langevin, 38 - Grenoble (France)); Koczon, P.; Mutterer, M. (Technische Hochschule Darmstadt (Germany, F.R.). Inst. fuer Kernphysik); Goennenwein, F. (Tuebingen Univ. (Germany, F.R.). Physikalisches Inst.); Asghar, M. (Universite des Sciences et de la Technologie Houari Boumediene, Algiers (Algeria). Inst. de Physique); Quade, U.; Rudolph, K. (Muenchen Univ. (Germany, F.R.)); Engelhardt, D. (Karlsruhe Univ. (T.H.) (Germany, F.R.)); Piasecki, E. (Warsaw Univ. (Poland))

    1990-09-03

    The recoil mass separator LOHENGRIN of the Laue-Langevin Institute Grenoble has been used to measure for the first time, the yields of light fission fragments from the fissioning system: {sub 93}{sup 239}Np; this odd-Z nucleus is formed after double thermal neutron capture in a {sub 93}{sup 237}Np target. The mass distributions were measured for different kinetic energies between 92 and 115.5 MeV, but the nuclear charge distributions were determined only up to 112 MeV. These distributions are compared to the distributions obtained for the even-even system {sub 94}{sup 240}Pu. At high kinetic energy, the mass distribution shows a prominent peak around mass number A{sub L}=106. These cold fragmentations are discussed in terms of a calculation based on a scission point model extrapolated to the cold fission case. As expected for an odd-Z fissioning nucleus, the nuclear charge distributions do not reveal any odd-even effect. The global neutron odd-even effect is found to be (8.1{plus minus}1.5)%. A simple model has been used to show that most of the neutron odd-even effect results from prompt neutron evaporation from the fragments. (orig.).

  19. Neutron-induced transmutation reactions in 237Np, 238Pu, and 239Pu at the massive natural uranium spallation target

    Transmutation reactions in the 237Np, 238Pu, and 239Pu samples were investigated in the neutron field generated inside a massive (m = 512 kg) natural uranium spallation target. The uranium target assembly QUINTA was irradiated with the deuteron beams of kinetic energy 2, 4, and 8 GeV provided by the Nuclotron accelerator at the Joint Institute for Nuclear Research (JINR) in Dubna. The neutron-induced transmutation of the actinide samples was measured off-line by implementing methods of gamma-ray spectrometry with HPGe detectors. Results of measurement are expressed in the form of both the individual reaction rates and average fission transmutation rates. For the purpose of validation of radiation transport programs, the experimental results were compared with simulations of neutron production and distribution performed by the MCNPX 2.7 and MARS15 codes employing the INCL4-ABLA physics models and LAQGSM event generator, respectively. In general, a good agreement between the experimental and calculated reaction rates was found in the whole interval of provided beam energies

  20. Measurement of transuranic elements, chiefly 237Np (by neutron activation analysis), in the physical and biological compartments of the French shore of the English Channel

    The behaviour of transuranic elements in the marine environment has been studied via both in situ sampling and laboratory tracer experiments. In particular, the radionuclide 237Np was investigated and techniques for its quantitative determination are described. In the field investigations, a preliminary separation of Np from samples was performed prior to neutron activation analysis, with subsequent γ-ray spectrometry of 238Np. In the laboratory studies, 237Np was determined by a radiochemical method followed by α-spectrometry. The results obtained from the in situ study in the English Channel (sea water, seaweed, molluscs) and from laboratory-uptake experiments (water and mussels) are described. Levels of Pu, Am and Np are compared and the characteristics of neptunium transfer to molluscs are discussed. (author)

  1. Emission probabilities of gamma rays from the decay of 233Pa and 238Np, and the thermal neutron capture cross section of 237Np

    In order to determine the thermal neutron capture cross section of 237Np, the relevant γ emission probabilities of the 312-keV γ-ray from the decay of 233Pa and the 984-keV γ-ray from the decay of 238Np are deduced from the ratio of the emission rate to the activity. The emission rate and activity are measured with a Ge detector and a Si detector, respectively. The measured emission probability for 312-keV γ-ray is 41.6±0.9% and that for 984-keV γ-ray is 25.2±0.5%. The emission probabilities are used to correct the thermal neutron capture cross section of 237Np reported previously, and gives 168±6b. The neutron capture cross section is also determined as 169±6b by α-ray spectroscopic method. The measured emission probabilities and capture cross section are compared with others from references. By averaging these values deduced by different methods, the value of 169±4b is recommended as the thermal neutron capture cross section of 237Np for 2,200 m/s neutrons. (author)

  2. Study of (n,p) and (n,α) cross-sections for 232Th, 231Pa, 233U isotopes

    The study of neutron induced reaction cross-sections in the charged particle emission in this energy region will help us to understand the energy dependence of activation cross-sections in detail, thereby providing a complete database that will lead to better understanding of mechanisms of the nuclear reactions. The present study describes nuclear model calculations of (n,p) and (n,α) reaction cross-sections for 232Th, 231Pa and 233U isotopes

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

    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: 234U, 235U, 237Np and 238Np. The experiment is led at the heavy-ion accelerator GSI in Darmstadt, Germany. This facility provides intense relativistic primary beam of 238U. 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 the

  4. Sediment 231Pa/230Th as a recorder of the rate of the Atlantic meridional overturning circulation: insights from a 2-D model

    S. E. Allen

    2009-11-01

    Full Text Available A two dimensional scavenging-circulation model is used to investigate the patterns of sediment 231Pa/230Th generated by the Atlantic Meridional Overturning Circulation (AMOC and further advance the application of this proxy for ocean paleocirculation studies. The scavenging parameters and the geometry of the overturning circulation cell have been chosen so that the model generates meridional sections of dissolved 230Th and 231Pa consistent with published water column profiles and an additional 12 previously unpublished profiles measured in the North and Equatorial Atlantic. The processes that generate the meridional sections of dissolved and particulate 230Th, dissolved and particulate 231Pa, dissolved and particulate 231Pa/230Th, and sediment 231Pa/230Th are discussed in detail. The results indicate that the relationship between sediment 231Pa/230Th at any given site and the overturning circulation is very complex. They clearly show that constraining past changes in the strength and geometry of the AMOC requires an extensive data set and they suggest strategies to maximize information from a limited number of samples.

  5. Sediment 231Pa/230Th as a recorder of the rate of the Atlantic meridional overturning circulation: insights from a 2-D model

    S. E. Allen

    2010-03-01

    Full Text Available A two dimensional scavenging model is used to investigate the patterns of sediment 231Pa/230Th generated by the Atlantic Meridional Overturning Circulation (AMOC and further advance the application of this proxy for ocean paleocirculation studies. The scavenging parameters and the geometry of the overturning circulation cell have been chosen so that the model generates meridional sections of dissolved 230Th and 231Pa consistent with published water column profiles and an additional 12 previously unpublished profiles measured in the North and Equatorial Atlantic. The processes that generate the meridional sections of dissolved and particulate 230Th, dissolved and particulate 231Pa, dissolved and particulate 231Pa/230Th, and sediment 231Pa/230Th are discussed in detail. The results indicate that the relationship between sediment 231Pa/230Th at any given site and the overturning circulation is very complex. They clearly show that constraining past changes in the strength and geometry of the AMOC requires an extensive data set and they suggest strategies to maximize information from a limited number of samples.

  6. Measurement at n-TOF of the 237Np(n, γ) and 240Pu(n, γ) cross sections for the transmutation of nuclear waste

    The final design, safety assessment and precise performance analysis of transmutation devices such as Accelerator Driven Systems (ADS) or Fast Critical Reactors, need accurate and reliable nuclear data. The cross sections of 237Np and 240Pu have been measured in 2004 at n-TOF with good accuracy due to a combination of features unique in the world: high instantaneous neutron fluence and excellent energy resolution of the n-TOF facility [1], innovative Data Acquisition System based on flash ADCs and the use of a high performance BaF2 Total Absorption Calorimeter as a detection device. (authors)

  7. Rapid determination of 237Np in soil samples by multi-collector inductively-coupled plasma mass spectrometry and gamma spectrometry

    Yi, Xiaowei; Shi, Yanmei; Xu, Jiang; He, Xiaobing; ZHANG, HAITAO; Lin, Jianfeng

    2013-01-01

    A radiochemical procedure is developed for the determination of 237Np in soil with multi-collector inductively-coupled plasma mass spectrometry (MC-ICP-MS) and gamma-spectrometry. 239Np (milked from 243Am) was used as an isotopic tracer for chemical yield determination. The neptunium in the soil is separated by thenoyl-trifluoracetone extraction from 1 M HNO3 solution after reducing Np to Np(IV) with ferrous sulfamate, and then purified with Dowex 1 × 2 anion exchange resin. 239Np in the resu...

  8. The neutron capture cross sections of 237Np(n,γ) and 240Pu(n,γ) and its relevance in the transmutation of nuclear waste

    Neutron capture cross sections of actinides are of great relevance for the Transmutation of Nuclear Waste in Accelerator Driven Systems (ADS) and Generation-IV reactors. The neutron capture cross sections of 237Np and 240Pu in the range of 1 eV to 2 keV were measured at the n-TOF facility with a Total Absorption Calorimeter. The data have been analyzed with the SAMMY code. The corresponding covariance matrices have been generated. The final cross sections are presented and compared to the previously existing ones.The n-TOF 237Np σ(n,γ) is in agreement with the evaluated data files below 300 eV and its is lower by 10 to 15% up to 2 keV. This discrepancy with the evaluated data files is also observed in the capture cross section derived from the transmission measurements of Gressier et al. In the case of the 240Pu σ(n,γ), the n-TOF σ(n,γ) agrees within uncertainties with JENDL-3.3 and JEFF-3.1, except for a group of resonances around 800 eV. Endf/B-VII data are lower than n-TOF and the mentioned evaluations, with differences that increase with neutron energy up to 15-20 per cent

  9. The neutron capture cross sections of {sup 237}Np(n,{gamma}) and {sup 240}Pu(n,{gamma}) and its relevance in the transmutation of nuclear waste

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

    2008-07-01

    Neutron capture cross sections of actinides are of great relevance for the Transmutation of Nuclear Waste in Accelerator Driven Systems (ADS) and Generation-IV reactors. The neutron capture cross sections of {sup 237}Np and {sup 240}Pu in the range of 1 eV to 2 keV were measured at the n-TOF facility with a Total Absorption Calorimeter. The data have been analyzed with the SAMMY code. The corresponding covariance matrices have been generated. The final cross sections are presented and compared to the previously existing ones.The n-TOF {sup 237}Np {sigma}(n,{gamma}) is in agreement with the evaluated data files below 300 eV and its is lower by 10 to 15% up to 2 keV. This discrepancy with the evaluated data files is also observed in the capture cross section derived from the transmission measurements of Gressier et al. In the case of the {sup 240}Pu {sigma}(n,{gamma}), the n-TOF {sigma}(n,{gamma}) agrees within uncertainties with JENDL-3.3 and JEFF-3.1, except for a group of resonances around 800 eV. Endf/B-VII data are lower than n-TOF and the mentioned evaluations, with differences that increase with neutron energy up to 15-20 per cent.

  10. Delayed neutron and delayed photon characteristics from photofission of 232Th, 235,238U, and 237Np with endpoint Bremsstrahlung photons in the giant dipole resonance region

    A renewed interest in photonuclear reactions was stimulated by applications as radioactive ion beam production, irradiation stations by high energy photons, shielding of electron accelerators, etc. Today, a particular attention is paid to the non-destructive characterization of waste barrels and the detection of nuclear materials, both based on photofission process and the associated delayed neutron (DN) and delayed photon (DP) emissions. The need of accurate and complete data for DN and DP yields and time characteristics of actinides was the motivation for an experimental campaign, started in 2004. In this paper, the experimental setup and the data analysis method will be presented and the modeling work will be described. Experimental results for DN and DP characteristics will be compared to calculations in the case of photofission of 232Th, 235,238U, and 237Np. (authors)

  11. Mass asymmetry dependence of fusion time-scales in 11B+237Np and 12C, 16O, 19F+232Th reactions in a dynamical trajectory model

    Dynamical trajectory calculations were carried out for the reactions of 11B+237Np and 12C, 16O and 19F+232Th, having mass asymmetries on either side of the Businaro-Gallone critical mass asymmetry αBG, in order to examine the mass asymmetry dependence of fusion reactions in these systems. The compound nucleus formation times were calculated as a function of the partial wave of the reaction for all the systems. This study brings out that for systems with αBG, the formation times are significantly larger than for α>αBG, which is caused by the dynamical effects involved in the large scale shape changes taking place in the fusion process as well as due to the interplay between the thermal and the collective motion during the collision process. The calculated time scales are comparable to the experimental values derived from the pre-fission neutron multiplicity measurements. (author). 16 refs., 4 figs., 1 tab

  12. Calculation of fusion time scales in 11B + 237Np, 12C + 232Th and 16O + 232Th reactions in a dynamical trajectory model

    There are several theoretical models which treat the fusion process and energy dissipation in heavy ion collision in terms of a fluctuating force represented by the coupling between macroscopic and intrinsic degrees of freedom. One such dynamical model has been developed by Feldmeier (1987), where the properties of the dissipative force are determined from a microscopic picture of particle exchange between two nuclei. The macroscopic shapes of the nuclear system are represented by axially symmetric configuration with sharp surfaces. We have used the above model to calculate the fusion time scales for the systems 11B + 237Np, 12C +232Th and 16O + 232Th at 77, 86 and 104 MeV bombarding energies to examine the effect of mass asymmetry in fusion dynamics. (author). 2 figs

  13. Transmutation of 129I and 237Np using spallation neutrons produced by 1.5, 3.7 and 7.4 GeV protons

    Small samples of approximately 1 g of 129I and 237Np, two long-lived radioactive waste nuclides, were exposed to spallation neutron fluences from relatively small metal targets of lead or uranium, surrounded with a paraffin moderator 6 cm thick irradiated with 1.5, 3.7 and 7.4 GeV protons. The (n, γ) transmutation rates have been determined for the two radioactive waste nuclides. Conventional radiochemical La and U sensors and a variety of solid-state nuclear track detectors were irradiated simultaneously with secondary neutrons. The observed secondary neutron fluences appear to be systematically larger, as compared to the calculations with the well-known cascade codes (LAHET from Los Alamos and DCM-CEM from Dubna)

  14. Study of the mass, nuclear charge and kinetic energy distribution of the fission fragments produced in the reaction 237 Np (2n th, f)

    In this work, we report fission fragment mass, energy and charge distributions measured for the fissioning nucleus: 239 Np 146, This odd Z nucleus is formed after double thermal neutron capture on to the 237 Np 144 target nucleus. These measurements were performed at the I.L.L. recoil mass spectrometer ''Lohengrin'' in Grenoble. The fission fragments were registered by an ionisation chamber placed at the focal plane of the spectrometer. The obtained distributions are compared to the 240 Pu 146 fragment mass, energy and charge distributions. They are discussed within the Wilkins' scission-point model. Cold fission has been studied while selecting fragmentations with final kinetic energies close to the maximum energy released in the reaction. These cold fission events are discussed according to a calculation based on the Wilkins' scission-point model extrapolated to the cold fragmentation case. 51 refs

  15. Neutron induced fission cross section ratios for 232Th, /sup 235,238/U, 237Np, and 239Pu from 1 to 400 MeV

    Time-of-flight measurements of neutron induced fission cross section ratios for 232Th, /sup 235,238/U, 237Np, and 239Pu, were performed using the WNR high intensity spallation neutron source located at Los Alamos National Laboratory. A multiple-plate gas ionization chamber located at a 20-m flight path was used to simultaneously measure the fission rate for all samples over the energy range from 1 to 400 MeV. Because the measurements were made with nearly identical neutron fluxes, we were able to cancel many systematic uncertainties present in previous measurements. This allows us to resolve discrepancies among different data sets. In addition, these are the first neutron-induced fission cross section values for most of the nuclei at energies above 30 MeV. 8 refs., 3 figs

  16. Neutron-induced fission cross section of 237Np in the keV to MeV range at the CERN n_TOF facility

    Diakaki, M.; Karadimos, D.; Vlastou, R.; Kokkoris, M.; Demetriou, P.; Skordis, E.; Tsinganis, A.; Abbondanno, U.; Aerts, G.; Álvarez, H.; Alvarez-Velarde, F.; Andriamonje, S.; Andrzejewski, J.; Assimakopoulos, P.; Audouin, L.; Badurek, G.; Baumann, P.; Bečvář, F.; Berthoumieux, E.; Calviani, M.; Calviño, F.; Cano-Ott, D.; Capote, R.; Carrillo de Albornoz, A.; Cennini, P.; Chepel, V.; Chiaveri, E.; Colonna, N.; Cortes, G.; Couture, A.; Cox, J.; David, S.; Dolfini, R.; Domingo-Pardo, C.; Dorochenko, A.; Dridi, W.; Duran, I.; Eleftheriadis, Ch.; Embid-Segura, M.; Ferrant, L.; Ferrari, A.; Ferreira-Marques, R.; Fitzpatrick, L.; Frais-Koelbl, H.; Fuji, K.; Furman, W.; Goncalves, I.; Gallino, R.; Gonzalez-Romero, E.; Goverdovski, A.; Gramegna, F.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Haas, B.; Haight, R.; Heil, M.; Herrera-Martinez, A.; Igashira, M.; Ioannidis, K.; Isaev, S.; Jericha, E.; Kadi, Y.; Käppeler, F.; Karamanis, D.; Kerveno, M.; Ketlerov, V.; Koehler, P.; Kolokolov, D.; Konovalov, V.; Krtička, M.; Lamboudis, C.; Leeb, H.; Lindote, A.; Lopes, I.; Lozano, M.; Lukic, S.; Marganiec, J.; Marques, L.; Marrone, S.; Massimi, C.; Mastinu, P.; Mengoni, A.; Milazzo, P. M.; Moreau, C.; Mosconi, M.; Neves, F.; Oberhummer, H.; O'Brien, S.; Oshima, M.; Pancin, J.; Papadopoulos, C.; Paradela, C.; Patronis, N.; Pavlik, A.; Pavlopoulos, P.; Perrot, L.; Plag, R.; Plompen, A.; Plukis, A.; Poch, A.; Pretel, C.; Quesada, J.; Rauscher, T.; Reifarth, R.; Rosetti, M.; Rubbia, C.; Rudolf, G.; Rullhusen, P.; Salgado, J.; Sarchiapone, L.; Savvidis, I.; Sedysheva, M.; Stamoulis, K.; Stephan, C.; Tagliente, G.; Tain, J. L.; Tassan-Got, L.; Tavora, L.; Terlizzi, R.; Vannini, G.; Vaz, P.; Ventura, A.; Villamarin, D.; Vincente, M. C.; Vlachoudis, V.; Voss, F.; Wendler, H.; Wiescher, M.; Wisshak, K.; n TOF Collaboration

    2016-03-01

    The neutron-induced fission cross section of 237Np was experimentally determined at the high-resolution and high-intensity facility n_TOF, at CERN, in the energy range 100 keV to 9 MeV, using the 235U(n ,f ) and 238U(n ,f ) cross section standards below and above 2 MeV, respectively. A fast ionization chamber was used in order to detect the fission fragments from the reactions and the targets were characterized as far as their mass and homogeneity are concerned by means of α spectroscopy and Rutherford backscattering spectroscopy respectively. Theoretical calculations within the Hauser-Feshbach formalism have been performed, employing the empire code, and the model parameters were tuned in order to successfully reproduce the experimental fission cross-sectional data and simultaneously all the competing reaction channels.

  17. Delayed neutron and delayed photon characteristics from photofission of 232Th, 235,238U, and 237Np with endpoint Bremsstrahlung photons in the giant dipole resonance region

    Doré, D.; Dighe, P. M.; Berthoumieux, E.; Laborie, J.-M.; Ledoux, X.; Macary, V.; Panebianco, S.; Ridikas, D.

    2009-10-01

    A renewed interest in photonuclear reactions was stimulated by applications as radioactive ion beam production, irradiation stations by high energy photons, shielding of electron accelerators, etc. Today, a particular attention is paid to the non-destructive characterization of waste barrels and the detection of nuclear materials, both based on photofission process and the associated delayed neutron (DN) and delayed photon (DP) emissions. The need of accurate and complete data for DN and DP yields and time characteristics of actinides was the motivation for an experimental campaign, started in 2004. In this paper, the experimental setup and the data analysis method will be presented and the modeling work will be described. Experimental results for DN and DP characteristics will be compared to calculations in the case of photofission of 232Th, 235,238U, and 237Np.

  18. Transmutation of 129I, 237Np, 238Pu, 239Pu, and 241Am using neutrons produced in target-blanket system `Energy plus Transmutation' by relativistic protons

    J Adam; K Katovsky; A Balabekyan; V G Kalinnikov; M I Krivopustov; H Kumawat; A A Solnyshkin; V I Stegailov; S G Stetsenko; V M Tsoupko-Sitnikov; W Westmeier

    2007-02-01

    Target-blanket facility `Energy + Transmutation' was irradiated by proton beam extracted from the Nuclotron Accelerator in Laboratory of High Energies of Joint Institute for Nuclear Research in Dubna, Russia. Neutrons generated by the spallation reactions of 0.7, 1.0, 1.5 and 2 GeV protons and lead target interact with subcritical uranium blanket. In the neutron field outside the blanket, radioactive iodine, neptunium, plutonium and americium samples were irradiated and transmutation reaction yields (residual nuclei production yields) have been determined using -spectroscopy. Neutron field's energy distribution has also been studied using a set of threshold detectors. Results of transmutation studies of 129I, 237Np, 238Pu, 239Pu and 241Am are presented.

  19. Application of digital signal-processing technique to delayed-neutron yield measurements on thermal-neutron induced fission of 237Np

    The measurement procedure based on the continuous thermal-neutron beam modulation with a mechanical chopper was developed for delayed-neutron yield measurement of the thermal-neutron induced fission of 237Np. The idea of the procedure is similar to that which is widely used in modern communications for the nonauthorized data access prevention. The data is modulated with predefined pattern before transmission to the public network and only the recipient that has the modulation pattern is able to demodulate it upon reception. For the thermal-neutron induced reaction applications, the thermal-neutron beam modulation pattern was used to demodulate the measured delayed-neutron intensity signals on the detector output resulting in nonzero output only for the detector signals correlated with the beam modulation pattern. A comparison of the method with the conventional measurement procedure was provided, and it was demonstrated that the cross-correlation procedure has special features making it superior over the conventional one when the measured value difference from the background is extremely small. Due to strong sensitivity of measurement procedure on the modulation pattern of the neutron beam, one can implement the modulation pattern of specific shape to separate the effect of the thermal part of the beam from the higher energy one in the most confident way in a particular experiment

  20. Experimental study of the neutron induced fission cross-section of 234U, 237Np and 243Am with time-of-flight spectrometry technics

    The current work concentrates on the measurements of the nuclear data needed for solving the problem of transmutation of radiotoxic waste. It includes fission cross-section of 234U in the energy range from thermal to 1 MeV, 237Np and 243Am in the resonance region and average group capture cross-section of 234U and 236U. Almost all of these data are recommended by IAEA as a first priority needs for transmutation problem. Another objective is to obtain the high resolution data of the fission cross-section of the 234U on the fission barrier to confirm the existence of the fine structure, attributed to the vibrational resonances in the third well of the fission barrier. The Dissertation summarizes more, than 10 years of the experiments, performed on the pulsed neutron sources IBR-2 and IBR-30 of Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear Research (FLNP JINR) in Dubna, Russia; 'Fakel' of Russian Scientific Center 'Kurchatov's Institute' and n-TOF of CERN. The TOF technique was used for neutron energy spectrometry and various kinds of detectors to mark the fission events. The independent measurements of the same isotopes done on different neutron sources and sometimes with different techniques gives strong, self-consistent set of data. (author)

  1. Sequential leaching extraction of 239,240Pu, 238Pu, 241Pu, 237Np and 241Am from a mud sample: An intercomparison study

    The transuranics content of a mud sample taken from a nuclear waste storage container was analysed employing two different sequential extraction methods. The following fractions were isolated: (1)Water soluble (2) Readily available (3) Carbonate bound and specifically adsorbed (4) Organically bound (5) Oxide and hydroxides bound and (6) residual. Both methods differ in the reagents employed, the extraction sequence applied as well as the temperature and means of extraction. The 239,240Pu, 238Pu, 237Np and 241Am extracted in each phase were determined using standard radiochemical procedures. 241Pu was analysed through the 241Am in-growth on just one old disk of the residual fraction containing plutonium. Plutonium was mainly associated to organic-oxides fractions (89-92 %). The percentage extracted in each fraction depended on the method and the extraction sequence used. The soluble fraction of plutonium was less than 13%. Neptunium seemed to be the more soluble than the other transuranics (27%) and the americium showed a tendency to be associated to carbonates (30%). (author)

  2. Excited levels of 238Np from spectroscopic measurements of the 237Np(n,γ)238Np reaction and /sup 242m/Am alpha decay

    The gamma rays and conversion electrons emitted following neutron capture in a 237Np target have been measured by use of the GAMS and BILL spectrometers at Grenoble. Gamma ray and alpha particle measurements of /sup 242m/Am alpha decay (Ge(Li)γ singles, γ-γ coincidences, α singles) have been made at Livermore. The data from these measurements have been combined with earlier measurements (Ionescu 1979, Asaro 1964) to produce a more detailed level scheme for 238Np. Approximately 36 levels have been identified from all of the experimental evidence. The experimentally-observed bandhead energies can be compared with predicted values derived from a simple linear addition of excitation energies observed in neighboring odd-mass nuclei. Values for the Gallagher-Moszkowski splitting of each configurational pair were obtained from theoretical calculations (Piepenbring 1978). We have assigned configurations to ten rotational bands whose bandhead energies range from 0 to 342 keV and which represent all but one of the configurations predicted to occur below 385 keV

  3. Capture and Fission rate of 232-Th, 238-U, 237-Np and 239-Pu from spallation neutrons in a huge block of lead.

    Vlachoudis, Vasilis

    2000-01-01

    The study is centered on the research of the incineration possibility of nuclear waste, by the association of a particle accelerator with a multiplying medium of neutrons, in the project "Energy Amplifier" of C. Rubbia. It consists of the experimental determination of the rates of capture and fission of certain elements (232-Th, 238-U, 237-Np and 239-Pu) subjected to a fluence of fast spallation neutrons. These neutrons are produced by the interaction of high kinetic energy protons (several GeV) provided by the CERN-PS accelerator, on a large lead solid volume. The measurement techniques used in this work, are based on the activation of elements in the lead volume and the subsequent gamma spectroscopy of the activated elements, and also by the detection of fission fragment traces. The development, of a Monte Carlo code makes it possible, on one hand, to better understand the relevant processes, and on the other hand, to validate the code, by comparison with measurements, for the design and the construction of...

  4. High resolution measurement of the 231Pa(n,f) cross section from 0.4 eV to 12 MeV

    231Pa neutron induced fission is one of the reactions contributing to stop the generation of unwanted quantities of 232U produced along with 233U in a liquid-metal fast breeder reactor using the 232Th/233U breeding cycle. 232U production is troublesome because of its short lifetime and because its decay chain yields to many α particles and γ rays including a 2.61 MeV penetrating γ radiation causing extra shielding problems. Since 231Pa is not listed in ENDF/B-IV and only scarce data of the 231Pa(n,f) cross section, sigmasub(f), are available from ENDF/B-V, a high resolution measurement of sigmasub(f) is mostly desirable to estimate properly the production rate of 232U. In addition to this reactor physics preoccupation, the opportunity was taken to study some fundamental aspects of the fission process, mainly to shed a light on the existence of a shallow third minimum in the fission barrier of 232Pa as calculated by Moeller and Nix

  5. Nuclear data evaluation for 237Np, 241Am, 242gAm and 242mAm irradiated by neutrons and protons at energies up to 250 MeV

    Evaluation of nuclear data has been performed for 237Np, 241Am, 242gAm and 242mAm. Neutron data were obtained at energies from 20 to 250 MeV and combined with JENDL-3.3 data at 20 MeV. Evaluation of the proton data has been done from 1 to 250 MeV. The coupled channel optical model was used to obtain angular distributions for elastic and inelastic scattering and transmission coefficients. Pre-equilibrium exciton model and Hauser-Feshbach statistical model were used to describe neutron and charged particles emission from excited nuclei. These evaluation is the first work for producing full sets of evaluated file up to 250 MeV for 237Np and Americium isotopes. (author)

  6. Standard practice for the determination of 237Np, 232Th, 235U and 238U in urine by inductively coupled plasma-Mass spectrometry (ICP-MS) and gamma ray spectrometry.

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This practice covers the separation and preconcentration of neptunium-237 (237Np), thorium-232 (232Th), uranium-235 (235U) and uranium-238 (238U) from urine followed by quantitation using ICP-MS. 1.2 This practice can be used to support routine bioassay programs. The minimum detectable concentrations (MDC) for this method, taking the preconcentration factor into account, are approximately 1E-2Bq for 237Np (0.38ng), 2E-6Bq for 232Th (0.50ng), 4E-5Bq for 235U (0.50ng) and 6E-6Bq for 238U (0.48ng). 1.3 This standard does not purport to address all of the safety problems, 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.

  7. Evaluation and improvement of cross section accuracy for most important dosimetry reactions 27Al(n,p), 56Fe(n,p) and 237Np(n,f) including covariance data

    New evaluations of cross sections and their uncertainties for dosimetry reactions 27Al(n,p) , 56Fe(n,p) and 237Np(n,f) have been carried out in the frame work of IAEA Research Contract No. 11372/RB. Data files prepared for this reactions in the ENDF-6 format may be consider as candidates for the new International Reactor Dosimetry File: IRDF-2002. (author)

  8. Redundant 230Th/ 234U/ 238U, 231Pa/ 235U and 14C dating of fossil corals for accurate radiocarbon age calibration

    Chiu, Tzu-Chien; Fairbanks, Richard G.; Mortlock, Richard A.; Cao, Li; Fairbanks, Todd W.; Bloom, Arthur L.

    2006-09-01

    230Th/ 234U/ 238U dating of fossil corals by mass spectrometry is remarkably precise, but some samples exposed to freshwater over thousands of years may gain and/or lose uranium and/or thorium and consequently yield inaccurate ages. Although a δ 234U initial value equivalent to modern seawater and modern corals has been an effective quality control criterion, for samples exposed to freshwater but having δ 234U initial values indistinguishable from modern seawater and modern corals, there remains a need for additional age validation in the most demanding applications such as the 14C calibration (Fairbanks et al., 2005. Radiocarbon calibration curve spanning 0 to 50,000 years BP based on paired 230Th/ 234U/ 238U and 14C dates on pristine corals. Quaternary Science Reviews 24(16-17), 1781-1796). In this paper we enhance screening criteria for fossil corals older than 30,000 years BP in the Fairbanks0805 radiocarbon calibration data set (Fairbanks et al., 2005) by measuring redundant 230Th/ 234U/ 238U and 231Pa/ 235U dates via multi-collector magnetic sector inductively coupled plasma mass spectrometry (MC-MS-ICPMS) using techniques described in Mortlock et al. (2005. 230Th/ 234U/ 238U and 231Pa/ 235U ages from a single fossil coral fragment by multi-collector magnetic-sector inductively coupled plasma mass spectrometry. Geochimica et Cosmochimica Acta 69(3), 649-657.). In our present study, we regard paired 231Pa/ 235U and 230Th/ 234U/ 238U ages concordant when the 231Pa/ 235U age (±2 σ) overlaps with the associated 230Th/ 234U/ 238U age (±2 σ). Out of a representative set of 11 Fairbanks0805 (Fairbanks et al., 2005) radiocarbon calibration coral samples re-measured in this study, nine passed this rigorous check on the accuracy of their 230Th/ 234U/ 238U ages. The concordancy observed between 230Th/ 234U/ 238U and 231Pa/ 235U dates provides convincing evidence to support closed system behavior of these fossil corals and validation of their 230Th/ 234U/ 238U

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

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

    2002-01-01

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

  10. The nuclear structure of 229Pa from the 231Pa(p,t)229Pa and 230Th(p,2nγ)229Pa reactions

    The level structure of the 229Pa nucleus has been investigated by means of the 231Pa(p,t)229Pa and 230Th(p,2nγ)229Pa reactions. Triton angular-distribution measurements were subjected to a CCBA analysis and combined with the results of in-beam conversion electron and γ-ray spectroscopy to establish a level scheme. Two low-lying bands of opposite parity were observed up to spins (23/2)- and (17/2)+, respectively. Rotational bands built on some 0+ excitations of the even-even core can be assigned. The lowest states of three further low-lying bands are observed. The level scheme is interpreted in terms of an octupole-deformed core with an unpaired proton. From the E1/E2 branching ratio the electric dipole moment can be deduced vertical stroke D0vertical stroke =(0.09 ±0.04) e .fm. ((orig.))

  11. Investigation of Neutron Spectra and Transmutation of ^{129}I, ^{237}Np and Other Nuclides with 1.5 GeV Protons from the Dubna Nuclotron Using the Electronuclear Setup "Energy plus Transmutation"

    Krivopustov, M I; Balabekyan, A R; Batusov, Yu A; Bielewicz, M; Brandt, R; Chaloun, P; Chultem, D; Dwivedi, K K; Elishev, A F; Fragopoulou, M; Henzl, V; Henzlová, D; Kalinnikov, V G; Kievets, M K; Krása, A; Krizek, F; Kugler, A; Manolopoulou, Metaxia; Mariin, I I; Nourreddine, A; Odoj, R; Pavliouk, A V; Pronskikh, V S; Robotham, H; Siemon, K; Szuta, M; Stegailov, V I; Solnyshkin, A A; Sosnin, A N; Stoulos, S; Tsoupko-Sitnikov, V M; Tumendelger, T; Wojecehowski, A; Wagner, V; Wan, J S; Westmeier, W; Zamani-Valasiadou, M; Kumawat, H; Kumar, V; Zaverioukha, O S; Zhuk, I V

    2004-01-01

    Experiments which are part of the scientific program "Investigations of physical aspects of electronuclear method of energy production and transmutation for radioactive waste of atomic energetics using relativistic beams from the JINR Synchrophasotron/Nuclotron" (project "Energy plus Transmutation") are described. A large lead target surrounded by a four-section uranium blanket with total weight of 206.4 kg natural uranium was irradiated with 1.5 GeV protons from the new cryogenic accelerator Nuclotron. Radiochemical sensors were exposed to the secondary particle fluences inside and on top of the target assembly. Two long-lived radioactive waste of atomic energetics sensors ^{129}I and ^{237}Np (approximately 1 g weight each) and stable nuclides ^{27}Al, ^{59}Co, ^{127}I, ^{139}La, ^{197}Au and ^{209}Bi as well as natural and enriched uranium were used. In addition, various solid state nuclear track detectors and nuclear emulsions were exposed simultaneously. The experimental results confirm the theoretical e...

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

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

  13. 230Th/ 234U/ 238U and 231Pa/ 235U ages from a single fossil coral fragment by multi-collector magnetic-sector inductively coupled plasma mass spectrometry

    Mortlock, Richard A.; Fairbanks, Richard G.; Chiu, Tzu-chien; Rubenstone, James

    2005-02-01

    The 230Th/ 234U/ 238U age dating of corals via alpha counting or mass spectrometry has significantly contributed to our understanding of sea level, radiocarbon calibration, rates of ocean and climate change, and timing of El Nino, among many applications. Age dating of corals by mass spectrometry is remarkably precise, but many samples exposed to freshwater yield inaccurate ages. The first indication of open-system 230Th/ 234U/ 238U ages is elevated 234U/ 238U initial values, very common in samples older than 100,000 yr. For samples younger than 100,000 yr that have 234U/ 238U initial values close to seawater, there is a need for age validation. Redundant 230Th/ 234U/ 238U and 231Pa/ 235U ages in a single fossil coral fragment are possible by Multi-Collector Magnetic Sector Inductively Coupled Plasma Mass Spectrometry (MC-MS-ICPMS) and standard anion exchange column chemistry, modified to permit the separation of uranium, thorium, and protactinium isotopes from a single solution. A high-efficiency nebulizer employed for sample introduction permits the determination of both 230Th/ 234U/ 238U and 231Pa/ 235U ages in fragments as small as 500 mg. We have obtained excellent agreement between 230Th/ 234U/ 238U and 231Pa/ 235U ages in Barbados corals (30 ka) and suggest that the methods described in this paper can be used to test the 230Th/ 234U/ 238U age accuracy. Separate fractions of U, Th, and Pa are measured by employing a multi-dynamic procedure, whereby 238U is measured on a Faraday cup simultaneously with all minor isotopes measured with a Daly ion counting detector. The multi-dynamic procedure also permits correcting for both the Daly to Faraday gain and for mass discrimination during sample analyses. The analytical precision of 230Th/ 234U/ 238U and 231Pa/ 235U dates is generally better than ±0.3% and ±1.5%, respectively (2 Relative Standard deviation [RSD]). Additional errors resulting from uncertainties in the decay constant for 231Pa and from undetermined

  14. Fusion hybrids for generation of advanced (231Pa+232U+233U+234U)-fuel in closed (U-Pu-Th)-fuel cycle

    Technology of controlled thermonuclear fusion (CTF) is traditionally regarded as a practically inexhaustible energy source. However, development, mastering, broad deployment of fast breeder reactors and closure of nuclear fuel cycle (NFC) can also extend fuel base of nuclear power industry (NPI) up to practically unlimited scales. Under these conditions, it seems reasonable to introduce into a circle of the CTF-related studies the works directed towards solving some principal problems which can appear in a large-scale NPI in closed NFC. The first challenge is a large scale of operations in NFC back-end that should be reduced by achieving substantially higher fuel burn-up in power nuclear reactors. The use of 231Pa-232Th-232U-233U fuel in light-water reactor (LWR) opens a possibility of principle to reach very high (about 30% HM) or even ultra-high fuel burn-up. The second challenge is a potential unauthorized proliferation of fissionable materials. As is known, a certain remarkable quantity of 232U being introduced into uranium fraction of nuclear fuel can produce a serious barrier against switching the fuel over to non-energy purposes. Involvement of hybrid thermonuclear reactors (HTR) into NPI structure can substantially facilitate resolving these problems. If HTR will be involved into NPI structure, then main HTR mission consists not in energy generation but in production of nuclear fuel with a certain isotope composition. The present paper analyzes some neutron-physical features in production of advanced nuclear fuels in thorium HTR blankets. The obtained results demonstrated that such a nuclear fuel may be characterized by very stable neutron-multiplying properties during full LWR operation cycle and by enhanced proliferation resistance too. The paper evaluates potential benefits from involvement of HTR with thorium blanket into the international closed NFC. (author)

  15. 237Np在破碎凝灰岩和凝灰质砂上的吸附研究%STUDY OF 237Np ADSORPTION ON CRUSHED TUFF AND TUFFACEOUS SAND

    王旭东; 田中忠夫; 武部慎一

    2001-01-01

    The adsorption characteristics of 237Np on crushed tuff andtuffaceous sand from northeast Japan have been studied using sequential extraction analysis technique. With a large difference in cation exchange capacity (CEC), both materials show nearly same sorptivity for 237Np. Together with dominance of ion exchange processes in sorption indicated by sequential extraction analysis, it suggests an absence of specific sorption at sites of this kind. It's interesting to find that sorption on Fe-Mn oxyhydroxide-oxides, defined by our sequential extraction procedure, almost keep constant for both cases, without time dependency. A potential mechanism of surface chemical reaction is deduced. Kinetics with respect to exchangeable and residual portions in tuff and fuffaceous sand confirms that some slow processes are controlling radionuclide sorption.%用连续提取法研究了237Np在破碎凝灰岩和凝灰质砂上的吸附。研究结果表明,48h以后,2种介质对237Np的吸附百分数相近,而且吸附的237Np在各相间的分布均以离子交换态为主。凝灰岩和凝灰质砂的阳离子交换容量差异显著,说明离子交换作用仅发生在粘土矿物的表面,凝灰岩中蒙脱石的层间没有发生阳离子交换。在28d的实验时间内,237Np在介质中的Fe-Mn氧化物-氢氧化物上的吸附百分数几乎不随时间变化。这可能是因为该吸附过程为瞬时即可完成的表面化学反应。介质内离子交换相和残余相上的吸附则显示为缓慢的吸附过程。

  16. Fission Fragment Folding Angle Distributions for the Systems 11B+237Np, 12C+236U, and 16O+232Th in the Energy Range 1.1B<2.1

    Fission fragment folding angle distributions have been measured for the systems 11B+237Np, 12C+236U, and 16O+232Th, populating the same compound nucleus (248Cf) and at similar excitation energies (Ex 45-100 MeV). The full momentum transfer and incomplete momentum transfer fusion-fission components have been separated over the bombarding energy range 1.1c.m/VB 2.1. It is observed that the largest value of the ratio of the transfer fission to the total fission is around 10 to 15% at the highest energy investigated. Over the energy range mentioned above, it is found that the transfer fission corrected fission fragment anisotropies are not significantly different from the values already obtained from the analysis of the total fission data reported earlier and hence the conclusions reached from the inclusive data remain unchanged. The anisotropy data were analyzed for the two cases corresponding to fission events with sizable fission barriers (Bf>T) and with smaller fission barriers (Bf>T). It was interesting to find that the effective moment of inertia (Jeff) values deduced from the latter component were consistent with the values from Sierk prescription used in the former case

  17. About the first experiment on investigation of 129I, 237Np, 238Pu and 239Pu transmutation at the nuclotron 2.52 GeV deuteron beam in neutron field generated in U/Pb-assembly 'Energy plus transmutation'

    Preliminary results of the first experiment with energy 2.52 GeV at the electronuclear setup which consists of Pb-target (diameter 8.4 cm, length 45.6 cm) and natU-blanket (206.4 kg), transmutation samples of 129I, 237Np, 238Pu and 239Pu (radioecological aspect) are described. Hermetically sealed samples in notable amounts are gathered in atomic reactors and setups of industries which use nuclear materials and nuclear technologies were irradiated in the field of neutrons produced in the Pb-target and propagated in the natU-blanket. Estimates of transmutations were obtained as a result of measurements of gamma activities of the samples. The information about the space and energy distribution of neutrons in the volume of the lead target and the uranium blanket was obtained with the help of sets of activation threshold detectors (Al, Co, Y, I, Au, Bi and others), solid-state nuclear track detectors, 3He neutron detectors and nuclear emulsion. Comparison of the experimental data with the results of simulation with the MCNPX program was performed

  18. Investigation of neutron spectra and transmutation of 129I, 237Np and other nuclides with 1.5 GeV protons from the Dubna Nuclotron using the electronuclear setup 'Energy plus Transmutation'

    Experiments which are part of the scientific program 'Investigations of physical aspects of electronuclear method of energy production and transmutation for radioactive waste of atomic energetics using relativistic beams from the JINR Synchrophasotron/Nuclotron' (project 'Energy plus Transmutation') are described. A large lead target surrounded by a four-section uranium blanket with a total weight of 206.4 kg natural uranium was irradiated with 1.5 GeV protons from the new cryogenic accelerator Nuclotron. Radiochemical sensors were exposed to the secondary particle fluences inside and on top of the target assembly. Two long-lived radioactive waste of atomic energetics sensors 129I and 237Np (approximately 1 g weight each) and stable nuclides 27Al, 59Co, 127I, 139La, 197Au and 209Bi as well as natural and enriched uranium were used. In addition, various solid state nuclear track detectors and nuclear emulsions were exposed simultaneously. The experimental results confirm the theoretical estimations that the neutron spectra around the U/Pb-assembly are dominated by medium- and high-energy neutrons as shown by the observation of (n, xn)-reaction products in Co, Au and Bi sensors. The yield of thermal neutrons on the surface of the U-blanket is strongly reduced as compared to the surface of a smaller Pb target surrounded with paraffin. The latter data were determined with (n, γ) reactions in stable La sensors. In this experiment the technique of nuclear emulsions has been applied for the first time to measurements of neutron spectra in an accelerator driven system

  19. Neutron-induced fission cross-section of 231Pa

    A first series of fission cross-section measurements for incident neutron energies between 0.6 and 3.4 MeV has confirmed a first chance threshold value around 1b. In contrast to our findings for the fission cross-section in 233Pa, both the direct and the surrogate cross-section data lead to the same result. This seems to support the assumption, that only in cases, where ingoing and outgoing particle are similar, particle-transfer reactions give results that are in agreement with those obtained from direct compound nuclear reactions

  20. Speciation of 241Am molecular compounds through 237Np Moessbauer and 241Am XPS spectroscopy

    Light actinides (U to Am) can be found in several oxidation states from (II) to (VII) in the molecular form or in the condensed matter state. The large variety of oxidation states is usually attributed to the contribution of the 5f states to the valence orbitals. For the heavier actinides, for which the 5f electrons are non bonding, the actinides become rare-earth like with a smaller number of oxidation states (II and III). However it is still not understood what really decides on the stability of a given oxidation state, and how it is depending on the chemical environment (coordination sphere, nature of the counter-anion, etc). This work shows how Moessbauer spectroscopy and 4f photoelectron spectroscopy (XPS) can contribute to progress in the understanding of the electronic structure of the actinide, especially for Am compounds Moessbauer reverse experiments were undertaken to show in what manner the electronic structure of the Am is preserved during the decay process (oxidation state stability). The result of XPS measurements shows that it is possible to correlate the 4f binding energy of the Am to the charge at the actinide core. The obtained results are somewhat surprising. The formal oxidation state (V) is 'less oxidised' than expected. Some Am(III) have less electron density (that means are more ionic) than americyl (V) hydroxide or carbonate. The reason for these surprisingly results comes from the 'Am=O' multiple bond system which reduces dramatically the charge at the actinide by a pi-donation mechanism. The evolution of the 4f binding energy of the Am species does not follow the oxidation state order. Theoretical DFT calculation were done on Am(V) compounds for qualitative electronic modeling. (authors)

  1. Post-Irradiation Examination of 237Np Targets for 238Pu Production

    Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Hobbs, Randy W [ORNL; Schmidlin, Joshua E [ORNL

    2015-01-01

    Oak Ridge National Laboratory is recovering the US 238Pu production capability and the first step in the process has been to evaluate the performance of a 237Np target cermet pellet encased in an aluminum clad. The process proceeded in 3 steps; the first step was to irradiate capsules of single pellets composed of NpO2 and aluminum power to examine their shrinkage and gas release. These pellets were formed by compressing sintered NpO2 and aluminum powder in a die at high pressure followed by sintering in a vacuum furnace. Three temperatures were chosen for sintering the solution precipitated NpO2 power used for pellet fabrication. The second step was to irradiate partial targets composed of 8 pellets in a semi-prototypical arrangement at the two best performing sintering temperatures to determine which temperature gave a pellet that performed the best under the actual planned irradiation conditions. The third step was to irradiate ~50 pellets in an actual target configuration at design irradiation conditions to assess pellet shrinkage and gas release, target heat transfer, and dimensional stability. The higher sintering temperature appeared to offer the best performance after one cycle of irradiation by having the least shrinkage, thus keeping the heat transfer gap between the pellets and clad small minimizing the pellet operating temperature. The final result of the testing was a target that can meet the initial production goals, satisfy the reactor safety requirements, and can be fabricated in production quantities. The current focus of the program is to verify that the target can be remotely dissembled, the pellets dissolved, and the 238Pu recovered. Tests are being conducted to examine these concerns and to compare results to code predictions. Once the performance of the full length targets has been quantified, the pellet 237Np loading will be revisited to determine if it can be increased to increase 238Pu production.

  2. Sorption of 231Pa on silica and the effect of humic acid

    Sorption of protactinium on silica colloids was studied in the pH range of 1 to 12 in NaClO4 medium using radiotracer technique. Silica was characterized using X-ray diffraction, light scattering and surface area measurements. The point of zero charge for silica colloids was about pH 2. The sorption of protactinium was about 98% in the pH range of 3 to 9 and was lower (70-80 %) below pH 3 and above pH 9. The quantitative sorption in the pH range 3 to 9 could be explained by surface complexation model. The reduction in sorption was attributed to electrostatic repulsion as the fraction of protactinium exists as cationic species at pH ≤ 2 and anionic species above pH 9. There was reduction in the sorption of protactinium in the presence of humic acid below pH 2 and above pH 10. Sorption of protactinium on silica in presence of 0.05M HF was about 99% between pH 3 to 8 and below 30% in the low and high pH region. Isotherm study revealed an exponential decrease in protactinium activity in solution with increase in silica. (author)

  3. Speciation of {sup 241}Am molecular compounds through {sup 237}Np Moessbauer and {sup 241}Am XPS spectroscopy

    Fouchard, S.; Gouder, T.; Colineau, E.; Wastin, F.; Rebizant, J.; Simoni, E.; Guillaumont, D.; Meyer, D

    2004-07-01

    Light actinides (U to Am) can be found in several oxidation states from (II) to (VII) in the molecular form or in the condensed matter state. The large variety of oxidation states is usually attributed to the contribution of the 5f states to the valence orbitals. For the heavier actinides, for which the 5f electrons are non bonding, the actinides become rare-earth like with a smaller number of oxidation states (II and III). However it is still not understood what really decides on the stability of a given oxidation state, and how it is depending on the chemical environment (coordination sphere, nature of the counter-anion, etc). This work shows how Moessbauer spectroscopy and 4f photoelectron spectroscopy (XPS) can contribute to progress in the understanding of the electronic structure of the actinide, especially for Am compounds Moessbauer reverse experiments were undertaken to show in what manner the electronic structure of the Am is preserved during the decay process (oxidation state stability). The result of XPS measurements shows that it is possible to correlate the 4f binding energy of the Am to the charge at the actinide core. The obtained results are somewhat surprising. The formal oxidation state (V) is 'less oxidised' than expected. Some Am(III) have less electron density (that means are more ionic) than americyl (V) hydroxide or carbonate. The reason for these surprisingly results comes from the 'Am=O' multiple bond system which reduces dramatically the charge at the actinide by a pi-donation mechanism. The evolution of the 4f binding energy of the Am species does not follow the oxidation state order. Theoretical DFT calculation were done on Am(V) compounds for qualitative electronic modeling. (authors)

  4. Migration of the radionuclides 99Tc, 237Np, 238Pu and 241Am in the caprock of the Gorleben repository

    31 different combinations of sediments with groundwaters (sediment-groundwater-systems) formed the basis of the experiments to determine sorption data and sorption mechanisms of radionuclides in selected acquifer systems with loose rock samples taken from shafts 1 and 2. The waters were selected on the basis of pore water analyses. For cohesive materials from which no pore water could be extracted, groundwaters from the horizon below and above were chosen. (orig.)

  5. Transmutation study of 237Np in energy + transmutation setup using 1.6 GeV deuteron-beam

    In this paper, the results of the measurement of cross sections of (n, γ) and (n, f) reactions of long lived fission fragment, Np237 have been presented in the mixed neutron environment of the E + T set up

  6. Measurements of periods, relative abundances and absolute yields of delayed neutrons from fast neutron induced fission of {sup 237}Np

    Piksaikine, V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-03-01

    The experimental method for measurements of the delayed neutron yields and period is presented. The preliminary results of the total yield, relative abundances and periods are shown comparing with the previously reported values. (J.P.N.)

  7. Measurement of fission cross sections and fragment angular distributions using solid state track detectors

    Fission cross sections and angular distributions of fission fragments from fissions induced by 14.1 and 15.8 MeV neutrons, respectively, in 232Th, 231Pa, 233U, 235U, 238U, 237Np, 239Pu and 241Am have been studied using Lexan plastic track detectors. A novel experimental set-up evolved from considerations of neutron economy allows simultaneous measurement of angular distribution of fission fragments from five independently fissioning nuclides at a time. The data on angular anistropy were analysed in the perspective of different chances of fissions taking place simultaneously in this energy region. Third-chance fission thresholds for 231Pa and 241Am were estimated from the measured anisotropy values to be 13.2 and 11.1 MeV, respectively. (author)

  8. Radiochemical method for the simultaneous determination of 233U, 236U, 237Np, 236Pu, 238Pu, and 239Pu in biological materials

    A radiochemical method has been developed for the determination of multiple isotopes of uranium, neptunium, and plutonium in biological materials. The elements are separated from the other sample constituents and from each other by anion exchange in halide media. Their recoveries are monitored by isotopic diluents. The amounts of the analyte and diluent isotopes of each element are measured alpha spectrometrically. The interelemental separation factors are generally greater than 102, and the recovery of each element ranges from 60% to 90%. 4 references, 1 table

  9. Transmutation of 129I,237 Np, 239Pu, and 241Am using neutrons produced in target-blanket system "Energy plus Transmutation" by relativistic protons

    Adam, Jindřich

    2007-01-01

    Roč. 68, č. 2 (2007), s. 201-212. ISSN 0304-4289 R&D Projects: GA MŠk 1P04LA213 Institutional research plan: CEZ:AV0Z10480505 Keywords : neutro production Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.383, year: 2007

  10. High-temperature X-ray diffraction studies on La1-2xCaxThxPO4(s) (x = 0, 0.25, 0.5) solid-solution

    Advanced Heavy Water Reactor (AHWR), Compact High Temperature Reactor (CHTR) and Accelerator Driven System (ADS) are being developed in India to use 232Th-233U fuel cycle. Unlike natural uranium, natural thorium contains only trace amounts of fissile material (231Th), which are insufficient to initiate a nuclear chain reaction. It can be used with 233U, 235U and 239Pu as fissile fuel. Because of it, the back end of thorium fuel cycle contains long-lived 231Pa, 229Th, 230Th and low level of minor actinides (237Np, 241Am, 243Am, 244Cm) for final disposal in geological repositories. From a geochemical point of view, monazite (LnPO4 with Ln: rare earths) is the most abundant lanthanide phosphate observed in natural samples). Such minerals appear as the major thorium source on earth, especially in several ores which contain up to 29 wt% of ThO2, 16wt% of UO2, respectively

  11. First results studying the transmutation of 129I, 237Np, 238Pu, and 239Pu in the irradiation of an extended natU/Pb-assembly with 2.52 GeV deuterons

    Krivopustov, M. I.; Pavljuk, A. V.; Kovalenko, A.D.; Mariin, I.I.; Elishev, A.F.; Adam, Jindřich; Kovalík, Alojz; Batusov, Yu, A.; Kalinnikov, V. G.; Brudanin, V. B.; Čaloun, Pavel; Tsoupko-Sitnikov, V. M.; Solnyshkin, A. A.; Stegailov, V. I.; Gerbish, Sh.; Svoboda, Ondřej; Dubnická, Z.; Kala, M.; Kloc, M.; Krása, Antonín; Kugler, Andrej; Majerle, Mitja; Wagner, Vladimír; Brandt, R.; Westmeier, W.; Robotham, H.; Siemon, K.; Bielewicz, M.; Kilim, S.; Szuta, M.; Strugalska-Gola, E.; Wojeciechowski, A.; Hashemi-Nezhad, R. S.; Manolopoulou, M.; Fragopolou, M.; Stoulos, S.; Zamani-Valasiadou, M.; Jokic, S.; Katovsky, K.; Schastny, O.; Zhuk, I. V.; Potapenko, A.S.; Safronova, A.A.; Lukashevich, Zh.A.; Voronko, V.A.; Sotnikov, V.V.; Sidorenko, V.V.; Ensinger, W.; Severin, H.D.; Batsev, S.; Kostov, L.; Protokhristov, Kh.; Stoyanov, Ch,.; Yordanov, O.; Zhivkov, P.K.; Kumar, A.V.; Sharma, M.; Khilmanovich, A.M.; Marcinkevich, B.A.; Korneev, S.V.; Damdinsuren, Ts.; Togoo, Ts.; Kumawat, H.

    2009-01-01

    Roč. 279, č. 2 (2009), s. 567-584. ISSN 0236-5731 R&D Projects: GA AV ČR IAA100480803; GA MŠk LC07050 Institutional research plan: CEZ:AV0Z10480505 Keywords : MCNPX. * GeV Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.631, year: 2009

  12. Procedures for determination of 239,240Pu, 241Am, 237Np, 234,238U, 228,230,232Th, 99Tc and 210Pb-210Po in environmental material

    Since 1987, the Department of Nuclear Safety Research, Risoe National Laboratory has developed procedures for analysis of low-level amounts of radioactivity in large samples of 200 liters seawater, 10 gram sediment, soil and other environmental materials. These analytical procedures provide high chemical yields, good resolution and excellent decontamination factors for large environmental samples analysed by alpha spectrometry and mass spectrometry (ICPMS). The procedures have been checked through practical analysis work and are used in Norway, the Netherlands, Germany, Spain, France and Denmark. (au)

  13. Design study of long-life PWR using thorium cycle

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-01

    Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that 231Pa better than 237Np as burnable poisons in thorium fuel system. Thorium oxide system with 8% 233U enrichment and 7.6˜ 8% 231Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1% Δk/k and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53% Δk/k and reduced power peaking during its operation.

  14. A study of the generation of 232U in UO2 and MOX fuels

    To clarify the generation pathway of 232U, an important nuclide for dose evaluation at various stages in the reuse of uranium, concentrations of 232U generated through various pathways were evaluated for UO2 and mixed oxide (MOX) fuels. Burnup calculation was conducted with ORIGEN2.2 code adopting ORLIBJ40 library, a set of cross-section libraries based on JENDL-4.0. It was found that differences in 232U concentrations in UO2 and MOX fuels mainly arise from differences in the initial compositions of 234U, 235U, and 236U. It was also found that the contribution of plutonium and americium isotopes in MOX fuels is small compared with that of uranium isotopes. The results clarified that the capture cross sections of 230Th, 231Pa, 235U, and 236U, as well as the (n,2n) cross sections of 237Np and 238U, have a large effect on the generation of 232U. Additional investigation showed that 232U concentration is strongly affected not only by time after irradiation but also by time before irradiation. (author)

  15. Characterization of actinide physics specimens for the US/UK joint experiment in the Dounreay Prototype Fast Reactor

    The United States and the United Kingdom are engaged in a joint research program in which samples of the higher actinides are irradiated in the Dounreay Prototype Fast Reactor in Scotland. The purpose of the porogram is (1) to study the materials behavior of selected higher actinide fuels and (2) to determine the integral cross sections of a wide variety of the higher actinide isotopes. Samples of the actinides are incorporated in fuel pins inserted in the core. For the fuel study, the actinides selected are 241Am and 244Cm in the form of Am2O3, Cm2O3, and Am6Cm(RE)7O21, where (RE) represents a mixture of lanthanides. For the cross-section determinations, the samples are milligram quantities of actinide oxides of 248Cm, 246Cm, 244Cm, 243Cm, 243Am, 241Am, 244Pu, 242Pu, 241Pu, 240Pu, 239Pu, 238Pu, 237Np, 238U, 236U, 235U, 234U, 233U, 232Th, 230Th, and 231Pa encapsulated in vanadium. Coincident with the irradiations, neutron flux and energy spectral measurements are made with vanadium-encapsulated dosimeter materials located within the same fuel pins

  16. Preparation of actinide specimens for the US/UK joint experiment in the Dounreay Prototype Fast Reactor

    A joint research program involving the United States and the United Kingdom was initiated about four years ago for the purpose of studying the fuel behavior of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of integral cross sections of a wide variety of higher actinide isotopes (physics specimens) was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the fuel pellets and physics samples. The higher actinide samples chosen for the fuel study were 241Am and 244Cm in the forms of Am2O3, Cm2O3, and Am6Cm(RE)7O21, where (RE) represents a mixture of lanthanides. Milligram quantities of actinide oxides of 248Cm, 246Cm, 244Cm, 243Cm, 243Am, 241Am, 244Pu, 242Pu, 241Pu, 240Pu, 239Pu, 238Pu, 237Np, 238U, 236U, 235U, 234U, 233U, 232Th, 230Th, and 231Pa were encapsulated to obtain nuclear cross section and reaction rate data for these materials

  17. The use of fission foils for plasma neutron diagnostics

    Commonly used fission foil materials have been examined for their application to plasma diagnostics as activation foils. Such foils have been used extensively in the past for fission reactor dosiemetry. They have very well known fission cross sections, and in most cases the fission yields are reasonably well known. The materials included in this study are 226Ra, 228Th, 232Th, 231Pa, 233U, 235U, 238U, 237Np, 238Pu, and 239Pu. Of these materials 232Th, 235U, and 238U are considered to be very good candidates for this application. The others have been eliminated because of high background radioactivity, impurities which present high backgrounds, or lack of knowledge about yield distribution of fission products. Production cross sections for fission products in the vicinity of the yield maxima (A = 85 - 101, 133 143) have been calculated from known fission cross sections and independent or cumulative yields at thermal energies (where applicable) and 14 MeV. Recent measurements at 2.5 MeV are also included. For one foil (232Th) results for 3 MeV and 11 MeV are also available. The decay schemes of the more prominent fission products have been thoroughly studied and good measurement precision should result from their use

  18. Application of the Spanish methodological approach for biosphere assessment to a generic high-level waste disposal site.

    Agüero, A; Pinedo, P; Simón, I; Cancio, D; Moraleda, M; Trueba, C; Pérez-Sánchez, D

    2008-09-15

    A methodological approach which includes conceptual developments, methodological aspects and software tools have been developed in the Spanish context, based on the BIOMASS "Reference Biospheres Methodology". The biosphere assessments have to be undertaken with the aim of demonstrating compliance with principles and regulations established to limit the possible radiological impact of radioactive waste disposals on human health and on the environment, and to ensure that future generations will not be exposed to higher radiation levels than those that would be acceptable today. The biosphere in the context of high-level waste disposal is defined as the collection of various radionuclide transfer pathways that may result in releases into the surface environment, transport within and between the biosphere receptors, exposure of humans and biota, and the doses/risks associated with such exposures. The assessments need to take into account the complexity of the biosphere, the nature of the radionuclides released and the long timescales considered. It is also necessary to make assumptions related to the habits and lifestyle of the exposed population, human activities in the long term and possible modifications of the biosphere. A summary on the Spanish methodological approach for biosphere assessment are presented here as well as its application in a Spanish generic case study. A reference scenario has been developed based on current conditions at a site located in Central-West Spain, to indicate the potential impact to the actual population. In addition, environmental change has been considered qualitatively through the use of interaction matrices and transition diagrams. Unit source terms of (36)Cl, (79)Se, (99)Tc, (129)I, (135)Cs, (226)Ra, (231)Pa, (238)U, (237)Np and (239)Pu have been taken. Two exposure groups of infants and adults have been chosen for dose calculations. Results are presented and their robustness is evaluated through the use of uncertainty and

  19. Analysis of trace neptunium in the vicinity of underground nuclear tests at the Nevada National Security Site

    A high sensitivity analytical method for 237Np analysis was developed and applied to groundwater samples from the Nevada National Security Site (NNSS) using short-lived 239Np as a yield tracer and HR magnetic sector ICP-MS. The 237Np concentrations in the vicinity of the Almendro, Cambric, Dalhart, Cheshire, and Chancellor underground nuclear test locations range from <4 × 10−4 to 2.6 mBq/L (6 × 10−17–4.2 × 10−13 mol/L). All measured 237Np concentrations are well below the drinking water maximum contaminant level for alpha emitters identified by the U.S. EPA (560 mBq/L). Nevertheless, 237Np remains an important indicator for radionuclide transport rates at the NNSS. Retardation factor ratios were used to compare the mobility of 237Np to that of other radionuclides. The results suggest that 237Np is less mobile than tritium and other non-sorbing radionuclides (14C, 36Cl, 99Tc and 129I) as expected. Surprisingly, 237Np and plutonium (239,240Pu) retardation factors are very similar. It is possible that Np(IV) exists under mildly reducing groundwater conditions and exhibits a retardation behavior that is comparable to Pu(IV). Independent of the underlying process, 237Np is migrating downgradient from NNSS underground nuclear tests at very low but measureable concentrations. - Highlights: • A high sensitivity analytical method for 237Np analysis in groundwater was developed. • Groundwater samples from the Nevada National Security Site (NNSS) were analyzed. • 237Np concentrations were well below the EPA maximum contaminant level in drinking water. • 237Np is less mobile than 3H and other non-sorbing radionuclides. • 237Np and Pu apparent retardation factors are similar

  20. Np-237 in peat and lichen in Finland

    Salminen, S.; Paatero, J.; Roos, Per;

    2009-01-01

    Activity concentrations of 237Np in peat and lichen samples in Finland were determined and contributions from nuclear weapons testing in 1950–1960s and the Chernobyl accident were estimated. 237Np was determined with ICP-MS using 235Np as a tracer. Activity concentrations of 237Np in peat samples...... varied between 1.98 ± 0.05 and 14.1 ± 0.3 mBq/m2. The contribution from the Chernobyl accident to the total 237Np deposition in peat was 0.1–13%, the Chernobyl-derived fraction of total 237Np in peat being much lower than the previously determined corresponding Chernobyl-derived fractions of 239+240Pu...

  1. Inorganic, radioisotopic and organic analysis of 241-AP-101 tank waste

    Battelle received five samples from Hanford waste tank 241-AP-101, taken at five different depths within the tank. No visible solids or organic layer were observed in the individual samples. Individual sample densities were measured, then the five samples were mixed together to provide a single composite. The composite was homogenized and representative sub-samples taken for inorganic, radioisotopic, and organic analysis. All analyses were performed on triplicate sub-samples of the composite material. The sample composite did not contain visible solids or an organic layer. A subsample held at 10 C for seven days formed no visible solids. The characterization of the 241-AP-101 composite samples included: (1) Inductively-coupled plasma spectrometry for Ag, Al, Ba, Bi, Ca, Cd, Cr, Cu, Fe, K, La, Mg, Mn, Na, Nd, Ni, P, Pb, Pd, Ru, Rh, Si, Sr, Ti, U, Zn, and Zr (Note: Although not specified in the test plan, As, B, Be, Co, Li, Mo, Sb, Se, Sn, Tl, V, W, and Y were also measured and reported for information only) (2) Radioisotopic analyses for total alpha and total beta activities, 3H, 14C, 60Co, 79Se, 90Sr, 99Tc as pertechnetate, 106Ru/Rh, 125Sb, 134Cs, 137Cs, 152Eu, 154Eu, 155Eu, 238Pu, 239+240Pu, 241Am, 242Cm, and 243+244Cm; (3) Inductively-coupled plasma mass spectrometry for 237Np, 239Pu, 240Pu, 99Tc, 126Sn, 129I, 231Pa, 233U, 234U, 235U, 236U, 238U, 241AMU, 242AMU, 243AMU, As, B, Be, Ce, Co, Cs, Eu, I, Li, Mo, Pr, Rb, Sb, Se, Ta, Te, Th, Tl, V, and W; (4) total U by kinetic phosphorescence analysis; (5) Ion chromatography for Cl, F, NO2, NO3, PO4, SO4, acetate, formate, oxalate, and citrate; (6) Density, inorganic carbon and organic carbon by two different methods, mercury, free hydroxide, ammonia, and cyanide. The 241-AP-101 composite met all contract limits (molar ratio of analyte to sodium or ratio of becquerels of analyte to moles of sodium) defined in Specification 7 for Envelope A. Except for a few cases, the characterization results met or surpassed the

  2. Inorganic, radioisotopic and organic analysis of 241-AP-101 tank waste

    SK Fiskum; PR Bredt; JA Campbell; LR Greenwood; OT Farmer; GJ Lumetta; GM Mong; RT Ratner; CZ Soderquist; RG Swoboda; MW Urie; JJ Wagner

    2000-06-28

    Battelle received five samples from Hanford waste tank 241-AP-101, taken at five different depths within the tank. No visible solids or organic layer were observed in the individual samples. Individual sample densities were measured, then the five samples were mixed together to provide a single composite. The composite was homogenized and representative sub-samples taken for inorganic, radioisotopic, and organic analysis. All analyses were performed on triplicate sub-samples of the composite material. The sample composite did not contain visible solids or an organic layer. A subsample held at 10 C for seven days formed no visible solids. The characterization of the 241-AP-101 composite samples included: (1) Inductively-coupled plasma spectrometry for Ag, Al, Ba, Bi, Ca, Cd, Cr, Cu, Fe, K, La, Mg, Mn, Na, Nd, Ni, P, Pb, Pd, Ru, Rh, Si, Sr, Ti, U, Zn, and Zr (Note: Although not specified in the test plan, As, B, Be, Co, Li, Mo, Sb, Se, Sn, Tl, V, W, and Y were also measured and reported for information only) (2) Radioisotopic analyses for total alpha and total beta activities, {sup 3}H, {sup 14}C, {sup 60}Co, {sup 79}Se, {sup 90}Sr, {sup 99}Tc as pertechnetate, {sup 106}Ru/Rh, {sup 125}Sb, {sup 134}Cs, {sup 137}Cs, {sup 152}Eu, {sup 154}Eu, {sup 155}Eu, {sup 238}Pu, {sup 239+240}Pu, {sup 241}Am, {sup 242}Cm, and {sup 243+244}Cm; (3) Inductively-coupled plasma mass spectrometry for {sup 237}Np, {sup 239}Pu, {sup 240}Pu, {sup 99}Tc, {sup 126}Sn, {sup 129}I, {sup 231}Pa, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 241}AMU, {sup 242}AMU, {sup 243}AMU, As, B, Be, Ce, Co, Cs, Eu, I, Li, Mo, Pr, Rb, Sb, Se, Ta, Te, Th, Tl, V, and W; (4) total U by kinetic phosphorescence analysis; (5) Ion chromatography for Cl, F, NO{sub 2}, NO{sub 3}, PO{sub 4}, SO{sub 4}, acetate, formate, oxalate, and citrate; (6) Density, inorganic carbon and organic carbon by two different methods, mercury, free hydroxide, ammonia, and cyanide. The 241-AP-101 composite met all

  3. Uptake from water and tissue distribution of neptunium-237 in crabs, shrimp and mussels

    The uptake of 237Np has been followed in the tissues of mussels, shrimp and crabs exposed to the actinide in sea-water under controlled conditions. Bioaccumulation was observed in all tissues examined with the highest concentration factors noted in the external shells of the three species. Despite some incorporation of 237Np into internal tissues, it is noteworthy that after prolonged exposure 92-98% of the organisms' total 237Np content was associated with the non-edible shells. The general behaviour of this toxic transuranic element in invertebrate tissues appears to be very similar to that of plutonium. (author)

  4. Nuclear Criticality Safety of the DOT 9975 Container for237NpO2Storage, Handling, and Transport

    Reed, D.A.

    2003-08-29

    Nuclear criticality safety considerations are presented to address use of the DOT 9975 shipping container for {sup 237}NpO{sub 2}. The DOT 9975 container will be used by multiple DOE sites and contractors. Various of site- and activity-specific NCS and facility safety documents are yet to be developed. For these reasons, an overall assessment of criticality safety of {sup 237}NpO{sub 2}-loaded DOT 9975 containers is considered useful to personnel involved in generating, reviewing, or approving these various documents. It is concluded that inherent container features, the loading per container (maximum of 6 kg {sup 237}Np), and the nuclear physics properties of {sup 237}NpO{sub 2} combine to preclude the potential for a nuclear criticality accident. This conclusion applies to storage, handling, and transport operations involving closed DOT 9975 packages, including credible off-normal conditions that may result in damage to packages during those operations.

  5. Investigation on migration behavior of TRU-nuclides

    The migration behavior of TRU nuclides in geological formation has been studied by a batch and a column method, considering the influence of humic complexation and colloid formation. The migration test of 237Np and 238Pu is carrying out in natural field, to verify the data and the conventional migration model for the safety evaluation of TRU waste disposal. It was found that the migration of 237Np and 238Pu in the natural field is largely retarded sorption onto soil. (author)

  6. Fission cross section calculations for Pa isotopes

    Based on the recently measured cross-section values for the neutron-induced fission of 231Pa and our experience gained with other isotopes, new self consistent neutron cross section calculations for n+231Pa have been performed up to 30 MeV. The results are quite different to the existing evaluations, especially above the first chance fission threshold. (authors)

  7. Preliminary Neutronics Analysis Of Fuel Pebble With Thorium Fuel Cycle

    A new fuel pebble was designed based on Thorium fuel cycle. 231Pa has been added into fuel pebble for obtaining the minimum reactivity swing. The results show that the new designed pebble fuel with 7.0 % 233U enrichment adding 3.2% 231Pa, the keff is to be controlled up to 65 GWd/t; the other design with 8.0 % 233U enrichment requires 3.9% 231Pa, the keff therefore is remain up to 80 GWd/t. About 95% of loaded 231Pa in fuel pebble is depleted after 120 GWd/t. The results imply that it is optimistic to design the fuel pebble with 233U, 231Pa and 232Th; but some effects such as fuel temperature effect, distribution of TRISO particle in pebble fuel, etc. are required to investigate. (author)

  8. Neptunium Transport Behavior in the Vicinity of Underground Nuclear Tests at the Nevada Test Site

    Zhao, P; Tinnacher, R M; Zavarin, M; Williams, R W; Kersting, A B

    2010-12-03

    We used short lived {sup 239}Np as a yield tracer and state of the art magnetic sector ICP-MS to measure ultra low levels of {sup 237}Np in a number of 'hot wells' at the Nevada National Security Site (NNSS), formerly known as the Nevada Test Site (NTS). The results indicate that {sup 237}Np concentrations at the Almendro, Cambric, Dalhart, Cheshire and Chancellor sites, are in the range of 3 x 10{sup -5} to 7 x 10{sup -2} pCi/L and well below the MCL for alpha emitting radionuclides (15 pCi/L) (EPA, 2009). Thus, while Np transport is believed to occur at the NNSS, activities are expected to be well below the regulatory limits for alpha-emitting radionuclides. We also compared {sup 237}Np concentration data to other radionuclides, including tritium, {sup 14}C, {sup 36}Cl, {sup 99}Tc, {sup 129}I, and plutonium, to evaluate the relative {sup 237}Np transport behavior. Based on isotope ratios relative to published unclassified Radiologic Source Terms (Bowen et al., 1999) and taking into consideration radionuclide distribution between melt glass, rubble and groundwater (IAEA, 1998), {sup 237}Np appears to be substantially less mobile than tritium and other non-sorbing radionuclides, as expected. However, this analysis also suggests that {sup 237}Np mobility is surprisingly similar to that of plutonium. The similar transport behavior of Np and Pu can be explained by one of two possibilities: (1) Np(IV) and Pu(IV) oxidation states dominate under mildly reducing NNSS groundwater conditions resulting in similar transport behavior or (2) apparent Np transport is the result of transport of its parent {sup 241}Pu and {sup 241}Am isotopes and subsequent decay to {sup 237}Np. Finally, measured {sup 237}Np concentrations were compared to recent Hydrologic Source Term (HST) models. The 237Np data collected from three wells in Frenchman Flat (RNM-1, RNM-2S, and UE-5n) are in good agreement with recent HST transport model predictions (Carle et al., 2005). The agreement

  9. Subcellular localization of neptunium-237 in lung and kidney after intratracheal administration in the rat: An ultrastructural and microanalytical study

    Chronic intratacheal administration of 237Np to rats was performed during 6 weeks. The total dose administered was 45.8 kBq. Two methods, electron microscopy and electron probe X-ray microanalysis, were used to determined the intracellular sites of localization of 237Np. Clusters of dense granules were observed in nuclei of pneumocytes and proximal tubular cells of the kidneys. These clusters have been shown to contain neptunium associated with phosphorus, sulfur and calcium. Alternations of nuclei and ultrastructural cytoplasmic lesions were observed. The absorbed doses in lungs and kidneys were very low. These results suggest that the chemical toxicity of 237Np is more important than its radiological toxicity. 30 refs., 2 figs

  10. Method for Determination of Neptunium in Large-Sized Urine Samples Using Manganese Dioxide Coprecipitation and 242Pu as Yield Tracer

    Qiao, Jixin; Hou, Xiaolin; Roos, Per

    2013-01-01

    A novel method for bioassay of large volumes of human urine samples using manganese dioxide coprecipitation for preconcentration was developed for rapid determination of 237Np. 242Pu was utilized as a nonisotopic tracer to monitor the chemical yield of 237Np. A sequential injection extraction...... 100% and high separation capacity of processing up to 5 L of human urine samples. The MnO2 coprecipitation process is simple and straightforward in which a batch (8–12) of samples can be pretreated within 4 h (i.e., <0.5 h/sample). In connection with the automated column separation and ICPMS...... quantification, which takes less than 1.5 h in total, the overall analytical time was on average less than 2 h for each sample. The high effectiveness and sample throughput make the developed method well suited for urine bioassay of 237Np in routine monitoring of occupationally internal radiation exposure and...

  11. Study on the radionuclide elution from quartz sand source and their speciation

    The study of the elution of the radionuclide from quartz sand source and their speciation is carried out by column method. The colloid form of nuclides in the eluted solution is studied by ultra-filtration method. The existing form of the nuclides on the sand after elution is studied with sequential extraction method. When the volume of eluted solution is up to 110 times of the column volume, the nuclide 237Np of more than 90% is eluted, and in the exchangeable form. The nuclide 237Np remained in the sand is mainly in the form of being coincident with the matrix of carbonate (5.5%) and the transition metals, iron and manganese (1.3%). The elution of 238Pu from sand is more difficult than that of 237Np. 238Pu may exist in colloid form in eluted solution

  12. Non-conventional measurement techniques for the determination of some long-lived radionuclides produced in nuclear fuel

    The results of a literature survey on non-radiometric analytical techniques for the determination of long-lived radionuclides are described. The methods which have been considered are accelerator mass spectrometry, inductively coupled plasma mass spectrometry, thermal ionization mass spectrometry, resonance ionization spectrometry, resonance ionization mass spectrometry and neutron activation analysis. Neutron activation analysis has been commonly used for the determination of 129I and 237Np in environmental samples. Inductively coupled mass spectrometry seems likely to become the method of choice for the determination of 99Tc, 237Np and Pu-isotopes. The methods are discussed and the chemical separation methods described. (orig.)

  13. Systematic study of anomalous fragment anisotropies in near- and sub-barrier fusion-fission reactions

    The fusion cross sections and fragment angular distributions for the complete fusion-fission reactions of 11B+238U, 237Np, 12C+237Np, 16O+232Th, 238U, and 19F+23Th at near- and sub-barrier energies have been measured by the fragment folding angle technique. It is revealed that the anomalous anisotropies of fission fragments in latter three systems are existence. Based on the experimental observations and Dressing and Randrup's theory, a new version model of preequilibrium fission is put forward to explain the anomaly. (author)

  14. Distribution of neptunium and plutonium in New Mexico lichen samples (Usnea arizonica) contaminated by atmospheric fallout

    The concentrations of 237Np, 239Pu and 240Pu were determined in lichen samples (Usnea arizonica) that were collected from ten locations in New Mexico between 2011 and 2013 using isotope dilution inductively-coupled plasma mass spectrometry (ID-ICP-MS). The observed isotopic ratios for 237Np/239Pu and 240Pu/239Pu indicate trace contamination from global and regional fallout (e.g. Trinity test and atmospheric testing at the Nevada Test Site). The fact that actinide contamination is detected in recent lichen collections suggests continuous re-suspension of fallout radionuclides even 50 years after ratification of the Limited Test Ban Treaty. (author)

  15. Non-conventional measurement techniques for the determination of some long-lived radionuclides produced in nuclear fuel

    The results of a literature survey on non-radiometric analytical techniques for the determination of long-lived radionuclides are described. The methods which have been considered are accelerator mass spectrometry, inductively coupled plasma mass spectrometry, thermal ionization mass spectrometry, resonance ionization spectrometry, resonance ionization mass spectrometry and neutron neutron activation analysis. Neutron activation analysis has been commonly used for the determination of 129I and 237Np in environmental samples. Inductively coupled mass spectrometry seems likely to become the method of choice for the determination of 99Tc, 237Np and Pu-isotopes. The methods are discussed and the chemical separation methods described. (author) 88 refs.; 8 tabs

  16. Neptunium determination by inductively coupled plasma mass spectrometry (ICP-MS)

    The determination of neptunium-237 (237Np) traditionally has been performed by alpha spectrometry or neutron activation analysis. These methods are labor intensive and require several days for completion. Inductively Coupled Plasma Mass Spectrometry (ICP-MS) is a possible alternative for 237Np determinations. This paper describes the analytical method developed for samples that have significant levels of uranium present. The lower reporting limits achievable by ICP-MS are competitive with the counting methods, but the real advantage for this laboratory lies in the lower cost and faster turnaround time provided by ICP-MS. (author)

  17. Neutron capture cross section measurement of Np-237 below 10 keV by linac time-of-flight method

    Lee, Samyol; Yamamoto, Shuji; Cho, Hyun-Je; Yoshimoto, Takaaki; Kobayashi, Katsuhei; Fujita, Yoshiaki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Ohkawachi, Yasushi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2001-03-01

    The neutron capture cross section of {sup 237}Np has been measured in the energy region from 0.01 eV to 10 keV by using the neutron time-of-flight (TOF) method with a 46 MeV electron linear accelerator (linac) at the Research Reactor Institute, Kyoto University (KURRI). A pair of C{sub 6}D{sub 6} scintillation detectors, which was placed at a distance of 12.0 {+-} 0.02 m from the pulsed neutron source, was employed for the prompt capture gamma-ray measurement from the {sup 237}Np sample. The measured result has been normalized to the reference value of the {sup 237}Np(n,{gamma}){sup 238}Np reaction in ENDF/B-VI at 0.0253 eV. The existing experimental and the evaluated capture cross sections in ENDF/B-VI and JENDL-3.2 have been compared with the present measurement. For the neutron capture cross section of {sup 237}Np, the data by Weston et al. and the evaluated data are in good agreement with the present measurement. However, the data by Hoffman et al. are obviously lower in the relevant energy region. The data, which were measured before using a lead slowing-down spectrometer at KURRI, have been in good agreement with the data obtained by energy-broadening the present TOF measurement. (author)

  18. Ionic speciation of Np in frozen solutions

    Light actinides exhibit many oxidation states (II) that are stable in salts and are relatively stable in solution. Np IV to Np VII solutions were prepared and quickly frozen. We report the preliminary results of the 237Np Mossbauer study. These results are discussed and compared to those available for Np compounds in the crystalline state or Np ions in solution. 3 refs

  19. Hanford intercontractor support

    Distribution coefficients (Kd values) were determined on subsoils from Washington and South Carolina for 241Am, 237Np, and 99Tc as a function of equilibrium solution concentration of calcium (Ca2+) and of sodium (Na+). Kd values decreased in all cases with increasing solution concentrations of Ca2+ and Na+. For the South Carolina subsoil Kd values ranged from 1.0 to 67 for 241Am as a function of Ca2+, from 0.2 to 0.002 M, respectively, 1.6 to 280 for 241Am as a function of Na+, 0.43 to 0.66 for 237Np as a function of Ca2+, and 0.16 to 0.25 for 237Np as a function of Na+ from 3.0 less than 0.015 M, respectively. For the Washington soil, Kd values were greater than 1200 for 241Am and ranged from 0.36 to 2.37 as a function of Ca2+ and from 3.19 to 3.90 for 237Np as a function of Na+ over the above concentration ranges, respectively. Kd values for 99Tc were essentially 0 at all NaHCO3 concentrations on the South Carolina subsoil

  20. Preparation of tracer 239Np from 243Am

    The aim of this work is the preparation of tracer 239Np from 243Am by extraction chromatography as a separation method that can be used as a tracer to determine the radiochemical yield of 237Np in the analysis of environmental samples. (authors)

  1. Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity - Irradiation Tests of Np and Np-U Samples in Experimental Fast Reactor Joyo (JAEA) and Advanced Thermal Reactor ATR (INL)

    A project of Produce Protected Plutonium (P3) was proposed by Tokyo-Tech as a part of non-proliferation research for Plutonium (Pu) utilization to nuclear reactor. The project is to reach the production of inherently protected Pu by addition 237Np to Uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. Experimental determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor Joyo was done to evaluate 238Pu production from 237Np under the fast neutron spectra. The amount of 238Pu in the irradiated 237Np samples was determined by a radiochemical analysis in Alpha-gamma Facility of JAEA. The produced 238Pu in the samples was found to depend on the neutron spectrum, ranging from that of a typical fast reactor to a nearly epi-thermal spectrum. The fast reactor can potentially control the 238Pu production from 237Np by the spectrum shift in the different irradiation position. Within the framework of the P3, the fast reactor can make roles of protected Pu production which can be better performed in the reflector region, the ratio of 238Pu is achieved up to around 90% produced from 237Np. On the other hand, 2, 5 and 10% 237Np containing U samples were also irradiated in Advanced Thermal Reactor of INL to evaluate the 238Pu production under thermal neutron region. The irradiation condition and its loading position of samples were fixed based on a calculation with MCWO (MCNP Coupled with ORIGEN2) code. The fuel specimens were removed from the core at 100, 200 and 300 effective full power days (EFPD), and then post irradiation examination was completed at Chemical lab. in MFC of INL. For the samples after 300 EFPD irradiation, Np depletion were about 60 % for 2% Np-U samples, about 50% for 5 and 10 % Np-U samples. The 238Pu to Pu ratio was about 20%, 30% and 45% for 2%, 5%, and 10% Np-U samples, respectively. The neutronics calculation results were coincident with the experimental ones. Acknowledgments: The

  2. Impact of fission product partitioning and transmutation of Np-237, I-127 and Tc-99 on waste disposal strategies

    From the point of view of waste management, three isotopes 237Np, 129I and 99Tc deserve particular attention since they cannot be confined within any presently known geological barrier. The safety assessment studies showed that 129I diffuses through 50 m of clay after 5000 years and creates a radiological burden which is superior to the other nuclides. 99Tc and 237Np emerge much later (million years) but cannot be confined in the repository. Separation of 237Np is already envisaged by the reprocessing plants and small modifications to the PUREX process permit to separate this nuclide quantitatively. However, the overall Np reduction factor cannot be higher than 3 to 4 because 237Np is formed by alpha-decay of 241Am which is immobilized in the glass. The irradiation technology to transmute 237Np into 238Pu is available. Iodine is separated from the dissolver off-gases in the reference fuel cycle and could be transformed into a stable Ba(IO3)2 matrix which can serve as irradiation 1291 matrix to destroy. Technetium is produced as insoluble residue and occurs in the waste streams together with Ru, Rh and Pd. Its recovery is still questionable and needs further developments. Irradiation of small targets in HFRs can lead to a 90% transmutation of these long-lived isotopes but the irradiation periods (particularly for 129I) are very long. The present HFR generation is not designed to handle large amounts of long-lived radionuclides and new more powerful units ought to be designed for that purpose. (author)

  3. Impact of fission product partitioning and transmutation of Np-237, I-127 and Tc-99 on waste disposal strategies

    Beatsle, L.H. [Studiecentrum voor Kernenergie - Centre d' Etude de l' Energie Nucleaire (SCK-CEN), Mol (Belgium)

    1991-07-01

    From the point of view of waste management, three isotopes {sup 237}Np, {sup 129}I and {sup 99}Tc deserve particular attention since they cannot be confined within any presently known geological barrier. The safety assessment studies showed that {sup 129}I diffuses through 50 m of clay after 5000 years and creates a radiological burden which is superior to the other nuclides. {sup 99}Tc and {sup 237}Np emerge much later (million years) but cannot be confined in the repository. Separation of {sup 237}Np is already envisaged by the reprocessing plants and small modifications to the PUREX process permit to separate this nuclide quantitatively. However, the overall Np reduction factor cannot be higher than 3 to 4 because {sup 237}Np is formed by alpha-decay of {sup 241}Am which is immobilized in the glass. The irradiation technology to transmute {sup 237}Np into {sup 238}Pu is available. Iodine is separated from the dissolver off-gases in the reference fuel cycle and could be transformed into a stable Ba(IO{sub 3}){sub 2} matrix which can serve as irradiation 1291 matrix to destroy. Technetium is produced as insoluble residue and occurs in the waste streams together with Ru, Rh and Pd. Its recovery is still questionable and needs further developments. Irradiation of small targets in HFRs can lead to a 90% transmutation of these long-lived isotopes but the irradiation periods (particularly for {sup 129}I) are very long. The present HFR generation is not designed to handle large amounts of long-lived radionuclides and new more powerful units ought to be designed for that purpose. (author)

  4. ZZ MATXSLIBJ33, JENDL-3.3 based, 175 N-42 photon groups (VITAMIN-J) MATXS library for discrete ordinates multi-group

    -156, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Tb-159, Er-162, Er-164, Er-166, Er-167, Er-168, Er-170, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, W-182, W-183, W-184, W-186, Hg-196, Hg-198, Hg-199, Hg-200, Hg-201, Hg-202, Hg-204, Pb-204, Pb-206, Pb-207, Pb-208, Bi-209, Ra-223, Ra-224, Ra-225, Ra-226, Ac-225, Ac-226, Ac-227, Th-227, Th-228, Th-229, Th-230, Th-232, Th-233, Th-234, Pa-231, Pa-232, Pa-233, U-232, U-233, U-234, U-235, U-236, U-237, U-238, Np-235, Np-236, Np-237, Np-238, Np-239, Pu-236, Pu-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Pu-244, Pu-246, Am-241, Am-242, Am-242m, Am-243, Am-244, Am-244m, Cm-240, Cm-241, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246, Cm-247, Cm-248, Cm-249, Cm-250, Bk-247, Bk-249, Bk-250, Cf-249, Cf-250, Cf-251, Cf-252, Cf-254, Es-254, Es-255, Fm-255 Temperatures: 300 K. Origin: JENDL-3.3. Weighting spectrum: -- iwt=11 for NJOY-99. Legendre expansion: P6. Thermal scattering: free gas model. Self shielding: sigma-0, infinity, 10000, 1000, 300, 100, 30, 10, 1, 0.1, 1. E-5. Kerma factors are provided. NEA-1707/03: Corrections were made to the continuous inelastic scattering matrices (MT=91), for all nuclides for which this channel is open. This replaces the previous version. 2 - Methods: The nuclear data processing system NJOY-99.67 was used to produce MATXSLIBJ33. It can be further processed using TRANSX-2.15. 3 - Restrictions on the complexity of the problem: Accuracy of pointwise cross-section reconstruction: 0.1%; Upper limit of thermal region: 4.6 eV

  5. ZZ KASHIL-E70, 199 N, 42 Photon Groups Cross Sections in MATXS Format Based on ENDF/B-VII.0 for Shielding Applications

    1 - Description: Format: MATXS, 204 nuclides processed with NJOY99.245. Number of groups: 199 neutron-, 42 photon-groups. 204 Nuclides including 8 thermal scattering law data: H-1, H-2, H-3, He-3, He-4, Li-6, Li-7, Be-9, Be-9, Be-9, B-10, B-11, C-nat, C-nat, N-14, N-15, O-16, O-17, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, S-32, S-33, S-34, S-36, Cl-35, Cl-37, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Sc-45, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Y-89, Zr-90, Zr-91, Zr-92, Zr-94, Zr-96, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-100, Pd-102, Pd-104, Pd-105, Pd-106, Pd-108, Pd-110, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-112, Cd-113, Cd-114, Cd-116, In-113, In-115, I-127, Xe-124, Xe-126, Xe-128, Xe-129, Xe-130, Xe-131, Xe-132, Xe-134, Xe-136, Cs-133, Ba-138, Pr-141, Nd-143, Nd-145, Nd-146, Nd-148, Nd-150, Pm-147, Sm-147, Sm-151, Sm-152, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Dy-164, Ho-165, Lu-175, Lu-176, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Ir-191, Ir-193, Au-197, Pb-206, Pb-207, Pb-208, Bi-209, Th-230, Th-232, Pa-231, Pa-233, U-232, U-233, U-234, U-235, U-236, U-237, U-238, Np-237, Np-238, Np-239, Pu-236, Pu-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Pu-243, Pu-244, Am-241, Am-242, Am-242m, Am-243, Cm-241, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246, Cm-247, Cm-248, Bk-249, Cf-249, Cf-250, Cf-251, Cf-252, Cf-253, Es-253. Origin: ENDF/B-VII.0. Weighting spectrum: 300, 600, 1000, 2100 K. The KASHIL-E70 is a MATXS-format, 199-group neutron and 42-group photon cross section library for shielding applications based on ENDF/B-VII.0. The library contains 204 nuclide data including 8 thermal scattering law data processed by the NJOY99.259 code patched with NEA

  6. ZZ MCB-JEF2.2, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K

    -154, Eu-155, Eu-156, Eu-157, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Tb-159, Tb-160, Dy-160, Dy-161, Dy-162, Dy-163, Dy-164, Ho-165, Er-166, Er-167, Lu-175, Lu-176, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-nat., Bi-209, Th-230, Th-232, Pa-231, Pa-233, U-232, U-233, U-234, U-235, U-236, U-237, U-238, Np-237, Np-238, Np-239, Pu-236, Pu-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Pu-243, Pu-244, Am-241, Am-242, Am-242m, Am-243, Cm-241, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246, Cm-247, Cm-248, Bk-249, Cf-249, Cf-250, Cf-251, Cf-252, Cf-253, Es-253. Temperatures: 300 K, 600 K, 900 K, 1200 K, 1500 K, and 1800 K. Origin: JEF-2.2. 2 - Methods: The library was generated using the NJOY processing code. An example of the input data is provided

  7. Optimization of small long-life PWR based on thorium fuel

    Subkhi, Moh Nurul; Suud, Zaki; Waris, Abdul; Permana, Sidik

    2015-09-01

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% 233U & 2.8% 231Pa, 6% 233U & 2.8% 231Pa and 7% 233U & 6% 231Pa give low excess reactivity.

  8. Optimization of small long-life PWR based on thorium fuel

    Subkhi, Moh Nurul, E-mail: nsubkhi@students.itb.ac.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung (Indonesia); Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung Djati Bandung Jalan A.H Nasution 105 Bandung (Indonesia); Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Waris, Abdul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {sup 231}Pa give low excess reactivity.

  9. Protactinium and isotopes of thorium in metalliferous sediments from the Bauer depression

    Results are presented of a study of the vertical distribution of 238U, 234U, 232Th, 230Th, and 231Pa isotopes in a column of metalliferous sediments from the Bauer depression (southeastern part of the Pacific Ocean). On the basis of the obtained data a hypothesis is formulated concerning the authigenic production of 230Th and 231Pa in these deposits, i.e, the similarity of the physicochemical behavior of 230Th and 231Pa found in pelagic sediments is found in these specific sediments also. We present arguments in favor of the identical behavior of these radionuclides in the marine environment. With the help of the ionium method of dating marine sediments, the average rate of sedimentation of the investigated column of metalliferous sediments from the Bauer depression was calculated

  10. Protactinium and thorium isotopes in metal-containing precipitates from Bauer depression

    The results studying vertical distribution of 238U, 234U, 232Th, 230Th and 231Pa isotopes in the colomn of metalliferous sediments from Bauer depression (South-East part of the Pacific ocean) are presented. On the basis of the data obtained the supposition about the authigenic origin of 230Th and 231Pa in these ocean sedimentations, is stated, i.e. even in such specific ocean sediments closeness of physical chemical behaviour of 230Th and 231Pa is fixed. Arguments for identical behaviour of these radionuclides in sea water are presented. Average rate of sedimentation of the column investigated of metalliferous sediments from Bauer depression is calculated using the ionic method of ocean deposition dating

  11. Optimization of small long-life PWR based on thorium fuel

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% 233U & 2.8% 231Pa, 6% 233U & 2.8% 231Pa and 7% 233U & 6% 231Pa give low excess reactivity

  12. Application of off/on line sequential injection system with high resolution ICP-MS to measurement radionuclides in environmental samples

    Chang-Kyu Kim (IAEA, Physics, Chemistry and Instrumentation Lab., Agency' s Laboratories, Seibersdorf (Austria))

    2010-03-15

    A sequential injection system was developed, which can be widely used for the separation and preconcentration of analytes from diverse environmental samples. The system enables the separation time to be shortened by maintaining a constant flow rate of solution and by avoiding clogging or bubbling in a chromatographic column. The SI system was successfully applied to the separation of 237Np and Pu isotopes in IAEA reference materials and environmental samples, and to the sequential separation of 210Po and 210Pb in a phosphogypsum candidate reference material. The replicate analysis results of 237Np, 239+240Pu 210Po and 210Pb in some IAEA reference materials using the SI system associated with HR-ICP-MS, alpha-spectrometry and LSC are in good agreement with the recommended value within 5% of standard deviation. The SI system enabled a halving of the separation time required for radionuclides. (author)

  13. A chromatographic separation of neptunium and protactinium using 1-octanol impregnated onto a solid phase support

    We have developed a new chromatographic method to efficiently separate and isolate neptunium (Np) and protactinium (Pa), based on the selective extraction of protactinium by primary alcohols. The effectiveness of the new technology is demonstrated by efficient separation of 233Pa from parent radionuclide 237Np, using a hydrochloric acid mobile-phase medium. Our new approach reproducibly isolated 233Pa tracer with a yield of 99 ± 1 % (n = 3; radiochemical purity 100 %) and enabled chemical recovery of 237Np parent material of 92 ± 3 % (radiochemical [99 %) for future 233Pa tracer preparations. Compared to previous methods, the new approach reduces radioactive inorganic and organic waste; simplifies the separation process by eliminating cumbersome liquid-liquid extractions; and allows isolation of radiochemically-pure fractions in less than 1 h. (author)

  14. The Geochemical Behavior of Tc, Np, and Pu in Spent Nuclear Fuel in an Oxidizing Environment

    Studies at the Nopal and Shinkolowbwe uranium deposits show that the primary uraninite (UO2) altered to a suite of secondary uranyl minerals similar to those observed in corrosion tests with uranium oxide . Although the Nopal I deposit tells us something about the possible fate of uranium, it tells us little about the likely fate of the important long-lived radionuclides; iodine (129I), cesium (135Cs), technetium (99Tc), neptunium (237Np), and plutonium (239Pu). Most performance assessment (PA) models, assume conservatively, that as the UO2 matrix corrodes, the key radionuclides (129I, 99Tc, 237Np, and 239Pu) will be released congruently. In so doing, these PA models force increased reliance on human engineered barriers

  15. The Geochemical Behaviour of Tc, Np, and Pu in Spent Nuclear Fuel in an Oxidizing Environment

    Buck, Edgar C.; Hanson, Brady D.; McNamara, Bruce K.; R. Giere and P. Stille

    2004-10-01

    Studies at the Nopal and Shinkolowbwe uranium deposits show that the primary uraninite (UO2) altered to a suite of secondary uranyl minerals similar to those observed in corrosion tests with uranium oxide . Although the Nopal I deposit tells us something about the possible fate of uranium, it tells us little about the likely fate of the important long-lived radionuclides; iodine (129I), cesium (135Cs), technetium (99Tc), neptunium (237Np), and plutonium (239Pu). Most performance assessment (PA) models, assume conservatively, that as the UO2 matrix corrodes, the key radionuclides (129I, 99Tc, 237Np, and 239Pu) will be released congruently. In so doing, these PA models force increased reliance on human engineered barriers.

  16. Standardization of 241Am by digital coincidence counting, liquid scintillation counting and defined solid angle counting

    Balpardo, C; Rodrigues, D; Arenillas, P

    2010-01-01

    The nuclide 241Am decays by alpha emission to 237Np. Most of the decays (84.6 %) populate the excited level of 237Np with energy of 59.54 keV. Digital Coincidence Counting was applied to standardize a solution of 241Am by alpha-gamma coincidence counting with efficiency extrapolation. Electronic discrimination was implemented with a pressurized proportional counter and the results were compared with two other independent techniques: Liquid Scintillation Counting using the logical sum of double coincidences in a TDCR array and Defined Solid Angle Counting taking into account activity inhomogeneity in the active deposit. The results show consistency between the three methods within a limit of a 0.3%. An ampoule of this solution will be sent to the International Reference System (SIR) during 2009. Uncertainties were analysed and compared in detail for the three applied methods.

  17. Neptunium distribution in PUREX process streams

    237Np is one of the most important minor actinides present in spent fuel both from environmental and application point of view. The routing of neptunium to the particular waste stream of PUREX process is required for its separation and purification as 237Np is the target nuclide for production of 238Pu. In addition, the routing of neptunium to a particular PUREX stream will help in better waste management, which in turn will reduce its bearing on the environment considering its long half life, alpha emitting properties and mobile nature. In order to route Neptunium to a particular waste stream of PUREX process, it is imperative to understand the distribution of neptunium in various process streams. Although, there are reports on Np distribution under simulated conditions of PUREX streams, the present study deals with neptunium determination in actual PUREX streams samples. (author)

  18. Solid state track recorder measurements in the poolside critical assembly

    Fission rate measurements using solid state track recorders (SSTR) have been performed at the PCA. A schematic representation of a cross-section of the PCA is shown. Fission rates were measured in the pressure vessel simulator at the T/4, T/2 and 3T/4 positions and in the void box (VB). SSTR measurements were carried out with 232Th, 235U (bare and cadmium covered), 238U and 237Np fissionable deposits. Midplane only measurements were carried out for 235U and 237Np, while 5 axial locations at 1/4T and 1/2T and 3 axial locations at 3/4T and in the VB were sampled for 232Th and 238U. The HEDL SSTR fission rate measurements reported herein for both configurations together with NBS and CEN/SCK fission chamber measurements will be used to establish absolute and relative fission reaction rates, and ratios for the PCA pressure vessel Benchmark Facility

  19. Angular distributions in the neutron-induced fission of actinides

    In 2003 the n_TOF Collaboration performed the fission cross section measurement of several actinides ($^{232}$Th, $^{233}$U, $^{234}$U, $^{237}$Np) at the n_TOF facility using an experImental setup made of Parallel Plate Avalanche Counters (PPAC). The method based on the detection of the 2 fragments in coincidence allowed to clearly disentangle the fission reactions among other types of reactions occurring in the spallation domain. We have been therefore able to cover the very broad neutron energy range 1eV-1GeV, taking full benefit of the unique characteristics of the n_TOF facility. Figure 1 shows an example obtained in the case of $^{237}$Np where the n_ TOF measurement showed that the cross section was underestimated by a large factor in the resonance region.

  20. Refinement of Pu parent-daughter isotopic and concentration analysis for forensic (dating) purposes

    Plutonium (Pu) metal samples from an interlaboratory exchange exercise and simulated swipe samples were dated using plutonium-uranium (Pu-U) and plutonium-americium (Pu-Am). Metal data were evaluated for consistency and the swipe data against its source material. Metal ages based on 239Pu versus 235U and 240Pu versus 236U agreed to within a few percent, while the 238Pu-234U and 241Pu-241Am measurements had larger uncertainties. Swipe ages compared favorably with the material's known history. Neptunium (237Np) analyses were examined in the context of the 241Pu-241Am-237Np system to estimate whether Np can provide insights on material from which Am, Np, and U were removed. (author)

  1. Calculation of fission fragment angular anisotropy in heavy-ion induced fission

    Fission fragment angular anisotropies from 16O + 232Th, 12C + 236U, 11B + 237Np, 14N + 232Th, 11B + 235U and 12C + 232Th systems were calculated by means of the standard saddle point statistical model (SSPSM). The results were obtained with and without neutron emission correction in the reactions, and comparisons were made with the corresponding experimental data. The normal and anomalous behaviors of fission fragment anisotropies are extensively discussed. (author)

  2. Measurements of Fission Cross Sections of Actinides

    Wiescher, M; Cox, J; Dahlfors, M

    2002-01-01

    A measurement of the neutron induced fission cross sections of $^{237}$Np, $^{241},{243}$Am and of $^{245}$Cm is proposed for the n_TOF neutron beam. Two sets of fission detectors will be used: one based on PPAC counters and another based on a fast ionization chamber (FIC). A total of 5x10$^{18}$ protons are requested for the entire fission measurement campaign.

  3. Translation of selected papers published in Nuclear Constants, No. 3(57), Moscow 1984

    The document contains the English translation of 6 papers selected for the nuclear data interest, (mainly the determination and the evaluation of the fast fission cross sections of 232Th, 235U, 238U, 237Np, 243Am) which were published in ''Topics in Atomic Science and Technology'', Series Nuclear Constants, No. 3(57), Moscow (1984). The original report was distributed as INDC(CCP)-240/G. Refs, figs and tabs

  4. Selected articles translated from Jadernye Konstanty (Nuclear Constants) volumes 1-2, 1995

    This report contains selected articles translated from Russian language which are dealing with nuclear constants. The MENDL-2 cross section library, universal library of fission products, evaluation of fission cross section for 237Np, evaluation of the photoneutron reaction cross sections, analysis of reaction rate measurements for determination of neutron energy spectra are specially dealt with. It contains five articles and each one is separately indexed. Refs, figs, tabs

  5. Analysis of the evaluated data discrepancies for minor actinides and development of improved evaluation

    Ignatyuk, A. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-03-01

    The work is directed on a compilation of experimental and evaluated data available for neutron induced reaction cross sections on {sup 237}Np, {sup 241}Am, {sup 242m}Am and {sup 243}Am isotopes, on the analysis of the old data and renormalizations connected with changes of standards and on the comparison of experimental data with theoretical calculation. Main results of the analysis performed by now are presented in this report. (J.P.N.)

  6. Average resonance parameters evaluation for actinides

    Porodzinskij, Yu.V.; Sukhovitskij, E.Sh. [Radiation Physics and Chemistry Problems Inst., Minsk-Sosny (Belarus)

    1997-03-01

    New evaluated <{Gamma}{sub n}{sup 0}> and values for {sup 238}U, {sup 237}Np, {sup 243}Cm, {sup 245}Cm, {sup 246}Cm and {sup 241}Am nuclei in the resolved resonance region are presented. The applied method based on the idea that experimental resonance missing results in correlated changes of reduced neutron widths and level spacings distributions is discussed. (author)

  7. Actinides separation and long-lived fission products from the high activity effluent

    The aim of this document is to study the decontamination of a high activity effluent in minor actinides-α transmitters (241Am, 243Am, 243Cm, 245Cm, 237Np, 238Pu, 242Pu, 235U, 238U) and long-life fissions products (133Cs, 137Cs) and then the separation of Am, Cm, Np, Cs and Pu, U traces. (TEC). 16 figs., 1 tab

  8. Isoscaling and fission modes in the yields of the Kr and Xe isotopes from photofission of actinides

    Drnoyan, J.; Zhemenik, V. I.; Mishinsky, G. V.

    2016-05-01

    Yields of Kr and Xe isotopes in photofission of 232Th, 238U, 237Np, 244Pu, 243Am, and 248Cm were tested for isoscaling dependence. Isoscaling for Kr is revealed. For Xe, isoscaling is found to be affected by the STI and STII fission modes governed by the N = 82 and N = 88 neutron shells. The work was performed at the Flerov Laboratory of Nuclear Reactions, Joint Institute for Nuclear Research (JINR).

  9. Modelling of reaction cross sections and prompt neutron emission

    Oberstedt S.; Tudora A.; Hambsch F.-J.

    2010-01-01

    Accurate nuclear data concerning reaction cross sections and the emission of prompt fission neutrons (i.e. multiplicity and spectra) as well as other fission fragment data are of great importance for reactor physics design, especially for the new Generation IV nuclear energy systems. During the past years for several actinides (238U(n, f) and 237Np(n, f)) both the reaction cross sections and prompt neutron multiplicities and spectra have been calculated within the frame of the EFNUDAT project.

  10. The fission fragment yields at the photofission of actinide nuclei

    The fission fragment yields of isotopes 101Mo, 135I, 135mCs were measured at the photo-fission of actinide nuclei 232Th, 238U, 237Np. These fission fragments have some peculiarities in nuclear structure or in practical using. The measurements were performed on the microtron bremsstrahlung at the Flerov Laboratory of Nuclear Reactions, JINR, at the electron energy 22 MeV. The activation method with an HPGe detector was used in these measurements of the yields

  11. FFTF (FAST FLUX TEST FACILITY) REACTOR CHARACTERIZATION PROGRAM ABSOLUTE FISSION RATE MEASUREMENTS

    FULLER JL; GILLIAM DM; GRUNDL JA; RAWLINS JA; DAUGHTRY JW

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  12. FFTF (Fast Flux Test Facility) Reactor Characterization Program: Absolute Fission-rate Measurements

    Fuller, J.L.; Gilliam, D.M.; Grundl, J.A.; Rawlins, J.A.; Daughtry, J.W.

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  13. Nordic collaboration on the use of mass-spectrometers for the analysis of radioisotopes. NKS-project NORCMASS. Final report

    This report cover an overview of the work performed during a three year (2003-2005) project initialized with the purpose of identifying and work on problems in isotope ratio and ultra trace measurements of primarily plutonium and uranium isotopes and 237Np using ICPMS. The project also included an educational part aiming to describe fundamental aspects and practical steps for radioisotope measurements using ICP-MS. (au)

  14. Electron paramagnetic resonance and electron nuclear double resonance of 237-neptunium hexafluoride in uranium hexafluoride single crystals

    Butler, James E.; Hutchison, Clyde A., Jr.

    1981-03-01

    The EPR and ENDOR spectra of 237NpF6 molecules dilutely substituted for host molecules in single crystals of UF6 at temperatures between 1.2 and 2.1 °K have been obtained at microwave frequencies, ˜9.4 and ˜9.7 GHz. Approximate values are given for the parameters in a spin Hamiltonian formalism that describes the measurements. The results are discussed.

  15. Optimization of solvent extraction cycles to separate plutonium and neptunium

    This paper describes the steps taken to assess the potential of a solvent extraction process for separating plutonium from neptunium. The plutonium, namely 238Pu, is of importance because it is the isotope used in making radioisotope thermoelectric generators and radioisotope heaters and is obtained by irradiating 237Np targets in a high neutron flux environment. After the neptunium targets have been irradiated, the targets could then be processed by solvent extraction methods

  16. High-sensitivity fast neutron detector KNK-2-7M

    The construction of the fast neutron detector KNK-2-7M is briefly described. The results of the study of the detector in the pulse-counting mode are given for the fissions of 237Np nuclei in the radiator of the neutron-sensitive section and in the current mode with the separation of sectional currents of functional sections. The possibilities of determining the effective number of 237Np nuclei in the radiator of the neutronsensitive section are considered. The diagnostic possibilities of the detector in the counting mode are shown by example of the analysis of the reference data from the neutron-field characteristics in the working hall of the BR-K1 reactor. The diagnostic possibilities of the detector in the current operating mode are shown by example of the results of measuring the 237Np-fission intensity in the BR-K1 reactor power start-ups implemented in the mode of fission-pulse generation on delayed neutrons at the detector arrangement inside the reactor core cavity under conditions of a wide variation of the reactor radiation field

  17. Investigation of the nuclear data on the neutron beam of the IBR-30 and IBR-2 reactors

    Methodical aspects and results of the measurements, conducted on reactors IBR-30 and IBR-2 are presented. Measurements of the multiplicity spectra of gamma-quanta from the neutron capture reaction on the isotopes 48Ti, 113,115In, 117Sn, 127J, 149Sm, 165Ho, 175Lu, 177Hf, 178Hf, 185Re, 187Re, and 232Th have been done. These spectra were used for determination of the capture cross-section and according resonance parameters in the fission of 235U, 239Pu. The measurements were performed of resonance spins of isotopes 113,115In, 117Sn, 185,187Re, and 235U by the low-lying level population method in the thermal neutron energy area. The transmission and self-indication functions of filter samples 232Th, 237Np, and 238U were measured with different temperatures for determination the neutron cross-sections in the neutron energy range from 1 eV to 100 keV. The fission cross-section measurements of the minor actinides 234U, 237Np, and 243Am by the neutrons with energy below the fission barrier were carried out. The measurements of total yields and decay curve of the delayed neutrons from fission of 233,235U, 237Np by thermal and cold neutrons were performed

  18. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2013-03-01

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%).

  19. The (3He, tf) as a surrogate reaction to determine (n, f) cross sections in the 10-20 MeV energy range

    The surrogate reaction 238U(3He, tf) is used to determine the 237Np(n, f) cross section indirectly over an equivalent neutron energy range from 10 to 20 MeV. A self-supporting ∼761 μg/cm2 metallic 238U foil was bombarded with a 42 MeV 3He2+ beam from the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory (LBNL). Outgoing charged particles and fission fragments were identified using the Silicon Telescope Array for Reaction Studies (STARS) consisted of two 140 μm and one 1000 μm Micron S2 type silicon detectors. The 237Np(n, f) cross sections, determined indirectly, were compared with the 237Np(n, f) cross section data from direct measurements, the Evaluated Nuclear Data File (ENDF/B-VII.0), and the Japanese Evaluated Nuclear Data Library (JENDL 3.3) and found to closely follow those datasets. Use of the (3He, tf) reaction as a surrogate to extract (n, f) cross sections in the 10-20 MeV equivalent neutron energy range is found to be suitable.

  20. Radiochemical determination of Np-237 in soil samples contaminated with weapon grade plutonium

    The Palomares terrestrial ecosystem (Andalusia, southwestern Spain) is known to constitute a natural laboratory to study the distribution, behaviour and migration of certain actinides, such as plutonium, americium and neptunium. This scenario is partially contaminated with weapon grade plutonium since the burn-out and fragmentation of two of the four thermonuclear bombs accidentally dropped by a B-52 from the USA Air Force back in 1966. While performing radiometric measurements on the field, with the goal of gathering information about the surface contamination levels, the possible presence of 237Np was observed through its 29 keV gamma emission. To accomplish a more detailed characterization of the source term in the contaminated area using the isotopic ratios Pu-Am-Np, the radiochemical isolation and later quantification by alpha spectrometry of 237Np was initiated. As a first approach, a bibliographic study was undertaken, considering different radiochemical methods applied to a variety of samples such as soils, sediments, sea water, lichens, etc. The radiochemical procedure selected in our laboratory involves separation of neptunium from americium, uranium and plutonium with ionic resin (AG 1x2), given that in soil samples from Palomares 239+240Pu levels are several orders of magnitude higher than 237Np. Then neptunium is isolated using TEVA organic resin. After electro-deposition, quantification is performed by high resolution alpha spectrometry. Different analyses have been performed with blank solutions spiked with 236Pu, 237Np and 244Cm, solutions resulting from the total dissolution of isolated radioactive particles and soil samples. The lack of an appropriate tracer in our lab led to the determination of an average percentage of Np recovery using a certified solution of 237Np. Decontamination percentages obtained during the Pu-Np separation ranged from 98 % to 100 %. Some tests to investigate the effect of the addition or absence of NaNO2 (responsible for

  1. Toxicty of thorium cycle nuclides

    The purpose of this project is to investigate the biological hazards associated with uranium-thorium breeder fuels and fuel recycle process solutions. Initial studies emphasize the metabolism and long-term biological effects of inhaled 233U-232U nitrate and oxide fuel materials andof 231Pa, a major, long-lived, radioactive waste product

  2. The Tautavel cave and dating methods

    Main age estimation methods are described, datable materials and limits of application are given in each case. Methods studied are: carbon 14, 230Th/234U, 231Pa/235U, potassium-argon, thermoluminescence, electron spin resonance, fission tracks. Indirect methods like paleomagnetism and paleoclimates are briefly described

  3. Fission fragment angular distributions and fission cross section validation

    The present knowledge of angular distributions of neutron-induced fission is limited to a maximal energy of 15 MeV, with large discrepancies around 14 MeV. Only 238U and 232Th have been investigated up to 100 MeV in a single experiment. The n-TOF Collaboration performed the fission cross section measurement of several actinides (232Th, 235U, 238U, 234U, 237Np) at the n-TOF facility using an experimental set-up made of Parallel Plate Avalanche Counters (PPAC), extending the energy domain of the incident neutron above hundreds of MeV. The method based on the detection of the 2 fragments in coincidence allowed to clearly disentangle the fission reactions among other types of reactions occurring in the spallation domain. I will show the methods we used to reconstruct the full angular resolution by the tracking of fission fragments. Below 10 MeV our results are consistent with existing data. For example in the case of 232Th, below 10 MeV the results show clearly the variation occurring at the first (1 MeV) and second (7 MeV) chance fission, corresponding to transition states of given J and K (total spin and its projection on the fission axis), and a much more accurate energy dependence at the 3. chance threshold (14 MeV) has been obtained. In the spallation domain, above 30 MeV we confirm the high anisotropy revealed in 232Th by the single existing data set. I'll discuss the implications of this finding, related to the low anisotropy exhibited in proton-induced fission. I also explore the critical experiments which is valuable checks of nuclear data. The 237Np neutron-induced fission cross section has recently been 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 n-TOF data, we simulate a criticality experiment performed at Los Alamos with a 6 kg sphere of 237Np. This sphere was

  4. Partial neutron capture cross sections of actinides using cold neutron prompt gamma activation analysis

    Nuclear waste needs to be characterized for its safe handling and storage. In particular long-lived actinides render the waste characterization challenging. The results described in this thesis demonstrate that Prompt Gamma Neutron Activation Analysis (PGAA) with cold neutrons is a reliable tool for the non-destructive analysis of actinides. Nuclear data required for an accurate identification and quantification of actinides was acquired. Therefore, a sample design suitable for accurate and precise measurements of prompt γ-ray energies and partial cross sections of long-lived actinides at existing PGAA facilities was presented. Using the developed sample design the fundamental prompt γ-ray data on 237Np, 241Am and 242Pu were measured. The data were validated by repetitive analysis of different samples at two individual irradiation and counting facilities - the BRR in Budapest and the FRM II in Garching near Munich. Employing cold neutrons, resonance neutron capture by low energetic resonances was avoided during the experiments. This is an improvement over older neutron activation based works at thermal reactor neutron energies. 152 prompt γ-rays of 237Np were identified, as well as 19 of 241Am, and 127 prompt γ-rays of 242Pu. In all cases, both high and lower energetic prompt γ-rays were identified. The most intense line of 237Np was observed at an energy of Eγ=182.82(10) keV associated with a partial capture cross section of σγ=22.06(39) b. The most intense prompt γ-ray lines of 241Am and of 242Pu were observed at Eγ=154.72(7) keV with σγ=72.80(252) b and Eγ=287.69(8) keV with σγ=7.07(12) b, respectively. The measurements described in this thesis provide the first reported quantifications on partial radiative capture cross sections for 237Np, 241Am and 242Pu measured simultaneously over the large energy range from 45 keV to 12 MeV. Detailed uncertainty assessments were performed and the validity of the given uncertainties was demonstrated. Compared

  5. Study of extraction chromatography methods of separation of neptunium and its radiometric determination

    In presented dissertation thesis the separation procedure for determination of neptunium in model solutions and also in soil samples was designed. 1. The sorbent was prepared on basis of hydrophobised silica gel with silica oil for hydrophobisation - Lukooil H. Extraction agent was anchored on this modified surface -quarternary ammonium salt Aliquat 336. In experiments was proved, that sorbent prepared like this is resist enough to the elution solutions using in separation procedures. It was found out that maximal concentration of deposited Aliquat 336 in benzene on hydrophobised silica gel selected granulate size is 40%. Sorbent prepared with silica gel lower granulate size and also with higher concentration of Aliquat 336 in benzene became sticky and unsuitable for using in separation columns. 2. The prepared sorbent is similar to commercial produced TEVA resin made by company Eichrom and for purpose of this project it was cheaper but the same effective alternative in comparing to mentioned commercial resin. 3. The separation procedure for isolation of 237Np from radionuclides interfering during its radiometric determination with alpha spectrometry was designed -in scientific publications hasn't been described yet. The procedure consists of using formic acid for reduction of radionuclides interfering by alpha spectrometric determination of 237Np and subsequent purification of neptunium on the next column by using ferrous sulphamate as reduction agent. This procedure was applied to the model solutions and also to the soil samples. The model solutions and soil samples for these experiments were contaminated with known amounts of 237Np, 238pU, 232Th and mixture of 234U/238U. The effective separation of neptunium from described radionuclides was achieved with application of the described separation procedure. 4. The separation procedure for separation of 239Np from its mother radionuclide 241Am was designed with prepared sorbent. Different washing and elution

  6. ZZ FSXJ32, MCNP nuclear data library based on JENDL-3.2. ZZ FSXLIBJ33, MCNP nuclear data library based on JENDL-3.3

    -103, Rh-105, Pd-102, Pd-104, Pd-105, Pd-106, Pd-107, Pd-108, Pd-110, Ag-107, Ag-109, Ag-110m,Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-116, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-126, Sb-121, Sb-123, Sb-124, Sb-125, Te-120, Te-122, Te-123, Te-124, Te-125, Te-126, Te-127m,Te-128, Te-129m,Te-130, I -127, I -129, I -131, Xe-124, Xe-126, Xe-128, Xe-129, Xe-130, Xe-131, Xe-132, Xe-133, Xe-134, Xe-135, Xe-136, Cs-133, Cs-134, Cs-135, Cs-136, Cs-137, Ba-130, Ba-132, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Ba-140, La-138, La-139, Ce-140, Ce-141, Ce-142, Ce-144, Pr-141, Pr-143, Nd-142, Nd-143, Nd-144, Nd-145, Nd-146, Nd-147, Nd-148, Nd-150, Pm-147, Pm-148, Pm-148m,Pm-149, Sm-144, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Sm-153, Sm-154, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Eu-156, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Tb-159, Er-162, Er-164, Er-166, Er-167, Er-168, Er-170, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, W-182, W-183, W-184, W-186, Hg-196, Hg-198, Hg-199, Hg-200, Hg-201, Hg-202, Hg-204, Pb-204, Pb-206, Pb-207, Pb-208, Bi-209, Ra-223, Ra-224, Ra-225, Ra-226, Ac-225, Ac-226, Ac-227, Th-227, Th-228, Th-229, Th-230, Th-232, Th-233, Th-234, Pa-231, Pa-232, Pa-233, U-232, U-233, U-234, U-235, U-236, U-237, U-238, Np-235, Np-236, Np-237, Np-238, Np-239, Pu-236, Pu-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Pu-244, Pu-246, Am-241, Am-242, Am-242m, Am-243, Am-244, Am-244m, Cm-240, Cm-241, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246, Cm-247, Cm-248, Cm-249, Cm-250, Bk-247, Bk-249, Bk-250, Cf-249, Cf-250, Cf-251, Cf-252, Cf-254, Es-254, Es-255, Fm-255 Temperatures: 300 K. Origin: JENDL-3.3. Thermal scattering: Free gas model Kerma factors are provided. The original JENDL-3.3 has two problems in Am-241 data. One is the missing of MF/MT=4/18, and the other is the incorrect neutron spectra for MT=18 below 500 keV. The updated data have been produced as JENDL-3

  7. ZZ VITJEF22.BOLIB, JEF-2.2 Multigroup Coupled (199 n + 42 gamma) X-Section Library in AMPX Format for Nuclear Fission Applications

    1 - Description or function: VITJEF22.BOLIB /1/ is a multigroup coupled (199 neutron groups + 42 photon groups) pseudo-problem-independent cross section library in AMPX /2/ format for nuclear fission applications. VITJEF22.BOLIB is based on the JEF-2.2 /3/ European nuclear data file and it was processed, through the NJOY /4/ and SCAMPI /5/ systems, in the VITAMIN-B6 /6/ group structure using the same parameters and calculation procedures. Original nuclear data file: JEF-2.2. Data processing systems: NJOY-94.66 and SCAMPI. Format: AMPX. Number of groups: 199 neutron groups and 42 photon groups. Thermal neutron groups: 36 groups below 5.043 eV with up-scattering cross sections. Neutron energy range: 1.0 E-5 eV - 19.64 MeV. Photon energy range: 1.0 keV - 30.0 MeV. Temperatures [K]: 300, 600, 1000, and 2100 (same values as in VITAMIN-B6). Background cross sections (SIGMA-Zeros) [barns]: 1, 10, 50, 100, 300, 1.0 E+3, 1.0 E+4, 1.0 E+5, 1.0 E+6, 1.0 E+10 (infinite dilution) (same values as in VITAMIN-B6). Legendre order: P7 for materials with Z ≤ 29 (copper); P5 for the remainders. Number of materials: 138. Materials included: one file per material (nat=natural): H-H2O, H-CH2, H2-D2O, H-3, He-3, He-4, Li-6, Li-7, Be-9, Be-TH, B-10, B-11, C-nat, C-GPH, N-14, N-15, O-16, O-17, F-19, Na-23, Mg-nat, Al-27, Si-nat, P-31, S-32, S-33, S-34, S-36, Cl-nat, K-nat, Ca-nat, Ti-nat, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-nat, Ga-nat, Y-89, Zr-nat, Nb-93, Mo-nat, Ag-107, Ag-109, Cd-nat, Cd-106, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115m, Cd-116, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-125, Sn-126, Ba-138, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-nat, Bi-209, Th-230, Th-232, Pa-231, Pa-233, U-232, U-233, U-234, U-235, U-236

  8. Determination of 236Np(n,f) cross-section using surrogate ratio method

    Surrogate strategy is widely used to determine (n,f) cross-sections. Hybrid ratio approach is used to get 236Np(n,f) cross-section for which there is no experimental data available beyond 7 MeV till date. 237Np and 239Pu are produced over the similar excitation energy range (20.4-25.4 MeV) using 235U(6Li,α)237Np and 235U(6Li,d)239Pu reactions respectively. The ground state Q values for these two reactions are 7.707 MeV and - 6.718 respectively. Hence 237Np and 239Pu can be populated at overlapping excitation energies in 6Li + 235U transfer reactions for bombarding energies around 44.4 MeV. Fission probabilities of these two compound nuclei are measured and the desired cross-sections are determined using the following formula based on Weisskopf-Ewing approximation. Measurements were carried out using 44.4 MeV 6Li beam (from BARC-TIFR Pelletron) and 1.6 mg/cm2 thick 235U target (with 4.5 mg/cm2 thick Ni-Cu backing). Four telescopes (ΔE-E) made of silicon surface barrier detectors, used to detect light charged particles, were kept 100 apart covering 700-1000. A large area silicon detector (with a solid angle ∼ 33 msr and an angular coverage of 1540- 1660) was used to detect fission fragments. Two monitor detectors were placed at forward angles. Results for 236Np(n,f) cross sections using the yields from the forward most telescope are shown. Contributions from Ni-Cu backing in inclusive counts have been subtracted out. The experimental results are consistent with the EMPIRE3.1 calculations

  9. Radionuclide Transport in Fracture-Granite Interface Zones

    Hu, Q; Mori, A

    2007-09-12

    In situ radionuclide migration experiments, followed by excavation and sample characterization, were conducted in a water-conducting shear zone at the Grimsel Test Site (GTS) in Switzerland to study diffusion paths of radionuclides in fractured granite. In this work, we employed a micro-scale mapping technique that interfaces laser ablation sampling with inductively coupled plasma-mass spectrometry (LA/ICP-MS) to measure the fine-scale (micron-range) distribution of actinides ({sup 234}U, {sup 235}U, and {sup 237}Np) in the fracture-granite interface zones. Long-lived {sup 234}U, {sup 235}U, and {sup 237}Np were detected in flow channels, as well as in the adjacent rock matrix, using the sensitive, feature-based mapping of the LA/ICP-MS technique. The injected sorbing actinides are mainly located within the advective flowing fractures and the immediately adjacent regions. The water-conducting fracture studied in this work is bounded on one side by mylonite and the other by granitic matrix regions. These actinides did not penetrate into the mylonite side as much as the relatively higher-porosity granite matrix, most likely due to the low porosity, hydraulic conductivity, and diffusivity of the fracture wall (a thickness of about 0.4 mm separates the mylonite region from the fracture) and the mylonite region itself. Overall, the maximum penetration depth detected with this technique for the more diffusive {sup 237}Np over the field experimental time scale of about 60 days was about 10 mm in the granitic matrix, illustrating the importance of matrix diffusion in retarding radionuclide transport from the advective fractures. Laboratory tests and numerical modeling of radionuclide diffusion into granitic matrix was conducted to complement and help interpret the field results. Measured apparent diffusivity of multiple tracers in granite provided consistent predictions for radionuclide transport in the fractured granitic rock.

  10. Distribution of Np and Pu in Swedish lichen samples (Cladonia stellaris) contaminated by atmospheric fallout

    Lindahl, Patric E-mail: patric.lindahl@radfys.lu.se; Roos, Per; Eriksson, Mats; Holm, Elis

    2004-07-01

    The activity concentrations of {sup 237}Np and the two Pu isotopes, {sup 239}Pu and {sup 240}Pu, were determined in lichen samples (Cladonia stellaris) contaminated by fallout from atmospheric nuclear test explosions and the Chernobyl accident. The samples were collected at 18 locations in Sweden, from north to south, between 1986 and 1988 and analysed with high-resolution inductively coupled plasma mass spectrometry (HR-ICP-MS) and alpha spectrometry. Data on the activity ratios {sup 238}Pu/{sup 239+240}Pu and {sup 134}Cs/{sup 137}Cs measured previously were also included in this study for comparison. The {sup 237}Np activity concentration ranged from 0.08{+-}0.01 to 2.08{+-}0.17 mBq kg{sup -1}, depending on the location of the sampling site and time of collection. The {sup 239+240}Pu activity concentration ranged from 0.09{+-}0.01 to 4.09{+-}0.15 Bq kg{sup -1}, with the {sup 240}Pu/{sup 239}Pu atomic ratio ranging between 0.16{+-}0.01 and 0.44{+-}0.03, the higher ratios indicating a combination of weapons test fallout and Chernobyl fallout. The {sup 237}Np/{sup 239}Pu atomic ratios ranged between 0.06{+-}0.01 and 0.42{+-}0.04, the lower ratios indicating combination of weapons test fallout and Chernobyl fallout. At a well-defined sampling site at Lake Rogen (62.32 deg. N, 12.38 deg. E), additional lichen samples were collected between 1987 and 1998 to study the distribution of Np and Pu in different layers. The concentrations of the two elements follow each other quite well in the profile.

  11. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

  12. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focuses on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by (8.7±2.1)% and (6.5±2.1)% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be under-predicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4)%. (authors)

  13. Modeling of neptunium(V), plutonium(IV) and americium(III) sorption on soils in the presence of humic acid

    Sorption experiments of 237Np(V)O2+, 238Pu(IV)4+ and 241Am(III)3+ onto soils in the presence of humic acid have been performed by a batch system, in order to clarify effects of humic acid on sorption of the radionuclides on soils. Soils used in the present experiments were a coastal sand which does not sorb humic acid, and an ando soil which sorbs humic acid very well and is known to have a high content of humic substances. The distribution coefficient of 237Np for both soils was not affected by the presence of humic acid, since 237Np much little interact with humic acid. The distribution coefficient of 241Am for both soils decreased as the humic acid concentration increased. Also the distribution coefficient of 238Pu for the coastal sand decreased with increasing humic acid concentration. On the other hand, as to the ando soil, the distribution coefficient of 238Pu in the presence of humic acid was larger than that in the absence of humic acid, in the humic acid concentration range below 5 mg/dm3, although it decreased with increasing humic acid concentration over 5 mg/dm3. These results suggest that apparent sorption behavior of 238Pu and 241Am on the soils may be dependent on the sorption ability of their humic complexes. The distribution coefficient of the three radionuclides for the soils could be evaluated by the sorption equilibrium model taking account of the effects of the sorption of both humic acid itself and humic complexes of radionuclide on the value of the distribution coefficient, besides the complexation in aqueous phase. (author)

  14. Marine Biological Investigations at the Eniwetok Test Site

    The results of marine biological investigations conducted at the Eniwetok Test Site since 1952 are summarized. Radioisotopes introduced into the sea from the tests at various times since then include fission products and other radioisotopes (U237, Np239, Mn54, Fe55,59, Co57,58,60, Zn65 and W185). The levels of radioisotopes in plankton samples taken 4 days to 6 weeks after contamination are reported and the distribution of the radioactivity between plankton and water is given. Grazing fishes contained Zn65, Fe55, Co57,58,60 and Mn54. Carnivorous fishes contained mostly Fe55 and Zn65. (author)

  15. Fission fragment angular distribution in heavy ion induced fission

    S. Soheyli

    2006-06-01

    Full Text Available   We have calculated the fission fragment angular anisotropy for 16O + 232Th,12C + 236U , 11B + 237 Np , 14 N + 232 Th , 11B + 235U , 12C + 232Th systems with the saddle point statistical model and compared the fission fragment angular anisotropy for these systems. This comparison was done with two methods a without neutron correction and b with neutron correction. Also we studied normal and anomalous behavior of the fission fragment angular anisotropy. Finally, we have predicted the average emitted neutron from compound nuclei considering the best fit for each system.

  16. Modelling of reaction cross sections and prompt neutron emission

    Oberstedt S.

    2010-10-01

    Full Text Available Accurate nuclear data concerning reaction cross sections and the emission of prompt fission neutrons (i.e. multiplicity and spectra as well as other fission fragment data are of great importance for reactor physics design, especially for the new Generation IV nuclear energy systems. During the past years for several actinides (238U(n, f and 237Np(n, f both the reaction cross sections and prompt neutron multiplicities and spectra have been calculated within the frame of the EFNUDAT project.

  17. Fission fragment angular distribution in heavy ion induced fission

    S. Soheyli; I. Ziaeian

    2006-01-01

      We have calculated the fission fragment angular anisotropy for 16O + 232Th,12C + 236U , 11B + 237 Np , 14 N + 232 Th , 11B + 235U , 12C + 232Th systems with the saddle point statistical model and compared the fission fragment angular anisotropy for these systems. This comparison was done with two methods a) without neutron correction and b) with neutron correction. Also we studied normal and anomalous behavior of the fission fragment angular anisotropy. Finally, we have predicted the averag...

  18. Photonuclear measurements on fissionable isotopes using monoenergetic photons

    The LLL linac monochromatic photon facility is used for measurements, which to date have included at least some data on the eight isotopes: 232Th, 233U, 234U, 235U, 236U, 237Np, 238U, and 239Pu. Photofission events are determined from an analysis of the emitted neutron multiplicity distributions as determined with the LLL 4π neutron detector. A sampling of data taken to date is included and is organized as follows: low-energy photofission cross sections, neutrino and fission neutron multiplicity distribution width parameter data, ring-ratio multiplicity coincidence data, delayed-neutron fraction data, and photonuclear and photofission cross sections. 2 refs., 32 figs

  19. Determination of 236Np(n,f) and 238Pu(n,f) cross-sections using surrogate reactions

    Cross sectional data for 236Np(n,f) reaction is not available for neutron energies beyond 4.32 MeV. In this work, we have determined the cross-section of the above reaction for the neutron energy range En=10-18 MeV using ‘Hybrid Surrogate Ratio’ method via 235U(6Li, α)237Np and 235U(6Li,d)239Pu reactions. The cross-sections for 238Pu(n,f) reaction available in literature have been used as a reference for the above surrogate method

  20. Neutron-induced fission cross section of 240,242Pu up to En = 3 MeV

    SALVADOR CASTINEIRA PAULA; BRYS TOMASZ; Hambsch, Franz-Josef; Oberstedt, Stephan; Pretel, C.; Vidali, Marzio

    2014-01-01

    The neutron-induced fission cross sections of 240,242Pu have been measured at JRC-IRMM with incident neutron energy from 0.2 MeV up to 3 MeV. A Twin-Frisch Grid Ionization Chamber (TFGIC) has been used in a back-to-back geometry. The measurements have been performed using the secondary standards 237Np and 238U as a reference. The purity of the plutonium samples was 99.89% for 240Pu and 99.97% for 242Pu. The results obtained follow the ENDF/B-VII.1 evaluation for 240Pu, but some discrepancies ...

  1. On the role of energy separated in fission process, excitation energy and reaction channels effects in the isomeric ratios of fission product 135Xe in photofission of actinide elements

    Thiep, Tran Duc; An, Truong Thi; Cuong, Phan Viet; Vinh, Nguyen The; Mishinski, G. V.; Zhemenik, V. I.

    2016-07-01

    In this work we present the isomeric ratio of fission product 135Xe in the photo-fission of actinide elements 232Th, 233U and 237Np induced by end-point bremsstrahlung energies of 13.5, 23.5 and 25.0 MeV which were determined by the method of inert gaseous flow. The data were analyzed, discussed and compared with the similar data from literature to examine the role of energy separated in fission process, excitation energy and reaction channels effects.

  2. Neutron-induced fission cross section of 240,242Pu

    Salvador Castiñeira, Paula

    2014-01-01

    A recent sensitivity analysis done for the new generation of fast reactors [1] has shown the importance of improved cross section data for several actinides. Among them, the neutron-induced fission cross section of 240,242Pu requires a level of accuracy of 1-3% and 3-5%, respectively, from the current status of 6% and 20%. Moreover, nearly all the measurements in the literature have been done relative to 235U(n,f). Therefore, using other references samples such as 237Np or 238U will provide t...

  3. Effects of the shape of tracer source on 90Sr migration

    The method and the results of the tracer migration test in horizontal column, which was conducted in the experimental shaft of the CIRP's Field Test Site, are introduced. The effects of the shape of tracer source on 90Sr migration are discussed. The results indicates that adsorption capacity of loess to 90Sr, 237Np, 238Pu and 241Am are relatively strong. Humus retarded migration of 90Sr. Retardation coefficient of 90Sr, under the condition of plane source, is less than that under the condition of point source. (authors)

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

    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 of...

  5. Metabolism and toxicity of neptunium

    The biological behavior and toxicity of neptunium were studied. Neptunium was admimistered either intravenously or intramuscularly in rats. In contrast of other transuranium elements the distribution patterns of neptunium in case of intravenous injection is not dependent of the physico-chemical state. The urinary excretion is high. The distribution after intramuscular injection showed a rather fast migration from the injection site. 237Np excreted in urine was approximately equal to bone deposit. Neptunium behavior rather followed that of alkaline earths than that of transplutonium elements

  6. Activation Spectrometry of Fast Neutrons by IAEA Threshold Detectors at Neutron Generators

    The suitability of the IAEA set of threshold detectors for neutron accident purposes was investigated. A generator producing 14.3-MeV neutrons by the T(d, n)4He reaction was employed for this purpose. 237Np, 232Th, 58Ni and 27Al threshold detectors were used. The induced activity was determined by gamma spectrometry using a multichannel analyser. Fast neutron spectra have been estimated from the experimental results. Measurements at the surface and at the depth of a phantom were provided. Some difficulties from low induced and fission activities (caused by the small neutron flux density and the light weight of the detectors) are pointed out. (author)

  7. Sorption of Np (V) by U (VI) hydroxide solids

    Wruck, D A; Brachmann, A; Sylwester, E; Allen, C E A

    1999-09-20

    The distribution of {sup 237}Np(V) between aqueous NaHCO{sub 3} solutions and U(VI) hydroxide solids was investigated. Experiments were initiated by addition of U solids to Np solutions and by coprecipitation of U and Np. Analysis by U L{sub III} extended X-ray absorption fine structure (EXAFS) spectroscopy and infrared absorption spectroscopy indicated the solid phase was synthetic schoepite. Equilibrium Np distribution coefficients were 5-44 mL/g in the pH range 6-8. The results are consistent with adsorption of Np by the solids and provide no evidence of Np incorporation in the bulk solid.

  8. EDB-II validated, key fission product yields for fast reactor application

    Relative fission yields were measured for three different locations in the row 4 ''Test Region'' of the EBR-II reactor. Correlation of the relative fission yields to the measured average energy (anti E) and the measured 137Cs 238U/235U spectral indices have been made. The measured relative fission yields for selected fission products from 235U, 238U, 239Pu and 237Np have been compared with those values reported by the Interlaboratory Reaction Rate (ILRR) program, EBR-II fast reactor yields from destructive analysis and summation, and the March 1977 version of ENDF/B-V

  9. Simultaneous neutron and gamma spectrum adjustment

    The spectrum adjustment procedure was extended to simultaneous neutron and gamma spectrum adjustment, and the feasibility of this technique is demonstrated in the analysis of HFIR dosimetry experiments. Conditions in which gamma rays may contribute considerably to radiation damage in steels are discussed. Beryllium helium accumulation fluence monitors (HAFMs) were found to be good monitors in gamma fields of intensities high enough to contribute to steel embrittlement. Use of 237Np, 238U, and 9Be HAFM as gamma dosimeters is proposed for high-dose irradiations in high-energy, high-intensity gamma fields

  10. Application of inductively coupled plasma mass spectrometry to the measurement of long-lived radionuclides in environmental samples

    This review describes applications of inductively coupled plasma mass spectrometry (ICP-MS) to the determination of long-lived radionuclides in environmental samples. Simultaneous determination of 232Th and 238U in biological samples is described in detail; in this procedure an internal standard, Tl or Bi, is adopted for correction of the matrix effect. Determination of 237Np in soil samples by ICP-MS is also described. It is chemically separated to ensure no interference from matrix elements. The detection limits are several mBq (several pg) for the case of radionuclides having a half life of thirty or forty thousand years. (author)

  11. Alpha self-irradiation effects in ternary oxides of actinides elements: The zircon-like phases AmIIIVO4 and AIINpIV(VO4)2 (A=Sr, Pb)

    We report the experimental studies of irradiation damage from alpha decay in neptunium and americium vanadates versus cumulative dose. The isotopes used were the transuranium α-emitter 237Np and the α,γ-emitter 241Am. Neptunium and americium vanadates self-irradiation was studied by X-ray diffraction method (XRD). The comparison of the powder diffraction patterns reveal that the irradiation has no apparent effect on the neptunium phases while the americium vanadate swells and becomes metamict as a function of cumulative dose

  12. Calculation of waste disposal rating for the fusion experimental breeder FEB-E

    Using the neutron transport code BISON 3.0, activation calculation code FDKR and its associated data libraries, the activation calculation and analyses of all long-lived radionuclides are performed for the Fusion Experimental Breeder FEB-E. The results indicate that the first wall and blanket structure materials of the FEB-E can meet the nuclear waste disposal criteria for the NRC 10CFR61 Class C after a few weeks from shutdown. The inventory of important actinides during field reprocessing, such as 232U and 237Np, are also calculated. Their concentrations do not excess the limit value required for environmental safety

  13. Comparison of fission and capture cross sections of minor actinides

    Nakagawa, T

    2003-01-01

    The fission and capture cross sections of minor actinides given in JENDL-3.3 are compared with other evaluated data and experimental data. The comparison was made for 32 nuclides of Th-227, 228, 229, 230, 233, 234, Pa-231, 232, 233, U-232, 234, 236, 237, Np-236, 237, 238, Pu-236, 237, 238, 242, 244, Am-241, 242, 242m, 243, Cm-242, 243, 244, 245, 246, 247 and 248. Given in the present report are figures of these cross sections and tables of cross sections at 0.0253 eV and resonance integrals.

  14. INERT-MATRIX FUEL: ACTINIDE ''BURNING'' AND DIRECT DISPOSAL

    Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am, 244 Cm and 237 Np). In Europe, Canada and Japan studies have determined much improved efficiencies for burnup of actinides using inert-matrix fuels. This innovative approach also considers the properties of the inert-matrix fuel as a nuclear waste form for direct disposal after one-cycle of burn-up. Direct disposal can considerably reduce cost, processing requirements, and radiation exposure to workers

  15. Preparation and characterisation of 234U tracer for mass spectrometry and alpha spectrometry

    234U was milked from 15 years aged 238Pu prepared earlier by neutron irradiation of 237Np. Ion exchange procedure using Dowex 1 X 8 resin in the nitric acid medium was followed for this purpose, in a glove box. The purified 234U was characterised by alpha spectrometry and thermal ionisation mass spectrometry for its 238Pu content and the isotopic composition of uranium, respectively. Alpha activity ratio of 234U/238Pu was 0.015 and the abundance of 234U was about 99 atom percent. (author). 1 fig

  16. Comparison of fission and capture cross sections of minor actinides

    The fission and capture cross sections of minor actinides given in JENDL-3.3 are compared with other evaluated data and experimental data. The comparison was made for 32 nuclides of Th-227, 228, 229, 230, 233, 234, Pa-231, 232, 233, U-232, 234, 236, 237, Np-236, 237, 238, Pu-236, 237, 238, 242, 244, Am-241, 242, 242m, 243, Cm-242, 243, 244, 245, 246, 247 and 248. Given in the present report are figures of these cross sections and tables of cross sections at 0.0253 eV and resonance integrals. (author)

  17. Speciation of Long-Lived Radionuclides in the Environment

    Hou, Xiaolin

    , isotopes of Pu, and 237Np in seawater, fresh water, soil, sediment, vegetations, and concrete. The developed methods are used for the investigation of the chemical speciation of these radionuclides as well as their environmental behaviours, especially in Danish environment. In addition the speciation of Pu......This project started in November 2005 and ended in November 2008, the work and research approaches are summarized in this report. This project studied the speciation of radionuclides in environment. A number of speciation analytical methods are developed for determination of species of 129I, 99Tc...

  18. The nuclear excitation by electron transition (NEET) and its application to isotope separation

    The electron transition in atoms can excite nuclei under the particular condition, and the possibility of its application to isotope separation has been studied. The theory and the experiment are shown in this paper. The NEET of 189Os was first confirmed at Osaka University, and that of 237Np was carried out in succession. Though the application to isotope separation has not been established so far, the separation of 235U from 238U is very interesting. The possibility of the above isotope separation is discussed. (Yoshimori, M.)

  19. Systematics of fission cross sections at the intermediate energy region

    Fukahori, Tokio; Chiba, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    The systematics was obtained with fitting experimental data for proton induced fission cross sections of Ag, {sup 181}Ta, {sup 197}Au, {sup 206,207,208}Pb, {sup 209}Bi, {sup 232}Th, {sup 233,235,238}U, {sup 237}Np and {sup 239}Pu above 20 MeV. The low energy cross section of actinoid nuclei is omitted from systematics study, since the cross section has a complicated shape and strongly depends on characteristic of nucleus. The fission cross sections calculated by the systematics are in good agreement with experimental data. (author)

  20. Challenge in determination of long-lived radionuclides by ICP-MS

    The ultra-trace and isotope analysis of long-lived radionuclides in environmental materials (in biological or geological samples and waters) is relevant of increasing importance [1-5].E.g., the determination of long-lived radionuclides is for the detection of radionuclide contamination in environmental materials in which several radioactive nuclides (e.g.99Tc, 129I , 237Np, 239Pu, 240Pu, 241Am) are present from fallout due to nuclear weapons testing, nuclear power plants or nuclear accidents. Especially, isotope ratios of uranium and plutonium [6] can indicate the origin of contamination in the environmental samples

  1. HEU age determination

    A technique has been developed to determine the Highly Enriched Uranium (HEU) Age which is defined as the time since the HEU was produced in an enrichment process. The HEU age is determined from the ratios of relevant uranium parents and their daughters viz 230Th/234U and 231Pa/235U. Uranium isotopes are quantitatively measured by their characteristic gammas and their daughters by alpha spectroscopy. In some of the samples where HEU is enriched more than 99%, the only mode of HEU age determination is by the measurement of 231Pa since there is negligible quantity of 230Th due to very low atom concentrations of 234U in the sample. In this paper we have presented data and methodology of finding the age of two HEU samples

  2. Study for 228Th reduction in thermal reactor with Th-U fuel cycls

    1999-01-01

    By using computercode WIMS/CENDL, the effects of some parameters, core configuration such as fuel element structure, neutron flux and burn-up, are discussed in thispaper.It is shown that high neutron flux, small fuel rod diameter,large volume ratio of coolant to fuel, seed-blank heterogeneous corearrangement and 231Pa chemical separation are necessary for reducing 228Th production in reactor.

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

    A detailed review is made of the experimental techniques that are available, or are in the process of development, for the determination of 238U, 235U, 234U, 231Pa, 232Th, 230Th, 228Th, 228Ra, 226Ra, 223Ra, 210Po and 210Pb. 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

  4. Mound Facility activities in chemical and physical research: July-December 1979

    Research is reported in the following fields: isotope separation (Ar, C, He, Kr, Ne, O, Xe), low-temperature research (H intermolecular potential functions, gas analysis in trennschaukel), separation chemistry (229Th, 231Pa, 230Th, 234U), separation research (liquid thermal diffusion, Ca isotope separation, molecular beam scattering, mutual diffusion of noble gas mixtures, lithium chemical exchange with cryptands), and calculations in plutonium chemistry (algorithms, valence in natural water)

  5. Measurements of radium in water using impregnated fibers

    The technique perfected by Moore and Reid for sampling radium in seawater is well adapted for environmental sampling. Using this method, we have examined runoff from mine tailings and have observed relatively high amounts of 223Ra (from the 235U series). Apparently the fiber is able to absorb a precursor, 231Pa or 227Ac, and hence retains the 223Ra concentrations for long storage periods. Examples of high-resolution alpha spectrometry of these activities are presented

  6. Mound Facility activities in chemical and physical research: July-December 1979

    1980-06-18

    Research is reported in the following fields: isotope separation (Ar, C, He, Kr, Ne, O, Xe), low-temperature research (H intermolecular potential functions, gas analysis in trennschaukel), separation chemistry (/sup 229/Th, /sup 231/Pa, /sup 230/Th, /sup 234/U), separation research (liquid thermal diffusion, Ca isotope separation, molecular beam scattering, mutual diffusion of noble gas mixtures, lithium chemical exchange with cryptands), and calculations in plutonium chemistry (algorithms, valence in natural water). (DLC)

  7. Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

    Trianti, N.; Su'ud, Z.; Riyana, E. S.

    2012-06-01

    A preliminary design study for the utilization of thorium added with 231Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of 233U to 231Pa in burn-up process. Optimizations of the content of 231Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 ˜ 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

  8. Determination of the 233Pa(n, f) reaction cross section from 0.5 to 10 MeV neutron energy using the transfer reaction 232Th(3He, p)234Pa

    The fission probability distributions of 232,233,234Pa and 231Th have been measured up to an excitation energy of 15 MeV, using the transfer reactions 232Th(3He, t)232Pa, 232Th(3He, d)233Pa, 232Th(3He, p)234Pa and 232Th(3He, 4He)231Th. From these measurements, the neutron induced fission cross sections of 231Pa, 233Pa and 230Th have been determined from the product of the fission probabilities of 232Pa, 233Pa and 231Th respectively with the calculated compound nucleus formation cross sections in the 231Pa+n, 233Pa+n and 230Th+n reactions. The validity of the applied method has been successfully tested with the existing neutron induced fission cross sections of 230Th and 231Pa. Special emphasis is put on the 233Pa(n, f) reaction which is of importance for thorium fueled nuclear reactors. Based on a statistical model analysis of the neutron induced fission cross section as a function of neutron energy, it has been possible to determine the barrier parameters of the 234Pa fissioning nucleus. Cross sections for the compound nucleus inelastic scattering 233Pa(n, n') and radiative capture 233Pa(n, γ) reactions have also been calculated and compared with recent evaluations

  9. Proton-induced fission of heavy nuclei at intermediate energies

    Deppman, A; Guimaraes, V; Karapetyan, G S; Balabekyan, A R; Demekhina, N A

    2013-01-01

    The intermediate energy proton-induced fission of 241Am, 238$U and 237$Np is studied. The inelastic interactions of protons and heavy nuclei are described by a CRISP model, in which the reaction proceeds in two steps. The first one corresponds fast cascade, where a series of individual particle-particle collisions occurs within the nucleus. It leaves a highly excited cascade residual nucleus, assumed to be in thermal equilibrium. Subsequently, in the second step the excited nucleus releases its energy by evaporation of neutrons and light charged particles as well. Both the symmetric and asymmetric fission are regarded, and the fission probabilities are obtained from CRISP code calculations, by means of statistical weighting factors. The fission cross sections, the fissility of the fissioning nuclei, and the number of nucleons lost by the target - before and after fission - are calculated and compared to experiments for 660 MeV protons incident on 241Am, 238$U and 237$Np. Some of the model predictions are in f...

  10. EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2

    The neutronics and the transmutation properties of the IAEA ADS benchmark 3.2 setup, the 'Yalina' experiment or ISTC project B-70, have been studied through an extensive amount of 3-D Monte Carlo calculations at CERN. The simulations were performed with the state-of-the-art computer code package EA-MC, developed at CERN. The calculational approach is outlined and the results are presented in accordance with the guidelines given in the benchmark description. A variety of experimental conditions and parameters are examined; three different fuel rod configurations and three types of neutron sources are applied to the system. Reactivity change effects introduced by removal of fuel rods in both central and peripheral positions are also computed. Irradiation samples located in a total of 8 geometrical positions are examined. Calculations of capture reaction rates in 129I, 237Np and 243Am samples and of fission reaction rates in 235U, 237Np and 243Am samples are presented. Simulated neutron flux densities and energy spectra as well as spectral indices inside experimental channels are also given according to benchmark specifications. Two different nuclear data libraries, JAR-95 and JENDL-3.2, are applied for the calculations

  11. Capture cross section measurement of Np-237 below 1 keV with Lead Slowing-down Spectrometer

    Kobayashi, Katsuhei; Cho, Hyun-Je; Yamamoto, Shuji; Yoshimoto, Takaaki; Fujita, Yoshiaki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Ohkawachi, Yasushi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2000-03-01

    Making use of the Kyoto University Lead slowing-down Spectrometer (KULS) driven by a 46 MeV electron linear accelerator (linac) at the Research Reactor Institute, Kyoto University (KURRI), the relative cross section for the {sup 237}Np(n,{gamma}) reaction has been measured from 0.01 eV to 1 keV with energy resolution of about 40% (FWHM). The neutron flux/spectrum has been measured by a BF{sub 3} counter. The cross section of the {sup 10}B(n,{alpha}) reaction in ENDF/B-VI was used as a reference one for the cross section measurement. The measured result has been normalized to the reference value of the {sup 237}Np(n,{gamma}){sup 238}Np reaction in ENDF/B-VI at 0.0253 eV, and the measurement has been compared with the experimental and the evaluated data in ENDF/B-VI and JENDL-3.2, whose data were broadened by the energy resolution of the KULS. (author)

  12. Analytical method for the determination of Np and Pu in sea water by AMS with respect to the Fukushima accident

    Hain, K.; Faestermann, T.; Famulok, N.; Fimiani, L.; Gómez-Guzmán, J. M.; Korschinek, G.; Kortmann, F.; Lierse v. Gostomski, Ch.; Ludwig, P.; Shinonaga, T.

    2015-10-01

    A chemical separation procedure for plutonium (Pu) and neptunium (Np) was developed using extraction chromatography, mass spectrometry and radiometric analysis to determine their concentrations and isotopic ratios in sea water. 241Am, which causes isobaric background to 241Pu in mass spectrometric measurements, was successfully separated from the Pu fraction by this method. Water samples which were spiked with 242Pu and 237Np or 239Np, respectively, were used for chemical yield determination. The chemical yields of Pu and Np, which were determined by alpha and gamma spectrometry at the Radiochemie München (RCM), of more than 85% were obtained. The developed method was applied to analyze the concentration of Pu and Np in the certified reference material, IAEA-443, by Accelerator Mass Spectrometry (AMS) at the Maier-Leibnitz-Laboratory (MLL) to check the applicability of the method to sea water samples. The concentrations of 240Pu, 241Pu and 237Np obtained in this study are in agreement with the certified and literature values within the uncertainties. Due to strong isotopic interference of 239Pu with 238U, it was not possible to analyze the concentration of 239Pu. Some modifications of the chemical separation method to suppress the uranium (U) fraction are under consideration. This method can be used for the analysis of Pu and Np in Pacific Ocean water samples collected after the Fukushima accident.

  13. Moessbauer spectroscopy in neptunium compounds

    Nakamoto, Tadahiro; Nakada, Masami; Masaki, Nobuyuki; Saeki, Masakatsu [Japan Atomic Energy Research Inst., Tokyo (Japan)

    1997-03-01

    Moessbauer effects are observable in seven elements of actinides from {sup 232}Th to {sup 247}Cm and Moesbauer spectra have been investigated mainly with {sup 237}Np and {sup 238}U for the reasons of availability and cost of materials. This report describes the fundamental characteristics of Moessbauer spectra of {sup 237}Np and the correlation between the isomer shift and the coordination number of Np(V) compounds. The isomer shifts of Np(V) compounds had a tendency to increase as an increase of coordination number and the isomer shifts of Np(V) compounds showed broad distribution as well as those of Np(VI) but {delta} values of the compounds with the same coordination number were distributed in a narrow range. The {delta} values of Np(VI) complexes with O{sub x} donor set suggest that the Np atom in its hydroxide (NpO{sub 2}(OH){center_dot}4H{sub 2}O)might have pentagonal bipyramidal structure and at least, pentagonal and hexagonal bipyramidal structures might coexist in its acetate and benzoate. Really, such coexistence has been demonstrated in its nitrate, (NpO{sub 2}){sub 2}(NO{sub 3}){sub 2}{center_dot}5H{sub 2}O. (M.N.)

  14. The effects of actinide separation on the radiological consequences of geologic disposal of high-level waste

    It has often been suggested that the potential hazard to man from the disposal of high-level radioactive waste could be reduced by removing a substantial fraction of the actinide elements. In this report the effects of actinide separation on the radiological consequences of one of the disposal options currently under consideration, that of burial in deep geologic formations, are examined. The results show that the potential radiological impact of geologic disposal of high-level waste arises from both long-lived fission products and actinides (and their daughter radionuclides). Neither class of radionuclides is of overriding importance and actinide separation would therefore reduce the radiological impact to only a limited extent and over limited periods. There might be a case for attempting to reduce doses from 237Np. To achieve this it appears to be necessary to separate both neptunium and its precursor element americium. However, there are major uncertainties in the data needed to predict doses from 237Np; further research is required to resolve these uncertainties. In addition, consideration should be given to alternative methods of reducing the radiological impact of geologic disposal. The conclusions of this assessment differ considerably from those of similar studies based on the concept of toxicity indices. Use of these indices can lead to incorrect allocation of research and development effort. (author)

  15. Modelisation of the fission cross section

    The neutron cross sections of four nuclear systems (n+235U, n+233U, n+241Am and n+237Np) are studied in the present document. The target nuclei of the first case, like 235U and 239Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237Np and 241Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author)

  16. Sensitivity analysis for actinide production and depletion in fast reactors

    In sensitivity analysis of the actinide production and depletion in fast reactors, a mathematical method of calculating sensitivity coefficients is improved and simplified by combining the time-dependent generalized perturbation technique with the eigenvalue method. Numerical calculations show that the eigenvalue method is well applicable in solving the nuclide chain equation and its adjoint equation and the cylic chains in the decay scheme of the actinides can be interpreted by means of complex eigenvalues. The sensitivity coefficients of actinide production and depletion in a 1000 MWe fast reactor are strongly dependent on the type of Pu fuel used, i.e. Pu fuel from BWR or Pu fuel from the blanket of FBR. The sensitivity coefficients due to variations of capture cross sections, σsub(n,2n) of 238U, lambda sub(β) of 241Pu and lambda sub(α) of 242Cm are especially large. Sensitivity analyses for the 1000 MWe fast reactors show that higher priorily should be given to decay constants of 241Pu and 242Cm, capture cross sections of 237Np, 241Am, 243Am and 242Pu, and fission cross sections of 237Np, 242Pu, 241Am and sup(242m)Am. (author)

  17. Transuranium reference measurements and reference materials

    During the 30 years of its existence, the Central Bureau for Nuclear Measurements has been involved in numerous high-accuracy investigations of mainly nuclear properties of elements important for the nuclear fuel cycle and for application of relevant radionuclides. This paper reports that these studies were made possible by the availability of particle accelerators (150-MeV electron linear accelerator; 3.7- and 7-MV Van de Graaff accelerators) as neutron sources and the use of specialized laboratories for the preparation, characterization, and manipulation of radioactive samples. Examples of investigations with plutonium and neptunium are capture and fission cross sections of 239Pu at very low energies, spontaneous fission fragment mass and energy distributions of 242Pu at very low energies, spontaneous fission fragment mass and energy distributions of 242Pu, alpha-particle and gamma-ray spectra of 237Np, half-life of 241Pu, use of 237Np in reactor neutron dosimetry, preparation and characterization of plutonium oxide reference materials for nondestructive analysis, and demonstration of the potential of synthetic isotope mixtures for analytical control of the nuclear fuel cycle

  18. Radionuclide migration experiments in a natural fracture in a quarried block of granite

    A radionuclide migration experiment was performed over a distance of 1 m in a natural fracture in a quarried block of granite. The fracture in the block was characterized hydraulically by measuring the pressure drop in borehole-to-borehole pump tests. The effective fracture volume in the block was ∼100 mL. A silicone coating was applied to the exterior, and the block was immersed in a tank of water to which hydrazine was added to provide a chemically reducing barrier. Migration experiments were performed at a flow rate of 2.2 mL h-1 using 85Sr, 131I,137Cs, 144Ce, 152Eu, 237Np and 238Pu. A total of 9.5 L of groundwater was pumped through the fracture, corresponding to ∼95 fracture volumes. Only 85Sr, 131I, 137Cs, 237Np and 238Pu were observed in the eluent. Scanning of the fracture surface at the end of the migration experiment showed limited mobility of α-emitting radionuclides and of the rare-earth elements, consistent with static sorption data obtained on representative fracture surface material. The mobility of 137Cs was higher than that of the rare-earth elements, but it was lower than that of 85Sr. When samples of fracture-coating material were separated into fractions with different specific gravity, there was a clear indication of radionuclide association with mineral groups. (author)

  19. Radionuclide migration experiments in a natural fracture in a quarried block of granite

    Vandergraaf, Tjalle T.; Drew, Douglas J.; Masuda, Sumio

    1996-02-01

    A radionuclide migration experiment was performed over a distance of 1 m in a natural fracture in a quarried block of granite. The fracture in the block was characterized hydraulically by measuring the pressure drop in borehole-to-borehole pump tests. The effective fracture volume in the block was ˜ 100 mL. A silicone coating was applied to the exterior, and the block was immersed in a tank of water to which hydrazine was added to provide a chemically reducing barrier. Migration experiments were performed at a flow rate of 2.2 mL h -1 using 85Sr, 131I, 137Cs, 144Ce, 152Eu, 237Np and 238Pu. A total of 9.5 L of groundwater was pumped through the fracture, corresponding to ˜95 fracture volumes. Only 85Sr, 131I, 137Cs, 237Np and 238Pu were observed in the eluent. Scanning of the fracture surface at the end of the migration experiment showed limited mobility of α-emitting radionuclides and of the rare-earth elements, consistent with static sorption data obtained on representative fracture surface material. The mobility of 137Cs was higher than that of the rare-earth elements, but it was lower than that of 85Sr. When samples of fracture-coating material were separated into fractions with different specific gravity, there was a clear indication of radionuclide association with mineral groups.

  20. Distribution of radionuclides in mussels, winkles and prawns: Pt. 2

    Under laboratory conditions mussels, winkles and prawns have exhibited the ability to accumulate 237Np, 239Pu and 241Am from both sea water and food. In general, the soft tissues involved in feeding and digestion accumulated radionuclides most effectively. In digestive glands/hepatopancreas, the site of nuclide uptake was the digestive tubules. Other active tissues were the gill and heart of the Dublin Bay prawn and the pallial complex and operculum of the winkle. The prawn's gill was the only tissue to exhibit a clear preference between food and sea water labelling media -higher accumulation occurring via sea water. Heart tissue contained enhanced levels of 239Pu relative to 237Np and 241Am. The various glandular and secretory functions of the winkle's pallial complex may account for the comparable magnitude of nuclide activities in this tissue and the digestive gland. Other mucous secretions, on the external layers of the winkle's head and foot, have been observed to accumulate radionuclides but not as efficiently as the pallial complex. The most intense α-track distributions encountered were found in the winkle's operculum, these being attributable to its chitinous nature. Despite the relatively low activities present in Ravenglass mussels and winkles, their α-autoradiographs exhibited tissue activity trends in general accordance with those obtained experimentally. This finding provides some further support for the validity of laboratory-derived information and its extrapolation to environmental conditions. (Author)

  1. Calculations of neutron and proton induced reaction cross sections for actinides in the energy region from 10MeV to 1GeV

    Several nuclear model codes were applied to calculations of nuclear data in the energy region from 10MeV to 1GeV. At energies up to 100MeV the nuclear theory code GNASH was used for nuclear data calculation for neutrons incident for on 238U, 233-236U, 238-242Pu, 237Np, 232Th, 241-243Am and 242-247Cm. At energies from 100MeV to 1GeV the intranuclear cascade exciton model including the fission process was applied to calculations of protons and neutrons with 233U, 235U, 238U, 232Th, 232Pa, 237Np, 238Np, 239Pu, 241Am, 242Am and 242-248Cm. Determination of parameter systematics was a major effort in the present work that was aimed at improving the predictive capability of the models used. An emphasis was placed upon a simultaneous analysis of data for a variety of reaction channels for the nuclei considered, as well as of data that are available for nearby nuclei or for other incident particles. Comparisons with experimental data available on multiple reaction cross sections, isotope yields, fission cross sections, particle multiplicities, secondary particle spectra, and double differential cross sections indicate that the calculations reproduce the trends, and often the details, of the measurements data. (author) 82 refs

  2. Identifying Sources of Non-fallout Nuclear Contamination in Hudson River Sediments by Plutonium and Neptunium isotope ratios.

    Kenna, T. C.; Chillrud, S. N.

    2002-12-01

    In an effort to identify and characterize nuclear contaminants released from sources contained within the Hudson River drainage basin, Pu isotopes and 237Np have been measured in a series of sediment cores collected from various locations within the region. During the last several decades, the Hudson River has received input of radioactive contamination from several sources. The first and most significant, has been global fallout, which was a result of atmospheric testing of nuclear weapons primarily by governments of the United States and Former Soviet Union in the 1950s and 1960s. The second, is contamination resulting from reactor releases at the Indian Point Nuclear Power Plant (IPNPP) located on the Hudson River about 35 miles north of New York City. This facility began operation in 1962. A third source of radioactive contamination to the region is contamination resulting from activities at the Knolls Atomic Power Laboratory (KAPL) located on the Mohawk River, which began operation in 1946. Our research entails identifying different sources of nuclear contamination by measurement of plutonium and neptunium isotopic ratios by inductively coupled plasma mass spectrometry (ICP-MS). The isotopic composition of a nuclear contaminant is a sensitive indicator of its origin. By comparing the isotopic composition measured in fluvial sediments to mean values reported for global fallout (i.e. 240Pu/239Pu = 0.18 ñ 0.014, 237Np/239Pu = 0.48 ñ 0.07, and 241Pu/239Pu = .00194 ñ 00028) it is possible to identify contaminants as non-fallout in origin. To date, we have analyzed selected samples from 3 sediment cores collected from the following locations: 1) the Mohawk River downstream of KAPL, 2) the Hudson River above its confluence with the Mohawk River, and 3) the lower Hudson River at a location in close proximity to IPNPP. Isotopic analysis of sediments from the Mohawk River indicates contamination that is clearly non-fallout in origin (240Pu/239Pu ranges between 0

  3. Proliferation Resistant Fuel for Pebble Bed Modular Reactors

    Proliferation of nuclear weapons produced with power reactors plutonium has always been amajor problem of the nuclear energy industry. This includes the PebbleBed Modular Reactor(PBMR), which is a specific design of a GenIV High-Temperature Reactor (HTR), mainly due to its online refueling feature, which may be misused for the production of weapons gradeplutonium. A promising approach toward preventing the proliferation of power reactorplutoniumis to denaturate the plutonium by increasing the ratio of 238Pu to total Pu in the spentfuel(1). The 238Pu isotope is characterized by a high heat rate (approximately 567 W/kg) due to thealpha decay of the 238Pu with half-life of 87.74 yr, in addition to its high spontaneous fissionneutron emission, which is higher than that of 240Pu. Thus, the presence of 238Pu in Pu considerably complicates the design and construction of nuclear weapons based on Pu, owing tothese characteristics of 238Pu. Recent papers(2,3) show that a Pu mixture is proliferation resistant given that the weight ratio of 238Pu to Pu is larger than 6%. In this paper we have studied afeasible technique for ensuring that the 238Pu to Pu ratio, in the Pu produced in PBMR, is larger than 6% during the entire fuel cycle. Contamination of the spent fuel with 238Pu may be achieved by doping the nuclear fuel witheither 241Am or 237Np(4-13). The 238Pu isotope is obtained from both 241Am and 237Np by a neutron-capture reaction and the subsequent decay of the reaction products(13).The 237Np isotopeis by itself a potential weapons grade material. However, its large critical mass of 57±4 kg(14) andthe difficulty of extracting it from irradiated fuel elements make it impractical for weapons purposes. On the other hand, the critical mass of 241Am is smaller, i.e. 34 to 45 kg. However, withdecay heat production of 114W/kg, the critical mass becomes a heat source of 3.9 to 5.1 KW,which makes 241Am unsuitable for weapons applications(3). As a result, it is a non

  4. Determination of 236U and transuranium elements in depleted uranium ammunition by α-spectrometry and ICP-MS

    It is well known that ammunition containing depleted uranium (DU) was used by NATO during the Balkan conflict. To evaluate the origin of DU (the enrichment of natural uranium or the reprocessing of spent nuclear fuel) it is necessary to directly detect the presence of activation products (236U, 239Pu, 240Pu, 241Am, and 237Np) in the ammunition. In this work the analysis of actinides by α-spectrometry was compared with that by inductively coupled plasma mass spectrometry (ICP-MS) after selective separation of ultratraces of transuranium elements from the uranium matrix. 242Pu and 243Am were added to calculate the chemical yield. Plutonium was separated from uranium by extraction chromatography, using tri-n-octylamine (TNOA), with a decontamination factor higher than 106; after elution plutonium was determined by ICP-MS (239Pu and 240Pu) and α-spectrometry (239+240Pu) after electroplating. The concentration of Pu in two DU penetrator samples was 7 x 10-12 g g-1 and 2 x 10-11 g g-1. The 240Pu/239Pu isotope ratio in one penetrator sample (0.12±0.04) was significantly lower than the 240Pu/239Pu ratios found in two soil samples from Kosovo (0.35±0.10 and 0.27±0.07). 241Am was separated by extraction chromatography, using di(2-ethylhexyl)phosphoric acid (HDEHP), with a decontamination factor as high as 107. The concentration of 241Am in the penetrator samples was 2.7 x 10-14 g g-1 and -15 g g-1. In addition 237Np was detected at ultratrace levels. In general, ICP-MS and α-spectrometry results were in good agreement.The presence of anthropogenic radionuclides (236U, 239Pu,240Pu, 241Am, and 237Np) in the penetrators indicates that at least part of the uranium originated from the reprocessing of nuclear fuel. Because the concentrations of radionuclides are very low, their radiotoxicological effect is negligible. (orig.)

  5. Estimation of neptunium in uranium product

    Neptunium- 237 is a long-lived (t1/2 = 2.14 x 106 years) alpha emitter, present in low concentration levels in process solutions. The oxidation state and the nature of species present in process solutions depend upon the concentration of nitrous acid, nitric acid, other redox elements, and ageing. Owing to the complicated redox behavior, 237Np is usually distributed in all streams such as highly active raffinate, plutonium product, and uranium product etc. Estimation of neptunium in various PUREX process streams is necessary for understanding the path and routing of neptunium during reprocessing of fast reactor fuels. Various methods such as neutron activation analysis (NAA), spectrophotometry and electroanalytical techniques have - been reported for quantitative estimation of neptunium in the process streams. The sample of uranium product was received from Reprocessing Group, IGCAR to estimate the concentration of neptunium in uranium product. About 1 gram of the sample was dissolved in 8 M HNO3 and evaporated to dryness. Then, the product was redissolved in 1 M HNO3, and centrifuged. The supernatant solution (1 mL) was analyzed by gamma spectrometry, alpha spectrometry and neutron activation analysis. The gamma emission of 233Pa, which was in secular equilibrium with the parent 237Np was determined for estimating of neptunium by gamma spectrometry. Analysis of the sample by alpha spectrometry showed the presence of plutonium and americium in the uranium product. These alpha emitters hindered the analysis of neptunium. Hence, Np(IV) was separated from other actinides by extraction into 1 mL of 0.5 M TTA/xylene and stripped with 10 M HNO3. An aliquot (0.1 mL) of organic phase was dried in an aluminum foil and irradiated at KAMINI reactor at the neutron flux of 1010 neutrons/cm/sec for 4 hours. The activation product, 238Np, formed was monitored by the characteristic gamma emissions at 984 keY and 1028 keY. The stripped solution was subjected to alpha spectrometry

  6. Effect of reducing groundwater on the retardation of redox-sensitive radionuclides

    Rose TP

    2008-12-01

    Full Text Available Abstract Laboratory batch sorption experiments were used to investigate variations in the retardation behavior of redox-sensitive radionuclides. Water-rock compositions were designed to simulate subsurface conditions at the Nevada Test Site (NTS, where a suite of radionuclides were deposited as a result of underground nuclear testing. Experimental redox conditions were controlled by varying the oxygen content inside an enclosed glove box and by adding reductants into the testing solutions. Under atmospheric (oxidizing conditions, radionuclide distribution coefficients varied with the mineralogic composition of the sorbent and the water chemistry. Under reducing conditions, distribution coefficients showed marked increases for 99Tc (from 1.22 at oxidizing to 378 mL/g at mildly reducing conditions and 237Np (an increase from 4.6 to 930 mL/g in devitrified tuff, but much smaller variations in alluvium, carbonate rock, and zeolitic tuff. This effect was particularly important for 99Tc, which tends to be mobile under oxidizing conditions. A review of the literature suggests that iodine sorption should decrease under reducing conditions when I- is the predominant species; this was not consistently observed in batch tests. Overall, sorption of U to alluvium, devitrified tuff, and zeolitic tuff under atmospheric conditions was less than in the glove-box tests. However, the mildly reducing conditions achieved here were not likely to result in substantial U(VI reduction to U(IV. Sorption of Pu was not affected by the decreasing Eh conditions achieved in this study, as the predominant sorbed Pu species in all conditions was expected to be the low-solubility and strongly sorbing Pu(OH4. Depending on the aquifer lithology, the occurrence of reducing conditions along a groundwater flowpath could potentially contribute to the retardation of redox-sensitive radionuclides 99Tc and 237Np, which are commonly identified as long-term dose contributors in the risk

  7. Target accuracy of MA nuclear data and progress in validation by post irradiation experiments with the fast reactor 'JOYO'

    This paper presents a discussion about the target accuracy of MA nuclear data for fast reactor cycle system development, as well as the validation work on those nuclear data by PIE analyses. The PIE analyses are in progress on fuels and MA samples (237Np, 241Am, 243Am, 244Cm) irradiated at the experimental fast reactor 'JOYO'. The analysis result on the first examined MA sample suggested the necessity of re-evaluation of the isomeric ratio for 241Am capture reaction both in ENDF/B-VI and in JENDL-3.3. The above result contributes to the uncertainty-reduction both of burnup reactivity loss and of gamma energy release from fuel assemblies. (author)

  8. Speciation of long-lived radionuclides in the environment

    Xiaolin Hou

    2008-11-15

    This project started in November 2005 and ended in November 2008, the work and research approaches are summarized in this report. This project studied the speciation of radionuclides in environment. A number of speciation analytical methods are developed for determination of species of 129I, 99Tc, isotopes of Pu, and 237Np in seawater, fresh water, soil, sediment, vegetations, and concrete. The developed methods are used for the investigation of the chemical speciation of these radionuclides as well as their environmental behaviours, especially in Danish environment. In addition the speciation of Pu isotopes in waste samples from the decommissioning of Danish nuclear facilities is also investigated. The report summarizes these works completed in this project. Through this research project, a number of research papers have been published in the scientific journals, the research results has also been presented in the Nordic and international conference/meeting and communicated to international colleagues. Some publications are also enclosed to this report. (au)

  9. Fission rates of nuclides in the neutron field formed on the surface of a massive lead target irradiated with protons at energy of 5.0 Gev

    Fission rates of the nuclides Th-232, U-235 (with and without cadmium filter), U-234, U-236, Np-237 and U-238 were measured on the side surface of a massive lead target irradiated at the Syncrophasotron accelerator (JINR, Dubna) with protons at kinetic energy of 5.0 GeV. These measurements were carried out using the technique of solid-state nuclear track detectors. Comparison of the measured values was made with the results obtained by means of Monte-Carlo simulation using the computer code DCM/CEM. - Highlights: • The spallation neutron source based on a massive lead target irradiated with 5.0 GeV protons. • The technique of solid-state nuclear track detectors was used for the measurements. • Fission rates of 232Th, 237Np and uranium isotopes were measured for a spallation neutron source. • The measured values are compared with the results of Monte-Carlo simulation

  10. Improved radioanalytical method for the simultaneous determination of Th, U, Np, Pu and Am(Cm) on a single TRU column by alpha spectrometry and ICP-MS

    Macsik, Z.; Groska, J.; Vajda, N. [RadAnal Ltd., Budapest (Hungary); Vogt, S.; Kim, C.S.; Maddison, A.; Donohye, D. [IAEA, Seibersdorf (Austria). Environmental Sample Lab.; Kis-Benedek, G. [IAEA, Seibersdorf (Austria). Terrestrial Environment Lab.

    2013-05-01

    A radioanalytical method based on the use of a single TRU extraction chromatographic column and selective, on-column oxidation state adjustment of actinides was developed for the determination of Th, U, Np, Pu and Am(Cm) in environmental samples (such as sediment and swipe samples). The procedure of Vajda et al. was further investigated and optimized focusing on the separation of Th, U and Np. The improved method combines two measurement techniques - alpha spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) - which allows to obtain more reliable information on a wider range of isotopes: {sup 228}Th, {sup 230}Th, {sup 232}Th, {sup 234}U, {sup 235}U, ({sup 236}U), {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 241}Am, {sup 242}Cm, {sup 243}Cm and {sup 244}Cm from one sample. (orig.)

  11. Evaluation of chelation concentration and cation separation of actinides at ultra-trace levels in urine matrix

    The feasibility of measuring picogram levels of actinides in a urine matrix using ion chromatography coupled on-line to an inductively coupled plasma quadrupole mass spectrometer (IC-Q-ICPMS) was investigated. A chelation column for separation of matrix ions and preconcentration of the actinides was combined with a cation-exchange column for separation of the actinides. Sample preparation required simple addition of ammonium acetate to adjust the pH of the urine matrix. Spike solutions containing 232Th, 237Np, 238U, 239Pu, and 241Am were added to undiluted urine, diluted urine (1 : 9) and water. This approach enhanced the signal sensitivities of all the tested actinides over two orders of magnitude in the water matrix, while certain elements (especially Am) can still be effectively concentrated in undiluted urine. (author)

  12. Reactor physics experiments related to transmutation in the KUCA

    Shiroya, Seiji [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1997-11-01

    At the Kyoto University Critical Assembly (KUCA), {sup 237}Np/{sup 235}U fission rate ratios are being measured using the back-to-back type double fission chamber to examine the nuclear data and the computational method for the transmutation of minor actinides (MA) in light water reactors (LWRs). The neutron spectra of cores are systematically being varied by changing the moderator-to-fuel volume ratio (V{sub m}/V{sub f}). The measured data are being compared with the calculated results by SRAC with three different nuclear data files. It has been indicated that the calculated results with JENDL-3.2 agreed better with the measured ones than those with JENDL-3.1 and ENDF/B-VI, although the calculated results underestimated the measured ones by around 10%. (author)

  13. Preparation of isotopes and sources of actinide elements

    As the C.E.A. possesses no isotopic separation facility, the productions of isotopes of actinide elements are performed: a) by neutron irradiation and chemical treatment of special targets, b) by milking decay products from stocks of aged actinide elements, c) by chemical treatment of alpha active wastes. These productions concern the following isotopes: 233U, 238Pu, 242Pu, 243Cm, 242Cm, 244Cm (a); 228Th, 229Th, 234U, 237U, 239Np, 240Pu, 241Am, 248Cm (b); 237Np, 241Am (c). These isotopes are produced to satisfy French and international needs and are sent to users in various forms: solutions, metals, oxides, fluorides, or in different sources forms. The preparation of the sources represents an important field of activities divided into two parts: 1/Industrial sources: production of large series of different sources, 2/ Scientific sources: production of sources suitable for a specific scientific problem. A large overview of these activities is given

  14. Photon and proton induced fission on heavy nuclei at intermediate energies

    Andrade-II, E.; Karapetyan, G.S.; Deppman, A.; Guimaraes, V. [Universidade de Sao Paulo (USP), Sao Paulo, SP (Brazil). Instituto de Fisica; Balabekyan, A.R. [Yerevan State University, Alex Manoogian 1, Yerevan (Armenia); Demekhina, N.A. [Yerevan Physics Institute, Alikhanyan Brothers 2, Yerevan (Armenia); Joint Institute for Nuclear Research (JINR), Flerov Laboratory of Nuclear Reactions (LNR), Moscow (Russian Federation)

    2014-07-01

    We present an analysis of fission induced by intermediate energy protons or photons on actinides. The 660 MeV proton induced reactions are on {sup 241}Am, {sup 238}U, and {sup 237}Np targets and the Bremsstrahlung-photons with end-point energies at 50 MeV and 3500 MeV are on {sup 232}Th and {sup 238}U targets. The study was performed by means of the Monte Carlo simulation code CRISP. A multimodal fission extension was added to the code within an approach which accounts for the contribution of symmetric and asymmetric fission. This procedure allowed the investigation of fission cross sections, fissility, number of evaporated nucleons and fission-fragment charge distributions. The comparison with experimental data show a good agreement between calculations and experiments. (author)

  15. Photon and proton induced fission on heavy nuclei at intermediate energies

    Andrade-II E.

    2014-04-01

    Full Text Available We present an analysis of fission induced by intermediate energy protons or photons on actinides. The 660 MeV proton induced reactions are on 241Am, 238U, and 237Np targets and the Bremmstrahlung-photons with end-point energies at 50 MeV and 3500 MeV are on 232Th and 238U targets. The study was performed by means of the Monte Carlo simulation code CRISP. A multimodal fission extension was added to the code within an approach which accounts for the contribution of symmetric and asymmetric fission. This procedure allowed the investigation of fission cross sections, fissility, number of evaporated nucleons and fission-fragment charge distributions. The comparison with experimental data show a good agreement between calculations and experiments.

  16. Intermediate mass fragment production in the proton-induced reactions of heavy targets

    Deppman, A; Guimaraes, V; Karapetyan, G S; Balabekyan, A R; Demekhina, N A; Adam, J

    2013-01-01

    The production of intermediate-mass fragments (IMFs) formed in the proton-induced reaction with $^{238}$U and $^{237}$Np at 660 MeV was measured in the LNP Phasotron and in U-400M Cyclotron, Joint Institute for Nuclear Research (JINR), Dubna, Russia. We have applied the induced-activation method in off-line analysis. A total of 115 isotopes of all elements in the range $7 \\leq A \\leq 69$ were unambiguously identified with high precision. There is a consideration that the formed nuclides could be produced in a very asymmetric binary decay of heavy nuclei originating from the spallation of heavy targets. Mass-yield distributions were derived from the data, and were compared with the the simulation code CRISP for multi modal fission.

  17. Method of measurement of cross sections of heavy nuclei fission induced by intermediate energy protons

    The purpose of this work is experimental studies of the energy dependence of the fission cross sections of heavy nuclei, natPb, 209Bi, 232Th, 233U, 235U, 238U, 237Np and 239Pu, 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 natPb, 209Bi, 235U and 238U are reported. (author)

  18. The Separation and Measurement of Np( Ⅴ, Ⅵ) and Its Daughter 237Pa

    2001-01-01

    By means of liquid scintillation measurement of the Np sample solution which reaches 237Np-233Pa equilibrium with its daughter Pa, the detection efficiency of low energy y spectrum at 28.54, 86.59 and 311.98 keV are obtained. Using these data, the separation of different valence of Np from its daughter Pa is studied. The initial valence of Np is Np(Ⅴ) in solution. In order to compare the separation results, H2O2 is used to prepare Np( Ⅴ). Np(Ⅵ) is prepared by K2Cr2O7 and concentrated HNO3, respectively. TTA and/or TOPO and the back-extraction of them are used to separate Np and Pa. It indicates that initial Np( Ⅴ ) can be separated by TTA or TOPO, and the Np

  19. Kinetic study of the reduction of Np(VI) with humic acid

    237Np is one of the most long-life and toxic nuclides in the high level nuclear waste, and will become the primary hazard in the final nuclear waste disposal after a long time of storage, so it is important to study the chemical behaviour of Np. Humic acid is a kind of organic compound reduction in nature. The study of the kinetics of the reduction of Np(VI) with humic acid will afford a basis for further study of the chemical behaviour of Np in the environmental water. Considering the rapid exchange of Np valences during this experiment, extractants TTA and TOPO are used to analyze simultaneously three Np valences (VI), (V) and (IV)

  20. Absolute reaction rate measurement with D-D neutron source in polyethylene spherical shell

    The absolute reaction rate distribution measurements in a polyethylene spherical shell with 38.6 cm outside diameter and 10 cm thickness were performed with D-D neutron source. By combining fission method and activation method, rich-uranium fission chamber, depleted-uranium fission chamber, 237Np fission chamber and 115In activation foils were placed at several positions on the equatorial line of the inner face of the shell, and the absolute reaction rates were obtained. The uncertainty of fission rates is 2.5%-4.3%, while the uncertainty of activation rates is about 6.3%. The reaction rates were calculated by MCNP and ENDF/B-VII. 0. The calculated results are lower than the measured results and 238U is typical. (authors)

  1. MANTRA: An Integral Reactor Physics Experiment to Infer Actinide Capture Cross-sections from Thorium to Californium with Accelerator Mass Spectrometry

    G. Youinou; C. McGrath; G. Imel; M. Paul; R. Pardo; F. Kondev; M. Salvatores; G. Palmiotti

    2011-08-01

    The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectrometry technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm and 248Cm.

  2. Progress report on safety research of high-level waste management for the period April 1987 to March 1988

    Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1987 are reviewed in the three sections of the report. The topics are as follows: 1) On performance and durability of waste forms and engineered barrier materials, accelerated alpha radiation stability of glass form and Synroc has been investigated and stress corrosion cracking of canister materials was examined under simulated conditions. 2) Sorption of 237Np on granite samples and behavior of iron during weathering of granites were studied with respect to safety evaluation for geological disposal. 3) Actual waste was transported from the Tokai Reprocessing Plant and hot operation using the actual waste was initiated at WASTEF. (author)

  3. Neutron-induced cross section of actinides via the surrogate-reaction method

    The surrogate-reaction method is an indirect way of determining cross sections for reactions that proceed through a compound nucleus. It consists of using a transfer reaction to produce the same decaying nucleus as the one formed in the desired neutron-induced reaction. This technique may enable neutron-induced cross sections to be extracted for short-lived nuclei that otherwise cannot be measured. However, the validity of the surrogate method for extracting capture cross sections has to be investigated. In this work we study the reactions 238U(d,p)239U, 238U(3He,t)238Np, 238U(3He,4He)237U as surrogates for neutron-induced reactions on 238U, 237Np and 236U, respectively, for which good quality data exist. The experimental set-up enabled the measurement of fission and gamma-decay probabilities. First results are presented and discussed. (authors)

  4. Nucleon-induced fission cross-sections at transitive energy region 20-200 MeV

    The new approach to the calculation of nucleon induced fission cross sections at energies 20-200 MeV is presented. The cross sections of multiconfiguration fission is calculated as a sum of fission cross-sections for nuclei formed in process of fast (direct) and precompound stage of fission reaction. The intranuclear cascade model is used for description of direct stage and precompound-statistical model for calculation of fission and de-excitation cross sections. Calculated with new optical model parameters sets fission cross sections are compared with experimental data for neutron-induced fission of 237Np, 239Pu, 235,238U and proton-induced fission of 235,238U. Brief information about new code system is also presented. (author)

  5. Measurement of the 240,242Pu Neutron-induced Fission Cross Sections

    Salvador-Castiñeira, P.; Bevilacqua, R.; Bryś, T.; Hambsch, F.-J.; Oberstedt, S.; Pretel, C.; Vidali, M.

    The neutron-induced fission cross section of 240,242Pu has been measured at the Van de Graaff facility of the Institute for Reference Materials and Measurements (JRC-IRMM). A Twin-Frisch Grid Ionization Chamber (TFGIC) has been used in a back-to-back geometry with the secondary standards 237Np and 238U to normalize the cross section. The energy range measured is from 0.2 keV up to 3 MeV. Preliminary results show some discrepancies around 1 MeV for the 242Pu with the ENDF/B-VII.1 evaluation. The spontaneous fission half-life has been measured for both isotopes, too. Preliminary results show reasonable agreement with the recommended values.

  6. Neutron-induced Fission Cross Section of 240242Pu up to En = 3 MeV

    Salvador-Castiñeira, P.; Bryś, T.; Hambsch, F.-J.; Oberstedt, S.; Pretel, C.; Vidali, M.

    2014-05-01

    The neutron-induced fission cross sections of 240,242Pu have been measured at JRC-IRMM with incident neutron energy from 0.2 MeV up to 3 MeV. A Twin-Frisch Grid Ionization Chamber (TFGIC) has been used in a back-to-back geometry. The measurements have been performed using the secondary standards 237Np and 238U as a reference. The purity of the plutonium samples was 99.89% for 240Pu and 99.97% for 242Pu. The results obtained follow the ENDF/B-VII.1 evaluation for 240Pu, but some discrepancies are visible around E/n = 1 MeV for 242Pu. In addition, the spontaneous fission half-life has been measured for both isotopes.

  7. Benchmark experiment on the model of fusion reactor blanket with uranium neutron multiplier

    Benchmark experiment on the model of thermonuclear reactor blanket with 14 MeV neutron source is described. The model design corresponds to the known concept of the fast hybrid blanket with 238U neutron multiplier and main tritium production on 6Li. Detailed measurements of the following process velocities were carried out: tritium production on lithium isotopes; reactions modelling tritium production; (n, γ) and (n, 2n) processes for 238U; fission reactions for 235,238U, 239Pu, 237Np. Neutron flux integral measurements were performed by a set of threshold detectors on the basic of the 115In(n, n'), 204Pb(n, n'), 64Zn(n, p), 27Al(n, p), 56Fe(n, p), 107Ag(n, 2n), 63Cu(n, 2n) and 64(n, 2n) reactions

  8. National Low-Level Waste Management Program Radionuclide Report Series

    Adams, James Paul; Carboneau, Michael Leonard; Allred, William Edgar

    1999-03-01

    The National Low Level Waste Management Program at the Idaho National Engineering and Environmental Laboratory has published a report containing key information about selected radionuclides that are most likely to contribute significantly to the radiation exposures estimated from a performance assessment of a low-level radioactive waste (LLW) disposal facility. The information includes physical and chemical characteristics, production means, waste forms, behavior of the radionuclide in soils, plants, groundwater, and air, and biological effects in animals and humans. The radionuclides included in this study comprise all of the nuclides specifically listed in 10CFR61.55, Tables 1 and 2, 3 H, 14 C, 59 Ni, 60 Co, 63 Ni, 90 Sr, 94 Nb, 99 Tc, 129 I, 137 Cs, 241 Pu, and 242 Cm. Other key radionuclides addressed in the report include 237 Np, 238 U, 239 Pu, and 241 Am. This paper summarizes key information contained within this report.

  9. National Low-Level Waste Management Program Radionuclide Report Series

    J.P. Adams; M.L. Carboneau; W.E. Allred

    1999-02-01

    The National Low Level Waste Management Program at the Idaho National Engineering and Environmental Laboratory has published a report containing key information about selected radionuclides that are most likely to contribute significantly to the radiation exposures estimated from a performance assessment of a low-level radioactive waste (LLW) disposal facility. The information includes physical and chemical characteristics, production means, waste forms, behavior of the radionuclide in soils, plants, groundwater, and air, and biological effects in animals and humans. The radionuclides included in this study comprise all of the nuclides specifically listed in 10CFR61.55, Tables 1 and 2, 3 H, 14 C, 59 Ni, 60 Co, 63 Ni, 90 Sr, 94 Nb, 99 Tc, 129 I, 137 Cs, 241 Pu, and 242 Cm. Other key radionuclides addressed in the report include 237 Np, 238 U, 239 Pu, and 241 Am. This paper summarizes key information contained within this report.

  10. Monte Carlo calculations on transmutation of transuranic nuclear waste isotopes using spallation neutrons. Difference of lead and graphite moderators

    Transmutation rates of 239Pu and some minor actinides (237Np, 241Am, 245Cm and 246Cm), in two accelerator driven systems (ADS) with lead or graphite moderating environments, were calculated using the LAHET code system. The ADS that were used had a large volume (∼ 30 m3) and contained no fissile material, except for a small amount of fissionable waste nuclei that existed in some cases. Calculations were performed at incident proton energy of 1.5 GeV and spallation target was lead. Also breeding rates of 239Pu and 233U as well as the transmutation rates of two long-lived fission products 99Tc and 129I were calculated at different locations in the moderator. It is shown that an ADS with graphite moderator is a much more effective transmuter than that with lead moderator

  11. A detailed study of the steam entry effect of a GCFR

    The results of detailed steam entry calculations for a realistic reactor model are reported. The calculations predict that the effect of steam entry into a 300 MWe GCFR fueled with normal Pu mixed oxide fuel is negative in the whole range of interest. The effect is very sensitive to the method used for compensating the excess reactivity to obtain criticality before steam ingress occurs. Steam entry reactivity is sensitive to fuel enrichment and boron control. Changes in the isotopic compositions of Pu can give raise to important changes of the steam entry effect. A small amount of 237Np added to the fuel can considerably improve steam entry reactivity without noticeably deteriorating breeding ratio or affecting reactor operation

  12. Radiochemical studies in the development of deep geological repository in the Czech Republic

    In this contribution the main achievements of radiochemical studies performed within the framework of the Czech DGR development programme are summarized and further plans outlined. The results of selection of the most dangerous radionuclides in spent fuel assemblies from VVER 440 reactors, based on spent fuel inventory calculations and analyses of migration rate of radionuclides to the environment, are presented in the first part of the contribution. It is shown that 14C, 129I, 126Sn, 135Cs, 36Cl, 79Se, 226Ra, 237Np, 229Th, and 242Pu belong among the most dangerous radionuclides in the Czech disposal concept. Problems with the determination of migration parameters of radionuclides are described in the second part of this contribution. (author)

  13. Sequential Injection Method for Rapid and Simultaneous Determination of 236U, 237Np, and Pu Isotopes in Seawater

    Qiao, Jixin; Hou, Xiaolin; Steier, Peter;

    2013-01-01

    target analytes, whereupon plutonium and neptunium were simultaneously isolated and purified on TEVA, while uranium was collected on UTEVA. The separation behavior of U, Np, and Pu on TEVA–UTEVA columns was investigated in detail in order to achieve high chemical yields and complete purification...... for the radionuclides of interest. 242Pu was used as a chemical yield tracer for both plutonium and neptunium. 238U was quantified in the sample before the separation for deducing the 236U concentration from the measured 236U/238U atomic ratio in the separated uranium target using accelerator mass spectrometry....... Plutonium isotopes and 237Np were measured using inductively coupled plasma mass spectrometry after separation. The analytical results indicate that the developed method is robust and efficient, providing satisfactory chemical yields (70–100%) of target analytes and relatively short analytical time (8 h/sample)....

  14. Rapid and simultaneous determination of neptunium and plutonium isotopes in environmental samples by extraction chromatography using sequential injection analysis and ICP-MS

    Qiao, Jixin; Hou, Xiaolin; Roos, Per;

    2010-01-01

    This paper reports an automated analytical method for rapid and simultaneous determination of plutonium isotopes (239Pu and 240Pu) and neptunium (237Np) in environmental samples. An extraction chromatographic column packed with TrisKem TEVA® resin was incorporated in a sequential injection (SI......) system for the isolation of plutonium and neptunium from matrix elements and interfering nuclides. Valence adjustment is a crucial step to ensure the same chemical behavior of plutonium and neptunium onto the TEVA column and consequently to accomplish their simultaneous separation and detection. Distinct...... procedures were investigated and compared for the adjustment of oxidation states of plutonium and neptunium to Pu(IV) and Np(IV), respectively. A two-step protocol using sulfite and concentrated nitric acid as redox reagents was proven to be the most effective method. The analytical results for both...

  15. Application of Moessbauer spectroscopy to the study of neptunium adsorbed on deep-sea sediments

    A Neptunium Moessbauer spectrometer (the first in Great Britain) was constructed and the Moessbauer spectra of NpAl Laves phase alloy obtained. Neptunium was sorbed onto a calcareous deep-sea sediment from sea water, using a successive-loading technique. Sorption appeared to be by an equilibrium reaction, and because of the low solubility of neptunium in seawater, this meant that the maximum loading that could be achieved was 8mg237Np/g sediment. This proved to be an adequate concentration for Moessbauer measurements and a Moessbauer spectrum was obtained. This showed that most of the neptunium was in exchange sites and not present as precipitates of neptunium compounds. It was probably in the 4+ state indicating that reduction had occurred during sorption. This work has demonstrated that Moessbauer Spectroscopy has great potential as an aid to understanding the mechanism of actinide sorption in natural systems. (author)

  16. Effect of the impurity on the results of the fission rate measurements and elimination method

    The nuclear fission method is an important method for neutron measurement. The 235U and 239Pu fission chamber can be used to measure the thermal neutron, and 238U, 232Th and 237Np fission chamber are often used to measure fast neutron. Because the materials for the fission chambers are got by the isotope separation or artificial method, they can contain a few other impurity atoms. The experimental measurements show that the measurement results for the fast neutron can be obviously affected by the few impurities which can be fissioned by the thermal neutron. The thermal fission can be corrected by the method of the thermal fission correction. It also can be eliminated by using the fast fission chamber with Cd

  17. Characterization of selected waste tanks from the active LLLW system

    From September 1989 through January of 1990, there was a major effort to sample and analyze the Active Liquid-Low Level Waste (LLLW) tanks at ORNL which include the Melton Valley Storage Tanks (MVST) and the Bethel Valley Evaporator Service Tanks (BVEST). The purpose of this report is to summarize additional analytical data collected from some of the active waste tanks from November 1993 through February 1996. The analytical data for this report was collected for several unrelated projects which had different data requirements. The overall analyte list was similar for these projects and the level of quality assurance was the same for all work reported. the new data includes isotopic ratios for uranium and plutonium and an evaluation of the denature ratios to address criticality concerns. Also, radionuclides not previously measured in these waste tanks, including 99Tc and 237Np, are provided in this report

  18. Response of actinides to flux changes in high-flux systems

    When discussing the transmutation of actinides in accelerator-based transmutation of waste (ATW) systems, there has been some concern about the dynamics of the actinides under high transient fluxes. For a pure neptunium feed, it has been estimated that the 238Np/237Np ratio increase due to an increasing flux may lead to an unstable, positive reactivity growth. In this analysis, a perturbation method is used to calculate the response of the entire set of actinides in a general way that allows for more species than just neptunium. The time response of the system can be calculated; i.e., a plot of fuel composition and reactivity versus time after a change in flux can be made. The effects of fission products can also be included. The procedure is extremely accurate on short time scales (∼ 1000 s) for the flux levels we contemplate. Calculational results indicate that the reactivity insertions are always smaller than previously estimated

  19. Energetic identification of ions of very low fluence; Identificacion energetica de iones de muy baja fluencia

    Mut C, D.A.; Balcazar, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2005-07-01

    The calibration of 2 types of plastics is presented (cellulose nitrate and polycarbonate) that detect and identify the energy of helium slight ions (1.5 to 10 MeV) and protons (0.3 to 6 MeV) in flows so low of a single particle /cm{sup 2}. This methodology is of importance in radiological protection to identify the actinides {sup 232} Th, {sup 241} Am, {sup 238} U, {sup 235} U, {sup 237} Np by means of its alpha emissions, or to carry out neutron spectroscopy in an ignored energy field by protons retrocession. The ion identification is adjusted for all the cases to a simple expression of the exponential type. The comparison is achieved among detection efficiencies for the detectors. (Author)

  20. Strontium and Actinides Removal from Savannah River Site Actual Waste Samples by Freshly Precipitated Manganese Oxide

    The authors investigated the performance of freshly precipitated manganese oxide and monosodium titanate (MST) for the removal of strontium (Sr) and actinides from actual high-level waste. Manganese oxide precipitation occurs upon addition of a reductant such as formate (HCO2-) or peroxide (H2O2) to a waste solution containing permanganate (MnO4-). An addition of non-radioactive strontium typically precedes the MnO4- and reductant addition, which serves primarily to isotopically dilute the strontium-90 (90Sr) present in the waste. Tests utilized a Tank 37H/44F composite waste solution. Personnel significantly increased the concentration of actinides in the waste by the addition of acidic americium/curium solution (F-Canyon Tank 17.1 solution), which contained a significant quantity of plutonium (Pu), and neptunium-237 (237Np) stock solution. Initial tests examined three manganese oxide treatment options

  1. Collective band properties in actinide nuclei well deformed

    In actinides, proton i13/2 orbitals and neutron j15/2 orbitals are both near Fermi surface. At a great rotation speed, driving and Coriolis forces change the surface forces, in particular, they lower pairing forces inside the nucleus. The use of Coulomb excitation with the help of heavy and very heavy projectiles such as 32S, 84Kr, 142Nd, 208Pb at 232Th together with the most recent techniques of spectroscopy allowed to populate yrast bands of 230Th, 232Th, 235U and 237Np nuclei up to high spin states and together the even-even nuclei states in different collective bands. Experimental results have been analyzed in the frame of the different current models. The low spin states of rotational bands have been reproduced in a previous calculation using the nucleon-nucleon effective interaction of Skyrme III

  2. Development of fission micro-chambers for nuclear waste incineration studies

    The incineration of transuranic elements by neutron induced fission is a very promising way to reduce long-term radiotoxicity of nuclear waste. The Mini-Inca aims to outline the ideal physical conditions to transmute minor actinides, mainly 241-243Am, 237Np and 244-245Cm. For some actinides there are large discrepancies of neutron cross sections taken from different evaluated nuclear data libraries. These cross sections play a dominant role in transmutation systems. For instance, a factor 20 was pointed out for the 242gsAm thermal neutron capture cross section from JEF-2.2 (5500 barns) and ENDF-B/VI (250 barns) libraries. Computer simulations can lead to controversial results depending on the nuclear data library that was used. To measure the incineration rate of minor actinides, and to provide an unambiguous experimental reference, fission micro-chambers are of great interest. (author)

  3. Analysis of americium-beryllium neutron source composition using the FRAM code

    Hypes, P. A. (Philip A.); Bracken, D. S. (David S.); Sampson, Thomas E.; Taylor, W. A. (Wayne A.)

    2002-01-01

    The FRAM code was originally developed to analyze high-resolution gamma spectra from plutonium items. Its capabilities have since been expanded to include analysis of uranium spectra. The flexibility of the software also enables a capable spectroscopist to use FRAM to analyze spectra in which neither plutonium nor uranium is present in significant amounts. This paper documents the use of FRAM to determine the {sup 239}Pu/{sup 241}Am, {sup 243}Am/{sup 241}Am, {sup 237}Np/{sup 241}Am, and {sup 239}Np/{sup 241}Am ratios in americium-beryllium neutron sources. The effective specific power of each neutron source was calculated from the ratios determined by FRAM in order to determine the americium mass of each of these neutron sources using calorimetric assay. We will also discuss the use of FRAM for the general case of isotopic analysis of nonplutonium, nonuranium items.

  4. Application of Novel Nanoporous Sorbents for the Removal of Heavy Metals, Metalloids, and Radionuclides

    Mattigod, Shas V.; Fryxell, Glen E.; Parker, Kent E.; Lin, Yuehe

    2005-10-01

    A new class of hybrid nanoporous materials for removing toxic heavy metals, oxyanions, and radionuclides from aqueous waste streams has been developed at the Pacific Northwest National Laboratory. These novel materials consist of functional molecules such as thiols, ethylenediamine complexed copper, and carbamoylphosphonates that are self-assembled as monolayers within the nanopores of a synthetic silica-based material. Tests indicated that these sorbents (self-assembled monolayers on mesoporous silica ? SAMMS) can achieve very high sorbate loadings ({approx}6 meq/g) very rapidly with relatively high specificity (Kd: 1?108 ml/g). Because of the specifically tunable nature of the functionalities, these nanoporous sorbents can be targeted to remove a selected category of contaminants such as heavy metals (Ag, Cd, Cu, Hg, and Pb), oxyanions (As and Cr), and radionuclides (137Cs, 129I, 237Np, and isotopes of Pu, Th, and U) from waste streams.

  5. Sorption experiments at oxic and anoxic conditions

    Sorption of U, Pu, Np, Th and Tc at anoxic conditions were studied using radionuclides 233U, 236Pu, 237Np, 234Th and 99Tc. Samples were selected to represent common rocks and minerals in Finnish bedrock. The determinations of K-values and sorption percents in quartz plagioclase, potassium feldspar, honrnblende, biotite, mica gneiss, tonalite, granodiorite, porphyritic granite, granitic gneiss and rapakivi granite were performed by the autoradiographic method. The sorpion of these elements was studied under anoxic conditions using synthetic granitic ground water and synthetic bentonite water containing ferrous iron. For redox-sensitive U, Np and Pu experiments were made by spiking samples with water containing the radioactive isotope at one specified oxidation state

  6. Paleocorrosion studies in deep sea sediments and the geological disposal of nuclear wastes

    Fehrenbach, L.; Maurette, M. (Centre National de la Recherche Scientifique, 91 - Orsay (France). Lab. Rene Bernas); Guichard, F. (Centre National de la Recherche Scientifique, 91 - Gif-sur-Yvette (France). Centre des Faibles Radioactivites); Havette, A. (Paris-11 Univ., 91 - Orsay (France). Dept. des Sciences de la Terre); Monaco, A. (College Scientifique Universitaire, 66 - Perpignan (France). Lab. de Recherches de Sedimentologie Marine)

    1984-09-01

    Uncertainties still surround assessment of the safety of disposal of nuclear wastes incorporated into 'radwaste' matrices. This is mostly due to the long time required for radioactive decay of /sup 237/Np. The present work explores the usefulness of an experimental approach in 'paleocorrosion', which should help in minimizing such uncertainties. In this approach, polished sections of sediments containing high concentrations of natural analogues of radwaste matrices are subjected to element micromapping. Thus it is possible to characterize the long-term interactions of such analogues in their geological repositories, and to identify which generate reaction aureoles and protective and/or unprotective coatings. These analogues include grains incorporated in deep sea sediments (uraninite and quartz from the Oklo uranium ore deposit; volcanic ash particles; magnetic cosmic spherules). The present results indicate that uraninite should be a much more durable radwaste matrix than any type of glass in deep sea sediments.

  7. Neutron-induced cross sections of actinides via the surrogate-reaction method

    Ducasse Q.

    2013-12-01

    Full Text Available The surrogate-reaction method is an indirect way of determining cross sections for reactions that proceed through a compound nucleus. This technique may enable neutron-induced cross sections to be extracted for short-lived nuclei that otherwise cannot be measured. However, the validity of the surrogate method has to be investigated. In particular, the absence of a compound nucleus formation and the Jπ dependence of the decay probabilities may question the method. In this work we study the reactions 238U(d,p239U, 238U(3He,t238Np, 238U(3He,4He237U as surrogates for neutron-induced reactions on 238U, 237Np and 236U, respectively, for which good quality data exist. The experimental set-up enabled the measurement of fission and gamma-decay probabilities. The first results are hereby presented.

  8. Radioactivity studies. Progress report, April 30, 1984-June 1, 1985

    This report includes information pertaining to metabolic studies of neptunium and protactinium in the adult baboon. Recent investigations have provided additional data on the uptake, distribution, retention and excretion of Np-237, Np-239 and Pa-233 in baboons following single intravenous and gavage administrations. Data is also presented on the gastrointestinal absorption of isotopes of uranium, neptunium and plutonium in individual baboons after receiving multiple gavage administrations at selected time intervals and nutritional states. The gastrointestinal (GI) absorption (f1 values) and retention factors have been calculated for each of these nuclides. We have begun metabolic studies on the adult tamarin (Saquinis labiatus). Data are presented in this report on the preliminary results of the metabolism of Np-239 bicarbonate intravenously injected into three females and one male tamarin. These data are discussed in comparison with similar results obtained with our baboons and with other species. 28 refs., 20 figs., 14 tabs

  9. Data basis for a site specific radioactive element migration analysis of a repository

    Migration analysis is of considerable importance in long-term safety analysis of radioactive waste repositories. As a first step the author calculates the transport of radionuclides using data, as far as possible, for an undisturbed hydrogeology. Thereby a reference case is defined. In a later step, possible events and processes can be considered leading to a deviation from the reference case. The present work gives the data base for a selected part of a comprehensive geosphere transport calculation. The report is restricted to a critical evaluation of parameters pertinent to the migration analysis of the 245Cm chain. This includes the important nuclide 237Np. For the first time it is possible to perform a site specific calculation for repositories planned in deep geologic formations in Switzerland. The well known fact that the data basis is extremely sparse is pointed out once more and concretized in detail. (Auth.)

  10. Extended probabilistic system assessment calculations within the SKI project-90

    The probabilistic system assessment calculation reported in the SKI Project-90 final documents were restricted to the following nuclides: 14C, 129I, 135Cs, 237Np and 240Pu. In this report we have extended those calculations to another five nuclides: 79Se, 243Am, 240Pu, 93Zr and 99Tc. The execution of probabilistic assessment calculations integrated in the context of SKIs first safety analysis exercise of an hypothetic final repository for high-level nuclear waste in Sweden, was a learning experience of relevance for the conduction of probabilistic safety assessment in future exercises. Some major conclusions and viewpoints of future need related with probabilistic assessment were withdrawn from this work and are presented in our report

  11. Neutron-induced cross sections of actinides via the surrogate-reaction method

    Tveten G. M.

    2013-03-01

    Full Text Available The surrogate-reaction method is an indirect way of determining cross sections for reactions that proceed through a compound nucleus. This technique may enable neutron-induced cross sections to be extracted for short-lived nuclei that otherwise cannot be measured. However, the validity of the surrogate method for extracting capture cross sections has to be investigated. In this work we study the reactions 238U(d,p239U, 238U(3He,t238Np, 238U(3He,4He237U as surrogates for neutroninduced reactions on 238U, 237Np and 236U, respectively, for which good quality data exist. The experimental set-up enabled the measurement of fission and gamma-decay probabilities. First results are presented and discussed.

  12. Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes

    This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of 241Am and 237Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the 241Am isomeric branching ratio were precisely measured at thermal energy point. (author)

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

    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; Gascó, 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; Tarján, 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 (40K, 137Cs, 232Th, 234U, 235U, 238U, 238Pu, 239+240Pu and 241Am) were certified for this material. Information on massic activities with 95% confidence intervals is given for six other radionuclides (90Sr, 210Pb(210Po), 226Ra, 239Pu, 240Pu 241Pu). Less frequently reported radionuclides (99Tc, 129I, 228Th, 230Th and 237Np) 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. PMID:16549351

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

    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 of radionuclides in water samples, for the development and validation of analytical methods and for training purposes. The material is available in 5 L units from IAEA (http://nucleus.iaea.org/rpst/index.htm). (author)

  15. Transmutation studies using SSNTD and radiochemistry and the associated production of secondary neutrons

    Brandt, R; Wan, J S; Schmidt, T; Langrock, E J; Vater, P; Adam, J; Bamblevski, V P; Bradnova, V; Gelovani, L K; Kalinnikov, V K; Krivopustov, M I; Kulakov, B A; Sosnin, A N; Perelygin, V P; Pronskikh, V S; Stegailov, V I; Tsoupko-Sitnikov, V M; Modolo, G; Odoj, R; Philippen, P W; Adloff, J C; Pape, F; Debeauvais, M; Zamani-Valassiadou, M; Hashemi-Nezhad, S R; Dwivedi, K K; Guo Shi Lun; Li, L; Wang, Y L; Wilson, B

    1999-01-01

    Experiments using 1.5 GeV, 3.7 GeV and 7.4 GeV protons from the Synchrophasotron, LHE, JINR, Dubna, Russia, on extended Pb- and U- targets were carried out using SSNTD and radiochemical sensors for the study of secondary neutron $9 fluences. We also carried out first transmutation studies on the long-lived radwaste nuclei /sup 129/I and /sup 237/Np. In addition, we carried out computer code simulation studies on these systems using LAHET and DCM/CEM codes. We $9 have difficulties to understand rather large transmutation rates observed experimentally when they are compared with computer simulations. There seems to be a rather fundamental problem understanding the large transmutation rates as $9 observed experimentally in Dubna and CERN, as compared to those theoretical computer simulations mentioned above. (10 refs).

  16. TANDEM. A mutual cooperation effort for transactinide nuclear data evaluation and measurement

    The need for accurate nuclear reaction data of actinides is well documented and several initiatives from international organizations for improvement have been initiated in the past. This need, particularly in view of method development for non-destructive assay of nuclear waste, has generated a joint effort to use prompt and delayed neutron activation techniques to enhance nuclear capture data of some long lived actinides such as 237Np, 242Pu and 241Am in the frame of a multilateral cooperation. This research initiative is targeted to lay grounds for the development of a non-destructive active neutron interrogation technique to quantify actinides in mixed waste and residues from decommissioning of nuclear installations for safe treatment and storage of such materials. (author)

  17. Current activities and future plans for nuclear data measurements at J-PARC

    In order to improve the data accuracy of neutron-capture cross-sections of minor actinides (MAs) and long-lived fission products (LLFPs), a new experimental instrument named ''Accurate Neutron-Nucleus Reaction measurement Instrument'' (ANNRI) has been constructed in the Materials and Life science experimental Facility (MLF) at the Japan Proton Accelerator Research Complex (J-PARC), and measurements of neutron-capture cross-sections of MAs, LLFPs and some stable isotopes with high-intensity pulsed neutrons have been started. The analyses for 244Cm, 246Cm, 241Am and 237Np were finished; those for 129I, 107Pd, 99Tc, 93Zr and some stable isotopes are in progress. These results will give significant contributions in the field of developing innovative nuclear systems. (orig.)

  18. Improved radioanalytical method for the simultaneous determination of Th, U, Np, Pu and Am(Cm) on a single TRU column by alpha spectrometry and ICP-MS

    A radioanalytical method based on the use of a single TRU extraction chromatographic column and selective, on-column oxidation state adjustment of actinides was developed for the determination of Th, U, Np, Pu and Am(Cm) in environmental samples (such as sediment and swipe samples). The procedure of Vajda et al. was further investigated and optimized focusing on the separation of Th, U and Np. The improved method combines two measurement techniques - alpha spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) - which allows to obtain more reliable information on a wider range of isotopes: 228Th, 230Th, 232Th, 234U, 235U, (236U), 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 241Am, 242Cm, 243Cm and 244Cm from one sample. (orig.)

  19. Critical comparison of radiometric and mass spectrometric methods for the determination of radionuclides in environmental, biological and nuclear waste samples.

    Hou, Xiaolin; Roos, Per

    2008-02-11

    The radiometric methods, alpha (alpha)-, beta (beta)-, gamma (gamma)-spectrometry, and mass spectrometric methods, inductively coupled plasma mass spectrometry, accelerator mass spectrometry, thermal ionization mass spectrometry, resonance ionization mass spectrometry, secondary ion mass spectrometry, and glow discharge mass spectrometry are reviewed for the determination of radionuclides. These methods are critically compared for the determination of long-lived radionuclides important for radiation protection, decommissioning of nuclear facilities, repository of nuclear waste, tracer application in the environmental and biological researches, these radionuclides include (3)H, (14)C, (36)Cl, (41)Ca, (59,63)Ni, (89,90)Sr, (99)Tc, (129)I, (135,137)Cs, (210)Pb, (226,228)Ra, (237)Np, (241)Am, and isotopes of thorium, uranium and plutonium. The application of on-line methods (flow injection/sequential injection) for separation of radionuclides and automated determination of radionuclides is also discussed. PMID:18215644

  20. Application of high resolution inductively coupled plasma mass spectrometry to the measurement of long-lived radionuclides

    The radioactivity determination of long-lived radionuclides in environmental samples is difficult due to their low concentration and low specific activity. Limits of detection (DL) for long-lived radionuclides (99Tc, 226Ra, 232Th, 237Np, 238U, 243Am) in standard solutions using Inductively Coupled Plasma Mass Spectrometry with a double focusing mass analyzer (HR-ICP-MS) with ultrasonic nebulizer were calculated. DL of most nuclides were under the fg g-1 level. It seems that the analysis by using HR-ICP-MS has the advantage of detection for low level radio nuclides whole half life is more than 1000 y. This method was applied to 226Ra analysis in hot spring waters and 230Th in lake sediments analysis after a simple chemical separation. (author)

  1. Chemical speciation of long-lived radionuclides in the environment

    This project started in November 2005 and ended in November 2008, the work and research approaches are summarized in this report. This project studied the speciation of radionuclides in environment. A number of speciation analytical methods are developed for determination of species of 129I, 99Tc, isotopes of Pu, and 237Np in seawater, fresh water, soil, sediment, vegetations, and concrete. The developed methods are used for the investigation of the chemical speciation of these radionuclides as well as their environmental behaviours, especially in Danish environment. In addition the speciation of Pu isotopes in waste samples from the decommissioning of Danish nuclear facilities is also investigated. The report summarizes these works completed in this project. Through this research project, a number of research papers have been published in the scientific journals, the research results has also been presented in the Nordic and international conference/meeting and communicated to international colleagues. Some publications are also enclosed to this report. (au)

  2. Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes; Mesures des sections efficaces de capture et potentiels d'incineration des actinides mineurs dans les hauts flux de neutrons: Impact sur la transmutation des dechets

    Bringer, O

    2007-10-15

    This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of {sup 241}Am and {sup 237}Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the {sup 241}Am isomeric branching ratio were precisely measured at thermal energy point. (author)

  3. Diffusion of Am, Pu, U, Np, Cs, I and Tc in compacted sand-bentonite mixture

    In order to predict the diffusion of actinides and fission products through a backfill mixture of sand and clay from a high-level waste repository, the diffusion of the actinides 241Am, 239Pu, 237Np and 233U and the fission products 134Cs, 131I and 99mTc have been measured in a mixture of 90% silica sand-10% bentonite (MX-80, Wyoming bentonite). The sand-bentonite mixture was compacted to a density of 2000 kg/m3. The water phase used was an artificial groundwater representative of Swedish deep granitic groundwater (pH∼8, I∼0.01). The apparent diffusivity is in all cases slightly higher than in pure clay. (author)

  4. Calculated medium energy fission cross sections

    An analysis has been made of medium-energy nucleon induced fission of 238U and 237Np using detailed models of fission, based upon the Bohr-Wheeler formalism. Two principal motivations were associated with these calculations. The first was determination of barrier parameters for proton-rich uranium and neptunium isotopes normally not accessible in lower energy reactions. The second was examination of the consistency between (p,f) experimental data versus new (n,f) data that has recently become available. Additionally, preliminary investigations were also made concerning the effect of fission dynamics on calculated fission cross sections at higher energies where neutron emission times may be significantly less than those associated with fission

  5. Monte Carlo modeling of spallation targets containing uranium and americium

    Neutron production and transport in spallation targets made of uranium and americium are studied with a Geant4-based code MCADS (Monte Carlo model for Accelerator Driven Systems). A good agreement of MCADS results with experimental data on neutron- and proton-induced reactions on 241Am and 243Am nuclei allows to use this model for simulations with extended Am targets. It was demonstrated that MCADS model can be used for calculating the values of critical mass for 233,235U, 237Np, 239Pu and 241Am. Several geometry options and material compositions (U, U + Am, Am, Am2O3) are considered for spallation targets to be used in Accelerator Driven Systems. All considered options operate as deep subcritical targets having neutron multiplication factor of k∼0.5. It is found that more than 4 kg of Am can be burned in one spallation target during the first year of operation

  6. Radionuclide desorption kinetics on synthetic Zn/Ni-labeled montmorillonite nanoparticles

    Huber, F. M.; Heck, S.; Truche, L.; Bouby, M.; Brendlé, J.; Hoess, P.; Schäfer, T.

    2015-01-01

    Sorption/desorption kinetics for selected radionuclides (99Tc(VII), 232Th(IV), 233U(VI), 237Np(V), 242Pu and 243Am(III)) under Grimsel (Switzerland) ground water conditions (pH 9.7 and ionic strength of ∼1 mM) in the presence of synthetic Zn or Ni containing montmorillonite nanoparticles and granodiorite fracture filling material (FFM) from Grimsel were examined in batch studies. The structurally bound Zn or Ni in the octahedral sheet of the synthetic colloids rendered them suitable as colloid markers. Only a weak interaction of the montmorillonite colloids with the fracture filling material occurs over the experimental duration of 10,000 h (∼13 months). The tri- and tetravalent radionuclides are initially strongly associated with nanoparticles in contrast to 99Tc(VII), 233U(VI) and 237Np(V) which showed no sorption to the montmorillonite colloids. Radionuclide desorption of the nanoparticles followed by sorption to the fracture filling material is observed for 232Th(IV), 242Pu and 243Am(III). Based on the conceptual model that the driving force for the kinetically controlled radionuclide desorption from nanoparticles and subsequent association to the FFM is the excess in surface area offered by the FFM, the observed desorption kinetics are related to the colloid/FFM surface area ratio. The observed decrease in concentration of the redox sensitive elements 99Tc(VII), 233U(VI) and 237Np(V) may be explained by reduction to lower oxidation states in line with Eh-pH conditions prevailing in the experiments and thermodynamic considerations leading to (i) precipitation of a sparingly soluble phase, (ii) sorption to the fracture filling material, (iii) possible formation of eigencolloids and/or (iv) sorption to the montmorillonite colloids. Subsequent to the sorption/desorption kinetics study, an additional experiment was conducted investigating the potential remobilization of radionuclides/colloids attached to the FFM used in the sorption/desorption kinetic

  7. The simultaneous neutron and photon interrogation method for fissile and non-fissile element separation in radioactive waste drums

    Measuring α-emitters such as (234,235,236,238U, 238,239,240,242,244Pu, 237Np, 241,243Am, ...), in solid radioactive waste allows us to quantify the α-activity in a drum and then to classify it. The simultaneous photon and neutron interrogation experiment (SIMPHONIE) method dealt with in this paper, combines both active neutron interrogation and induced photofission interrogation techniques simultaneously. Its purpose is to quantify fissile (235U, 239,241Pu, ...) and non-fissile (236,238U, 238,240Pu, ...) elements separately in only one measurement. This paper presents the principle of the method, the experimental setup, and the first experimental results obtained using the DGA/ETCA Linac and MiniLinatron pulsed linear electron accelerators located at Arcueil, France. First studies were carried out with U and Pu bare samples

  8. Determination of fission cross-section and absolute fission yields using track-cum gamma-ray spectrometric technique

    The fission cross-section of 233Pa(2nth, f) using fission track technique has been determined for the first time using thermal neutron flux of the reactor APSARA. This is important from the point of view of advance heavy water reactor (AHWR), which is to be described. On the other hand, the yields of fission products in the fast neutron induced fission of minor actinides are important from the point accelerator driven sub critical system (ADSS). In view of that, absolute yields of fission products in the fast neutron induced fission of 238U, 237Np, 238,240Pu, 243Am and 244Cm have been determined using the fission track-cum gamma-ray spectrometric technique. The total number of fission occurring in the target was estimated by track technique, whereas the activities of the fission products have been determined using gamma-ray spectrometric technique. Detailed procedure and its importance are to be discussed. (author)

  9. Studies of heavy ion reactions and transuranic nuclei. Progress report, August 1, 1979-July 31, 1980

    The study of heavy-ion reaction mechanisms at the SuperHILAC and LAMPF is reported. Preprints of five articles and manuscripts of four recent conference papers are given, along with complete citations of publications and a list of personnel. Significant work was performed in the following areas: the bombarding energy dependence of the 209Bi + 136Xe reaction; the fragment yields for specific Z and A for projectile-like fragments produced in the reaction of 8.3-MeV/u 56Fe ions with targets of 56Fe, 165Ho, 209Bi, and 238U; and time distributions of fragments from delayed fission after muon capture for muonic 235U, 238U, 237Np, 239Pu, and 242Pu

  10. Preparation of americium targets for nuclear chemistry experiments at DANCE

    Using 1 gram of 241Am from LANL stocks, the purification steps required to obtain a solution of 241Am from the original material are described. Part of the purified solution was submitted for purity analysis by mass spectrometry, radiochemistry and trace metals analysis. The impurities were expected to be 239Pu and 237Np. A second fraction of this material was used for electroplating three samples onto titanium disks that were suitable for insertion into an instrument package to be placed into the DANCE detector. The purification methods used, the electroplating setup and the solutions to various problems that were encountered in making these targets are discussed. The analytical results are discussed as well as the yields from the electrodeposition process. Comparison of these yields with those from similar experiments utilizing 235U and 243Am are also discussed. (author)

  11. AN INTEGRAL REACTOR PHYSICS EXPERIMENT TO INFER ACTINIDE CAPTURE CROSS-SECTIONS FROM THORIUM TO CALIFORNIUM WITH ACCELERATOR MASS SPECTROMETRY

    The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectroscopy (AMS) technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am and 248Cm.

  12. MANTRA: An Integral Reactor Physics Experiment to Infer Actinide Capture Cross-sections from Thorium to Californium with Accelerator Mass Spectrometry

    The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectrometry technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm and 248Cm.

  13. Preparation and characterization of 234U for mass spectrometry and alpha-spectrometry

    234U of high isotopic purity (>99 atom%) as well as of high radiochemical purity was separated from aged 238Pu prepared by neutron irradiation of 237Np. Methodologies based on ion exchange and solvent extraction procedures were used to achieve high decontamination factor from 238Pu owing to the very high α-specific activity of 238Pu (2800 times) in comparison to that of 234U. Isotopic composition of purified 234U was determined by thermal ionisation mass spectrometry. Alpha spectrometry was used for checking the radiochemical purity of 234U with respect to concomitant α-emitting nuclides. The separated 234U will be useful for different investigations using mass spectrometry and alpha spectrometry. (author)

  14. Experimental and computational benchmark tests

    A program involving principally NIST, LANL, and ORNL has been in progress for about four years now to establish a series of benchmark measurements and calculations related to the moderation and leakage of 252Cf neutrons from a source surrounded by spherical aqueous moderators of various thicknesses and compositions. The motivation for these studies comes from problems in criticality calculations concerning arrays of multiplying components, where the leakage from one component acts as a source for the other components. This talk compares experimental and calculated values for the fission rates of four nuclides - 235U, 239Pu, 238U, and 237Np - in the leakage spectrum from moderator spheres of diameters 76.2 mm, 101.6 mm, and 127.0 mm, with either pure water or enriched B-10 solutions as the moderator. Very detailed Monte Carlo calculations were done with the MCNP code, using a open-quotes light waterclose quotes S(α,β) scattering kernel

  15. Sequential separation of transuranic elements and fission products from uranium metal ingots in electrolytic reduction process of spent PWR fuels

    A sequential separation procedure has been developed for the determination of transuranic elements and fission products in uranium metal ingot samples from an electrolytic reduction process for a metallization of uranium dioxide to uranium metal in a medium of LiCl-Li2O molten salt at 650 deg C. Pu, Np and U were separated using anion-exchange and tri-n-butylphosphate (TBP) extraction chromatography. Cs, Sr, Ba, Ce, Pr, Nd, Sm, Eu, Gd, Zr and Mo were separated in several groups from Am and Cm using TBP and di(2-ethylhexyl)phosphoric acid (HDEHP) extraction chromatography. Effect of Fe, Ni, Cr and Mg, which were corrosion products formed through the process, on the separation of the analytes was investigated in detail. The validity of the separation procedure was evaluated by measuring the recovery of the stable metals and 239Pu, 237Np, 241Am and 244Cm added to a synthetic uranium metal ingot dissolved solution. (author)

  16. Current activities and future plans for nuclear data measurements at J-PARC

    Kimura, Atsushi; Harada, Hideo; Nakamura, Shoji; Iwamoto, Osamu; Toh, Yosuke; Koizumi, Mitsuo; Kitatani, Fumito; Furutaka, Kazuyoshi [Japan Atomic Energy Agency, Ibaraki (Japan); Igashira, Masayuki; Katabuchi, Tatsuya; Mizumoto, Motoharu [Tokyo Institute of Technology, Research Laboratory for Nuclear Reactors, Tokyo (Japan); Hori, Jun-ichi [Kyoto University, Research Reactor Institute, Osaka (Japan); Kino, Koichi [Hokkaido University, Faculty of Engineering, Sapporo (Japan); Kiyanagi, Yoshiaki [Nagoya University, Graduate School of Engineering, Nagoya (Japan)

    2015-12-15

    In order to improve the data accuracy of neutron-capture cross-sections of minor actinides (MAs) and long-lived fission products (LLFPs), a new experimental instrument named ''Accurate Neutron-Nucleus Reaction measurement Instrument'' (ANNRI) has been constructed in the Materials and Life science experimental Facility (MLF) at the Japan Proton Accelerator Research Complex (J-PARC), and measurements of neutron-capture cross-sections of MAs, LLFPs and some stable isotopes with high-intensity pulsed neutrons have been started. The analyses for {sup 244}Cm, {sup 246}Cm, {sup 241}Am and {sup 237}Np were finished; those for {sup 129}I, {sup 107}Pd, {sup 99}Tc, {sup 93}Zr and some stable isotopes are in progress. These results will give significant contributions in the field of developing innovative nuclear systems. (orig.)

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

    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 (4K, 137Cs, 232Th, 234U, 235U, 238U, 238Pu, 239+24Pu and 241Am) were certified for this material. Information on massic activities with 95% confidence intervals is given for six other radionuclides (9Sr, 21Pb(21Po), 226Ra, 239Pu, 24Pu 241Pu). Less frequently reported radionuclides (99Tc, 129I, 228Th, 23Th and 237Np) 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

  18. Updated multi-group cross sections of minor actinides with improved resonance treatment

    The study of minor actinide in transmutation reactors and other future applications makes resonance self-shielding treatment a significant issue for criticality and isotope depletion. Resonance treatment for minor actinides has been carried out by subgroup method with improved interference effect through interference correction. Subgroup data was generated using RMET21 and GENP codes along with multi-group cross section data by NJOY nuclear data processing system. Updated multi-group cross section data library for a neutron transport code nTRACER was compared with solutions from MCNPX. The resonance interaction of uranium with minor actinides has been included by modified interference treatment of interference correction in subgroup methodology. The comparison of cross sections and multiplication factor in pin and assembly problems showed significant improvement from systematic resonance treatment especially for 237Np and 243Am. (author)

  19. Fabrication and characterization of MCC [Materials Characterization Center] approved testing material---ATM-2, ATM-3, and ATM-4 glasses

    Materials Characterization Center glasses ATM-2, ATM-3, and ATM-4 are designed to simulate high-level waste glasses that are likely to result from the reprocessing of commercial nuclear reactor fuels. The three Approved Testing Materials (ATMs) are borosilicate glasses based upon the MCC-76-68 glass composition. One radioisotope was added to form each ATM. The radioisotopes added to form ATM-2, ATM-3, and ATM-4 were 241Am, 237Np, and 239Pu, respectively. Each of the ATM lots was produced in a nominal lot size of 450 g from feed stock melted in a nitrogen-atmosphere glove box at 1200/degree/C in a platinum crucible. Each ATM was then cast into bars. Analyzed compositions of these glasses are listed. The nonradioactive elements were analyzed by inductively coupled argon plasma atomic emission spectroscopy (ICP), and the radioisotope analyses were done by alpha energy analysis. Results are discussed. 7 refs., 3 figs., 5 tabs

  20. Photon and proton induced fission on heavy nuclei at intermediate energies

    We present an analysis of fission induced by intermediate energy protons or photons on actinides. The 660 MeV proton induced reactions are on 241Am, 238U, and 237Np targets and the Bremsstrahlung-photons with end-point energies at 50 MeV and 3500 MeV are on 232Th and 238U targets. The study was performed by means of the Monte Carlo simulation code CRISP. A multimodal fission extension was added to the code within an approach which accounts for the contribution of symmetric and asymmetric fission. This procedure allowed the investigation of fission cross sections, fissility, number of evaporated nucleons and fission-fragment charge distributions. The comparison with experimental data show a good agreement between calculations and experiments. (author)

  1. Neutron-induced cross sections of actinides via the surrogate-reaction method

    The surrogate-reaction method is an indirect way of determining cross sections for reactions that proceed through a compound nucleus. This technique may enable neutron-induced cross sections to be extracted for short-lived nuclei that otherwise cannot be measured. However, the validity of the surrogate method has to be investigated. In particular, the absence of a compound nucleus formation and the Jπ dependence of the decay probabilities may question the method. In this work we study the reactions 238U(d,p) 239U, 238U(3He,t)238Np, 238U(3He,4He)237U as surrogates for neutron-induced reactions on 238U, 237Np and 236U, respectively, for which good quality data exist. The experimental set-up enabled the measurement of fission and gamma-decay probabilities. The first results are hereby presented. (authors)

  2. Measurements of fast neutron-induced fission data of Np-237

    Win, Than; Saito, Keiichiro; Baba, Mamoru; Iwasaki, Tomohiko; Ibaraki, Masanobu; Miura, Takako; Sanami, Toshiya; Nauchi, Yasushi; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan). Faculty of Engineering

    1998-03-01

    We have performed the following measurements for {sup 237}Np using the 4.5 MV Dynamitron accelerator of Tohoku University as the pulsed neutron source: (1) Prompt fission neutron spectrum for 0.62 MeV incident neutrons, and (2) Neutron-Induced fission cross-section between 10 and 100 keV. The prompt fission neutron spectrum was measured using TOF method with a heavily shielded NE213 scintillation detector. The Maxwellian temperature T{sub m} derived is 1.28 MeV, which is lower than that of 1.38 MeV in JENDL-3.2. The fission cross sections were measured between 10 - 100 keV. The results are between JENDL-3.2 and ENDF/B-VI. (author)

  3. Communication of nuclear data progress: No.6 (1991)

    > (CNDP) in English is set up by Chinese Nuclear Committee and Chinese Nuclear Data center (CNDC). This is the sixth issue, in which the first part of CENDL-2 (Chinese Evaluated Nuclear Data Library, version 2.0) papers is published. The main purpose for making CENDL-2 is to develop and improve the CENDL-1. It includes the evaluation of 38 elements and isotopes with incident neutron energy from 10-5 eV to 20 MeV, which are 1H, 2H, 3He, 4He, 6Li, 7Li, 9Be, 10B, 11B, 14N, 16O, 19F,, Mg, Al, Si, S, K, Ti, V, Cr, Fe, Ni, Cu, Zr, Nb, Mo, Ag, 107Ag, 109Ag, Sn, Sb, 181Ta, W, 197Au, Pb, 238U, 237Np. The other evaluation reports of 16 elements and isotopes will be published in the supplement to CNDP No.6

  4. The importance of the giant resonances in hadron and muon induced fission

    In the first part of the thesis the fission probability of 238U by means of the reaction 238U(α,α'f) is studied at an incident energy of 480 MeV and a scattering angle of 3.40. In the measured spectrum of the inelastically scattered α particles a strong resonance is found in the excitation energy range from 8 to 13 MeV. The center of mass of the resonance lies at 11 MeV. Its width extends to 4.5 MeV. In the second part of the thesis the muon induced fission of 235U, 238U, 237Np, 242Pu, and 244Pu is studied. Thereby both fission fragments are detected in coincidence by two surface barrier detectors. By this it is possible for the first time to measure the mass and kinetic energy distribution of the fission fragments. (orig./HSI)

  5. WWER-1000 spent fuel nuclide inventory at the Kozloduy NPP

    This paper contains a presentation and discussion of selected isotope inventory results for different types of WWER-1000 spent fuel assemblies. Nuclide inventory calculations of spent fuel assemblies at the Kozloduy NPP are routinely performed using the SCALE 4.4a computer code system. Besides the standard 17x17 ORIGEN-S library, a specific library developed at the Kozloduy NPP for each different fuel type at typical irradiation conditions is used. The evaluated concentrations of the most important isotopes - 235U, 236U, 238U, 239Pu, 240Pu, 241Pu, 242Pu - depending on burnup, are compared with fuel supplier data and with data calculated using the HELIOS-1.5 lattice code. The buildup of some other isotopes (237Np, 241Am, 243Am) is presented, too. (authors)

  6. WWER-1000 spent fuel nuclide inventory at the Kozloduy NPP

    This paper contains a presentation and discussion of selected isotope inventory results for different types of WWER-1000 spent fuel assemblies. Nuclide inventory calculations of spent fuel assemblies at Kozloduy NPP are routinely performed using the SCALE 4.4a computer code system. Besides the standard 17x17 ORIGEN-S library, a specific library developed at the Kozloduy NPP for each different fuel type at typical irradiation conditions is used. The evaluated concentrations of the most important isotopes-235U, 236U, 238U, 239Pu, 240Pu, 241Pu, 242Pu-depending on burnup, are compared with fuel supplier data and with data calculated using the HELIOS-1.5 lattice code. The buildup of some other isotopes (237Np, 241Am, 243Am) is presented, too (Authors)

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

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

  8. Separation of 230Th (ionium) from uranium ores sulfuric acid and in nitric acid solutions

    230Th (ionium) is of interest for production of 231Pa, the Pa isotope with the longest half life, and for production of 232U which can be used in radionuclide batteries. Two procedures are presented which have been worked out for separation of 230Th from sulfuric acid and from nitric acid solutions. In the first case the effluents from the anion exchange resins are the starting material for recorvery of 230Th, in the second case the cation exchange resins which are for purification of U. The procedures selected are simple, economic and can be performed in any uranium mill or uranium purification plant without additional investments. (orig.)

  9. Total half-lives for selected nuclides

    Measurements of the half-lives of 3H, 10Be, 14C, 26Al, 40K, 39Ar, 53Mn, 87Rb, 92Nb, 129I, 138La, 147Sm, 176Lu, 174Hf, 180Ta, 187Re, 186Os, 190Pt, 204Pb, 210Pb, 210Po, 222Rn, 224Th, 226Ra, 227Ac, 228Ra, 228Th, 230Th, 232Th, 231Pa have been compiled and evaluated. The effect of the 14C half-life value on carbon dating ages is discussed as well as the stability of 204Pb. 237 refs., 30 tabs

  10. γ-ray spectrometry investigation of radioactive equilibrium in ancient uranium ores of Ukraine

    The nuclear-physics technique of determining the radioactive equilibrium. The degree of deviations from the radioactive equilibrium strongly correlates with the mineral content of ores. The possible models of loss/import of radioactive nuclides are investigated. As the more probable the complex model with the loss of 231Pa, 234U and 230Th and further secondary precipitation of 234U is examined. The geologically young ages of the losses of radionuclides were estimated to be equal to the times of the water level lifting for the periods of interglacial temperature optimums

  11. A review of the current status of nuclear data for major and minor isotopes of thorium fuel cycle

    In this paper, we present a critical overview of the status of the available nuclear data of isotopes of thorium fuel cycle, viz., 230Th, 232Th, 231Pa, 233Pa, 232U, 233U and 234U. Induced in the main body of the paper is a critical analysis of information contained in the two basic evaluated nuclear data files JENDL-3.2 and ENDF/B-VI (Rev.5) recently released by the IAEA/NDS as a result of truly international efforts. In some of the cases, the information and data given in EXFOR is examined to get an idea of the status of measured nuclear data of these isotopes. Some comments regarding gaps in experimental data as of 1999 are included in the discussion. Most of these experimental data were those generated two decades ago. In addition, generally, these experimental data are very limited in comparison to the voluminous nuclear data generated for the uranium-plutonium cycle. Experimental data is absent in most of the cases and, in such cases, evaluated cross sections in the two basic evaluated nuclear data files JENDL-3.2 and ENDF/B-VI(Rev.5) are based upon theoretical models and nuclear systematics. Some of these differences between JEF-2.2 and its source ENDF/B-V that were carried over to ENDF/B-VI(Rev.5) are explained. The role and the importance of 231Pa and 233Pa in the thorium fuel cycle in advanced concepts such as the Energy Amplifier are mentioned. New calculations of criticality property of 231Pa and 233Pa are presented using the neutron reaction data of JENDL-3.2 and ENDF/B-VI(Rev.5). The possible influence of 230Th is examined with respect to its cross sections and production of 231Pa in a typical Indian PHWR environment. The quality assurance in design and safety studies in nuclear energy in the next few decades and centuries require new and improved data with high accuracy and energy resolution. As a starter, the nuclear data of the set of isotopes of thorium fuel cycle discussed in this paper is a challenging sample for consideration as a trial project

  12. Critical experiment for large fast reactor at FCA XI-1 assembly

    The present report compiled the experimental results of sample worths, fission rates, Na void effects and B4C control rod reactivity effects. Preliminary analysis on some of the results was also made using 70 group constant set and the diffusion and transport theory code, and presented in this report. It was shown with the measured data that the test region of the assembly has the most softened neutron sepectrum compared with that of the previons FCA assemblies and well simulates the physics aspects of the large fast reactor core. Close agreements were obtained between the calculated and measured data for central sample worths. The calculated values of axial sample worth distributions, however, deviated to some extent from those of measured distributions. Concerning the axial fission rate distributions of 238U, 237Np, 239Pu and 235U, large discrepancies between the calculated and measured distributions were found in the axial blanket region. Transport calculations could imporve the result obtained by diffusion calculation for 238U fission rate distribution. The calculated central fission rate ratio of 238U/235U failed to predict the measured ratio, while calculated ratios of 237Np/235U and 239Pu/235U well agreed to these measured ratios. The calculated Na void effect with use of the first order perturbation method overestimated the measured effects by about 18% in the central part of the test region of the assembly. No significant discrepancy in the shape of axial distribution, however, was found between the calculated and measured Na void effects. B4C control rod worth was measured at core center region using mock-up control rod which consist of B4C pellet and SS tube. The effects of control rod worth were examined for 10B enrichment, 10B density and 10B distribution condition. (author)

  13. Effect of Reducing Groundwater on the Retardation of Redox-Sensitive Radionuclides

    Hu, Q; Zavarin, M; Rose, T P

    2008-04-21

    Laboratory batch sorption experiments were used to investigate variations in the retardation behavior of redox-sensitive radionuclides. Water-rock compositions used during these experiments were designed to simulate subsurface conditions at the Nevada Test Site (NTS), where a suite of radionuclides were deposited as a result of underground nuclear testing. Experimental redox conditions were controlled by varying the oxygen content inside an enclosed glove box and by adding reductants into the testing solutions. Under atmospheric (oxidizing) conditions, the radionuclide distribution coefficients varied with the mineralogical composition of the sorbent and the water chemistry. Under reducing conditions, distribution coefficients showed marked increases for {sup 99}Tc and {sup 237}Np in devitrified tuff, but much smaller variations in alluvium, carbonate rock, and zeolitic tuff. This effect was particularly important for {sup 99}Tc, which tends to be mobile under oxidizing conditions. Unlike other redox-sensitive radionuclides, iodine sorption may decrease under reducing conditions when I{sup -} is the predominant species. Overall, sorption of U to alluvium, devitrified tuff, and zeolitic tuff under atmospheric conditions was less than in the glove-box tests. However, the mildly reducing conditions achieved here were not likely to result in substantial U(VI) reduction to U(IV). Sorption of Pu was not affected by the decreasing redox conditions achieved in this study, as the predominant sorbed Pu species in all conditions was expected to be the low-solubility and strongly sorbing Pu(OH){sub 4}. Depending on the aquifer lithology, the occurrence of reducing conditions along a groundwater flowpath could potentially contribute to the retardation of redox-sensitive radionuclides {sup 99}Tc and {sup 237}Np, which are commonly identified as long-term dose contributors in the risk assessment in various nuclear facilities.

  14. Preconceptual Feasibility Study to Evaluate Alternative Means to Produce Plutonium-238

    John D. Bess; Matthew S. Everson

    2013-02-01

    There is currently no large-scale production of 238Pu in the United States. Feasibility studies were performed at the Idaho National Laboratory to assess the capability of developing alternative 238Pu production strategies. Initial investigations indicate potential capability to provision radioisotope-powered systems for future space exploration endeavors. For the short term production of 238Pu, sealed canisters of dilute 237Np solution in nitric acid could be irradiated in the Advanced Test Reactor (ATR). Targets in the large and medium “I” positions of the ATR were irradiated over a simulated period of 306 days and analyzed using MCNP5 and ORIGEN2.2. Approximately 0.5 kg of 238Pu could be produced annually in the ATR with purity greater than 92%. Optimization of the irradiation cycles could further increase the purity to greater than 98%. Whereas the typical purity of space batteries is between 80 to 85%, the higher purity 238Pu produced in the ATR could be blended with existing lower-purity inventory to produce useable material. Development of irradiation methods in the ATR provides the fastest alterative to restart United States 238Pu production. The analysis of 238Pu production in the ATR provides the technical basis for production using TRIGA® (Training, Research, Isotopes, General Atomics) nuclear reactors. Preliminary analyses envisage a production rate of approximately 0.7 kg annually using a single dedicated 5-MW TRIGA reactor with continuous flow loops to achieve high purity product. Two TRIGA reactors represent a robust means of providing at over 1 kg/yr of 238Pu annually using dilute solution targets of 237Np in nitric acid. Further collaboration and optimization of reactor design, radiochemical methods, and systems analyses would further increase annual 238Pu throughput, while reducing the currently evaluated reactor requirements.

  15. Rapid and simultaneous determination of neptunium and plutonium in environmental samples using anion exchange chromatographic and sequential injection setup combined with inductively coupled plasma mass spectrometry

    Full text: This paper presents an automated analytical method for the rapid and simultaneous determination of Pu and Np in the environmental samples. Anion exchange chromatographic column was incorporated in a sequential injection system to actualize the automated separation of Pu isotpes along with 237Np from the matrix elements and interfering radionuclides. K2S2O5-conc. HNO3 was applied as redox reagents for the valence adjustment and stabilization of Pu(IV) and Np(IV). 242Pu preformed well as a tracer for both Pu isotopes and 237Np. It was observed that the cross-link and particle size of the resins had significant effluence on the separation efficiency and anion exchange resin Bio-Rad AG 1 x 4 with the particle size of 100-200 mesh was chosen as the optimum. The investigation on the capacity showed small-sized column packed with 2mL resin sufficed up to 50g of soil sample, which provides an advantage of low consumption of the resin and low generation of acid waste after the column washing. The analytical results for Pu and Np in three reference materials showed good agreement with the certified or reference values at the 0.05 significance level. Chemical yields of Pu and Np equally range from 80% to 100%, and the decontamination factors for uranium, thorium and lead were in the range of 103 to 104. The total time of separation for a single sample was < 2.5 hours, which extremely improve the analysis efficiency and reduces the labor intensity, as well as enables a rapid determination of Pu and Np in emergency situations. (author)

  16. Studies on the reaction mechanism of the muon induced nuclear fission

    The mass and energy distribution of the fission fragments after muon induced nuclear fission allows the determination of the mean excitation energy of the fissioning nucleus after muon capture. By the systematic comparison with a mass distribution of a corresponding reaction for the first time for this an accuracy of about 1 MeV could be reached. Theoretical calculations on the excitation probability in the muon capture allow in connection with the fission probability an estimating calculation of this energy. The experimental result represents by this a test criterium for the valuation of the theoretical calculation. The measured probabilities for the occurrence of radiationless transitions in the muonic γ cascade of 237Np permit an indirect experimental determination of the barrier enhancement which causes the muon present during the fission process. The value found for this extends to 0.75+-0.1 MeV. A change of the mass distribution by the muon cannot be detected in the nuclides 235U, 237Np, and 242Pu studied here. Only the mean total kinetic energy of the fission products is reduced in these three nuclides in the prompt μ- induced fission by 1 to 2 MeV. For this result the incomplete screening of the nuclear charge during the fission process is made responsible. A mass dependence of this reduction has not been stated. Because the muon has appearently no influence on the mass splitting it can be valied as nearly ideal particle in order to study the hitherto little studied dynamics of the fission process. (orig.)

  17. Seasonality in the flux of natural radionuclides and plutonium in the deep Sargasso Sea

    A record of radionuclide fluxes at a deep-ocean station near Bermuda was obtained from analysis of a 3-year collection of sediment-trap samples. The trap was placed at a depth of 3200 m, 1000 m above the sea floor, and the samples were recovered at 2-month intervals. Concentrations of 238U, 234U, 232Th, 230Th, 228Th, 231Pa, 210Pb, 210Po, and sup(239,240)Pu were measured in the trapped material. Most of the radionuclide activity was found in the 230Th and 231Pa, considering that most of their production occurs in the water column below the euphotic zone. Evidently the seasonal influence is transmitted downward by the varying particle flux so that radionuclide scavenging rates at depth, as well as at the surface, are affected. It is suggested that this could be brought about by seasonal variations in the flux of marine snow or in the rate of fecal-matter production in the deep-water column. (author)

  18. High fuel burn-up and nonproliferation in PWR-type reactor on the basis of modified Th-fuel

    Neutronics-physical characteristics of the fuel lattice of a PWR-type reactor cooled by light water and by a mixture of light and heavy water have been analyzed. Th-fuel containing an essential amount of 231Pa and 232U is used, which allows an increase in fuel burn-up by a factor of 2-5 compared with that of traditional oxide uranium fuel with light water. It is important to underline that this is attained under the negative coolant density reactivity effect using cross sections of 231Pa and 232U from the updated JENDL-3.2 nuclear library. This radical increase of fuel burn-up is accompanied by a small change of reactivity during fuel irradiation (K∞=1.1 / 1.0), that favorably affects safety parameters of the reactor operation. A considerable percentage of 232U in fuel, and consequently in U, is a strong barrier against the proliferation of such weapon nuclide as 233U. (authors)

  19. Deep water provenance and dynamics of the (de)glacial Atlantic meridional overturning circulation

    Lippold, Jörg; Gutjahr, Marcus; Blaser, Patrick; Christner, Emanuel; de Carvalho Ferreira, Maria Luiza; Mulitza, Stefan; Christl, Marcus; Wombacher, Frank; Böhm, Evelyn; Antz, Benny; Cartapanis, Olivier; Vogel, Hendrik; Jaccard, Samuel L.

    2016-07-01

    Reconstructing past modes of ocean circulation is an essential task in paleoclimatology and paleoceanography. To this end, we combine two sedimentary proxies, Nd isotopes (εNd) and the 231Pa/230Th ratio, both of which are not directly involved in the global carbon cycle, but allow the reconstruction of water mass provenance and provide information about the past strength of overturning circulation, respectively. In this study, combined 231Pa/230Th and εNd down-core profiles from six Atlantic Ocean sediment cores are presented. The data set is complemented by the two available combined data sets from the literature. From this we derive a comprehensive picture of spatial and temporal patterns and the dynamic changes of the Atlantic Meridional Overturning Circulation over the past ∼25 ka. Our results provide evidence for a consistent pattern of glacial/stadial advances of Southern Sourced Water along with a northward circulation mode for all cores in the deeper (>3000 m) Atlantic. Results from shallower core sites support an active overturning cell of shoaled Northern Sourced Water during the LGM and the subsequent deglaciation. Furthermore, we report evidence for a short-lived period of intensified AMOC in the early Holocene.

  20. Measurements of neutron-induced fission cross section of protactinium-231 from 0.1 eV to 10 keV with lead slowing-down spectrometer and at 0.0253 eV with thermal neutron facility

    Protactinium(Pa)-231 is one of the most interesting nuclei which are related to the production of 232U in the 232Th-233U fuel cycle. The fission phenomena, fission energy and mass distributions for the neutron-induced fission of 231Pa have been investigated. Although several measurements of the 231Pa(n, f) cross section have been reported at higher energies, the fission cross section has rarely been measured in the lower/resonance energy region. One of the reasons may be due to sub-barrier fission which results in a low fission cross section. In addition, 231Pa is a radioactive actinide element. Then, a pure sample and an intense neutron source are required to overcome the severe experimental conditions for the fission cross section measurement. The thermal neutron-induced fission cross section of 231Pa was measured by Ghiorso and Van Winkle using a fission chamber and by Wagemans et al. with a surface barrier detector to detect the fission fragments at a neutron guide of the ILL. Mughabghab shows the thermal neutron cross section in the literature. Leonard and Odegaarden made the cross section measurement at 20 energy points between 0.37 eV and 0.52 eV by use of a crystal spectrometer. No experimental datum has been obtained at energies between 0.0253 eV and 100 keV except for those measured by Leonard and Odegaarden. The evaluated fission cross section data appear in ENDF/B-VI, JENDL-3.2 and JEF-2.2 (which is essentially same as ENDF/B-VI). The ENDF/B-VI data are markedly discrepant from the JENDL-3.2 data in the relevant energy region, especially above about 10 eV. In the present study, the neutron-induced fission cross section of 231Pa has been measured in the range of 0.1 eV to 10 keV relative to that of the 235U(n, f) reaction by making use of a back-to-back (BTB) type double fission chamber and a lead slowing-down spectrometer driven by a 46 MeV electron linear accelerator (linac) of the Research Reactor Institute, Kyoto University (KURRI). Below 1 keV, the

  1. Separation of radionuclides using host-guest materials

    This thesis is focused on the development of complex procedure with using commercially available sorbents to separate anthropogenic actinides 237Np, 241Am, 238Pu and 239,240Pu and their determination in the liquid radioactive wastes by using alpha spectrometry. Abilities of using commercially available sorbent AnaLig(R)Pu-02 gel from IBC Advanced Technologies, Inc. were tested, this product belongs to host-guest materials based on molecular recognition technology. This material is capable of selectively capturing actinides in oxidation state IV. To adjust the oxidation state of 238Pu and 239,240Pu was used NaNO2. Pu(IV) forms in the medium of nitric acid complexes, [Np(NO3)6]-, which are captured on the column. For the second monitored radionuclide, neptunium is typical valence V, Np(V) in the concentrated nitric acid produces strong complexes, [Np(NO3)6]-, which are capable of the sorption on the column of AnaLig(R)Pu-02 gel. The most common state of americium in aqueous solutions is III. Whereas in this oxidation state, americium do not form complexes in 8 mol·dm-3 nitric acid is the result of the flow-through. On the base of experimental obtained results, solution of 0.1 mol·dm-3 NH4I in 9 mol·dm-3 HCl was selected for elution of plutonium. Neptunium was eluted from the column using 9 mol·dm-3 of HCl with addition of 0.5 cm3 TiCl3. Optimizing conditions for the separation procedure was performed by using model solution of radioactive waste which was prepared according to the chemical composition of radioactive concentrate from NPP Mochovce and NPP Bohunice. The effect of the concentration of Fe3+, the effect of the concentration of the HCl, the effect of the concentration of the solution of NH4I and the effect of the volume of this solution to the yields of 238Pu were studied. And also was studied the effect of 9 mol·dm-3 of HCl and the effect of volume of 15 % TiCl3 to the yields of 237Np. Sorbent DGA(R) Resin from EiChrom Technologies, LLC was used for

  2. The production of Neptunium-236g

    Radiochemical analysis of 237Np is important in a number of fields, such as nuclear forensics, environmental analysis and measurements throughout the nuclear fuel cycle. However analysis is complicated by the lack of a stable isotope of neptunium. Although various tracers have been used, including 235Np, 239Np and even 236Pu, none are entirely satisfactory. However, 236gNp would be a better candidate for a neptunium yield tracer, as its long half-life means that it is useable as both a radiometric and mass spectrometric measurements. This radionuclide is notoriously difficult to prepare, and limited in scope. In this paper, we examine the options for the production of 236gNp, based on work carried out at NPL since 2011. However, this work was primarily aimed at the production of 236Pu, and not 236gNp and therefore the rate of production are based on the levels of 236Pu generated in the irradiation of (i) 238U with protons, (ii) 235U with deuterons, (iii) 236U with protons and (iv) 236U with deuterons. The derivation of a well-defined cross section is complicated by the relevant paucity of information on the variation of the 236mNp:236gNp production ratio with incident particle energy. Furthermore, information on the purity of 236gNp so produced is similarly sparse. Accordingly, the existing data is assessed and a plan for future work is presented. - Highlights: • We review the requirements for chemical yield tracers and the candidate tracers for 237Np analysis. • The current state of nuclear decay data and nuclear structure information for 236gNp and 236mNp are examined. • Existing cross section data for the production of 236Np by the irradiation of 235U, 236U and 238U with protons and deuterons are summarised. • We present the strategies employed at NPL for the recovery of 236Np from irradiated targets. • The results of these irradiations are presented

  3. Assessment and reduction of proliferation risk of reactor-grade plutonium regarding construction of ‘fizzle bombs’ by terrorists

    The approximately 23.7 wt% 240Pu in reactor-grade plutonium denatures the 239Pu to the extent that it cannot fuel high yield nuclear weapons. 240Pu has a high spontaneous fission rate, which increases the spontaneous neutron flux within the fuel. When such a nuclear weapon is triggered, these neutrons cause the nuclear fission chain reaction to pre-detonate which blows the imploding fuel shell apart before the designed level of compression and reactivity could be attained, thereby greatly reducing the average energy yield of such “fizzle” bombs. Therefore reactor-grade plutonium is normally viewed as highly proliferation resistant. In this article the literature on the proliferation resistance of reactor-grade plutonium and on the mechanism and effect of fizzle bombs is reviewed in order to test this view. It is shown that even very low yield fizzle bombs, exploded in urban areas, would still cause serious blast damage as well as radioactive contamination. Combined with the high levels of induced terror, fizzle bombs might thus be attractive psychological weapons for terrorists. Therefore reactor-grade plutonium may not be sufficiently proliferation resistant against nuclear terrorism. However, denaturisation with more than 9% 238Pu produces high levels of decay heat which will melt or explode the high explosives around uncooled implosion type weapons, rendering them useless. Unfortunately, reactor-grade Pu contains only 2.7% 238Pu and is thus not sufficiently proliferation resistant in this respect. It is also shown that the associated neptunium poses a substantial proliferation risk. In the present study strong improvement of the proliferation resistance was demonstrated by simulation of incineration of reactor-grade plutonium in the 400 MWth Pebble Bed Modular Reactor Demonstration Power Plant. Results for modified fuel cycles, aimed at transmutating 237Np to 238Pu are also reported. However, these modifications increased the disloaded heavy metal mass

  4. Dose rates as a function of time due to postulated radionuclide releases from the U.S. Yucca Mountain high-level radioactive waste repository

    The Yucca Mountain repository, which is located in a remote area in the State of Nevada, is being constructed for the long-term care and disposal of spent nuclear fuel and vitrified high-level radioactive waste. In accordance with U.S. law, the U.S. Environmental Protection Agency (USEPA) promulgated Standards that limit the dose rates to members of the public due to the consumption of ground water, alone, and the consumption of ground water plus agricultural products irrigated with the contaminated ground water, and other exposures, such as those from external sources and the inhalation of airborne radioactive materials. As part of this exercise, the USEPA identified eight specific radionuclides to which their Standards are to apply. These are: 14C, 99Tc, 129I, 226Ra, 228Ra, 237Np, 239Pu, and 241Am. For purposes of the associated dose rate estimates, a range of conservative assumptions have been applied, all of which are designed to assure that the estimated dose rates are well above what might be expected under 'real-world' conditions. As a first step, it was assumed that: (1) at 104 year after repository closure, a fractional release of 10-5 of the entire repository radionuclide inventory occurred; (2) the only prior reduction in the inventory was that due to radioactive decay; and (3) the sole path of exposure to neighboring population groups was through the consumption of 2 L d-1 of contaminated ground water. The accompanying analyses revealed that, of the eight radionuclides, only 226Ra, 237Np, and 239Pu, will represent a significant source of dose at that time. To provide perspective and insights, the next step was to estimate the committed effective dose rates for all eight radionuclides based on an assumed fractional release each year of 10-5 of the inventory from the time of repository closure up through the 106 year. For purposes of providing perspective, it was assumed that each dose rate estimate was independent, that is, no releases had occurred prior

  5. Assessment and reduction of proliferation risk of reactor-grade plutonium regarding construction of ‘fizzle bombs’ by terrorists

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School for Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001 (Internal Post Box 360), Potchefstroom 2520 (South Africa); Mulder, Eben J. [School for Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)

    2014-05-01

    The approximately 23.7 wt% {sup 240}Pu in reactor-grade plutonium denatures the {sup 239}Pu to the extent that it cannot fuel high yield nuclear weapons. {sup 240}Pu has a high spontaneous fission rate, which increases the spontaneous neutron flux within the fuel. When such a nuclear weapon is triggered, these neutrons cause the nuclear fission chain reaction to pre-detonate which blows the imploding fuel shell apart before the designed level of compression and reactivity could be attained, thereby greatly reducing the average energy yield of such “fizzle” bombs. Therefore reactor-grade plutonium is normally viewed as highly proliferation resistant. In this article the literature on the proliferation resistance of reactor-grade plutonium and on the mechanism and effect of fizzle bombs is reviewed in order to test this view. It is shown that even very low yield fizzle bombs, exploded in urban areas, would still cause serious blast damage as well as radioactive contamination. Combined with the high levels of induced terror, fizzle bombs might thus be attractive psychological weapons for terrorists. Therefore reactor-grade plutonium may not be sufficiently proliferation resistant against nuclear terrorism. However, denaturisation with more than 9% {sup 238}Pu produces high levels of decay heat which will melt or explode the high explosives around uncooled implosion type weapons, rendering them useless. Unfortunately, reactor-grade Pu contains only 2.7% {sup 238}Pu and is thus not sufficiently proliferation resistant in this respect. It is also shown that the associated neptunium poses a substantial proliferation risk. In the present study strong improvement of the proliferation resistance was demonstrated by simulation of incineration of reactor-grade plutonium in the 400 MW{sub th} Pebble Bed Modular Reactor Demonstration Power Plant. Results for modified fuel cycles, aimed at transmutating {sup 237}Np to {sup 238}Pu are also reported. However, these

  6. Uranium in the Nuclear Fuel Cycle: Creation of Plutonium (Invited)

    Ewing, R. C.

    2009-12-01

    One of the important properties of uranium is that it can be used to “breed” higher actinides, particularly plutonium. During the past sixty years, more than 1,800 metric tonnes of Pu, and substantial quantities of the “minor” actinides, such as Np, Am and Cm, have been generated in nuclear reactors - a permanent record of nuclear power. Some of these transuranium elements can be a source of energy in fission reactions (e.g., 239Pu), a source of fissile material for nuclear weapons (e.g., 239Pu and 237Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., 239Pu and 237Np). In fact, the new strategies of the Advance Fuel Cycle Initiative (AFCI) are, in part, motivated by an effort to mitigate some of the challenges of the disposal of these long-lived actinides. There are two basic strategies for the disposition of these heavy elements: 1.) to “burn” or transmute the actinides using nuclear reactors or accelerators; 2.) to “sequester” the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, such as zircon or isometric pyrochlore, A2B2O7 (A= rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage1. The radiation stability of these compositions is closely related to the structural distortions that can be accommodated for specific pyrochlore compositions and the electronic structure of the B-site cation. Recent developments in the understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms.

  7. The significance of agricultural vs. natural ecosystem pathways in temperate climates in assessments of long-term radiological impact

    Recent developments in performance assessment biosphere models have begun to emphasise the importance of natural accumulation pathways. In contrast to the agricultural pathways, the database for natural ecosystem pathways is less well developed, leading to a mismatch in quality of representations of the two types of system. At issue is the lack of reliable soil-plant and animal ingestion transfer factors for key radionuclides in natural ecosystems. The relative importance of the agricultural vs. natural ecosystem pathways is investigated here, in the context of a temperate site in present day, Eastern France. The BIOMASS Candidate Critical Group (CCG) methodology has been applied to map a set of eight candidate critical groups derived from the present-day societal context onto physical locations within a simple model of a river catchment system. The overall assessment model has been implemented using the Aquabios code. Annual individual dose to each of the CCGs has been calculated for each of the key radionuclides (79Se, 94Nb, 99Tc, 129I, 135Cs and 237Np) released to the valley aquifer and river. In addition to the traditional agricultural pathways, lifestyle groups exploiting natural habitats are explicitly addressed. Results show the susceptibility of different candidate critical groups to different radionuclides. A reference database typical of those employed in long-term performance assessment models is employed. Doses from external exposure (94Nb) and dust inhalation (237Np) are shown to dominate agricultural food consumption by factors of more than six, but, with the reference data set, foodstuffs obtained from natural ecosystems do not contribute significantly to critical group dose and, at most, show similar exposures to the agricultural pathways. This may lead to the conclusion that natural food can be ruled out of consideration in performance assessment models. However, systematic parametric sensitivity studies carried out on soil-plant and animal

  8. A comparison between the Example Reference Biosphere model ERB 2B and a process-based model: simulation of a natural release scenario

    The BIOMASS methodology was developed with the objective of constructing defensible assessment biospheres for assessing potential radiological impacts of radioactive waste repositories. To this end, a set of Example Reference Biospheres were developed to demonstrate the use of the methodology and to provide an international point of reference. In this paper, the performance of the Example Reference Biosphere model ERB 2B associated with the natural release scenario, discharge of contaminated groundwater to the surface environment, was evaluated by comparing its long-term projections of radionuclide dynamics and distribution in a soil-plant system to those of a process-based, transient advection-dispersion model (AD). The models were parametrised with data characteristic of a typical rainfed winter wheat crop grown on a sandy loam soil under temperate climate conditions. Three safety-relevant radionuclides, 99Tc, 129I and 237Np with different degree of sorption were selected for the study. Although the models were driven by the same hydraulic (soil moisture content and water fluxes) and radiological (Kds) input data, their projections were remarkably different. On one hand, both models were able to capture short and long-term variation in activity concentration in the subsoil compartment. On the other hand, the Reference Biosphere model did not project any radionuclide accumulation in the topsoil and crop compartments. This behaviour would underestimate the radiological exposure under natural release scenarios. The results highlight the potential role deep roots play in soil-to-plant transfer under a natural release scenario where radionuclides are released into the subsoil. When considering the relative activity and root depth profiles within the soil column, much of the radioactivity was taken up into the crop from the subsoil compartment. Further improvements were suggested to address the limitations of the Reference Biosphere model presented in this paper

  9. Mound Facility activities in chemical and physical research: July--December 1977

    Isotope separation of Ar, C, 3He, Kr, Ne, O, and Xe isotopes is reported. TiFeH/sub x/, TiCoH/sub x/, TiCuH/sub x/, and VH/sub x/ were studied using NMR (proton relaxation times). VD/sub x/ and VT/sub x/ were synthesized. The problem of calculating the valence state of Pu is discussed. A series solution to the plutonium (N,H) characteristic equation is suggested. Shipments of 231Pa, 230Th, and 229Th are reported. Separation and processing of 234U are also reported. Theoretical methods were developed to calculate temperature distributions as functions of water flow rate in liquid thermal diffusion columns. Diffusion coefficients were measured from 300 to 12000K for Kr-Xe and Kr-Ar. New thermal diffusion factors are submitted for Ne-Ar

  10. Pacific deep circulation: A velocity increase at the end of the interglacial stage 5?

    Mangini, A.; Dominik, J.; Müller, P. J.; Stoffers, P.

    1982-12-01

    Re-evaluation of 230Th and 231Pa data on 16 sediment cores recovered in the equatorial North Pacific, between the Clarion-Clipperton Fracture Zone and in the Aitutaki Passage, suggests that a major event modifying the sedimentary regime occurred about 70,000 y B.P. The change is recorded in 12 cores either as the onset of sediment accumulation following a period of sediment erosion or as a remarkable increase in the accumulation rate resulting from enhanced accumulation of redistributed sediment in abyssal plains. Both the onset of sediment accumulation and the enhanced accumulation of redistributed sediment could be attributed to bottom water velocities similar to present ones. Erosion, by contrast, is related to a period of maximum bottom water flow at the boundary of interglacial stage 5 and glacial stage 4.

  11. Developing 226Ra and 227Ac age-dating techniques for nuclear forensics to gain insight from concordant and non-concordant radiochronometers

    The model age or 'date of purification' of a nuclear material is an important nuclear forensic signature. In this study, chemical separation and MC-ICP-MS measurement techniques were developed for 226 Ra and 227Ac: grand-daughter nuclides in the 238U and 235U decay chains respectively. The 230Th-234U, 226Ra-238U, 231Pa-235U, and 227Ac-235U radiochronometers were used to calculate model ages for CRM-U100 standard reference material and two highly-enriched pieces of uranium metal from the International Technical Working Group Round Robin 3 Exercise. In conclusion, the results demonstrate the accuracy of the 226Ra-238U and 227Ac-235U chronometers and provide information about nuclide migration during uranium processing

  12. Developing 226Ra and 227Ac age-dating techniques for nuclear forensics to gain insight from concordant and non-concordant radiochronometers

    The model age or 'date of purification' of a nuclear material is an important nuclear forensic signature. In this study, chemical separation and MC-ICP-MS measurement techniques were developed for 226Ra and 227Ac: grand-daughter nuclides in the 238U and 235U decay chains, respectively. The 230Th-234U, 226Ra-238U, 231Pa-235U, and 227Ac-235U radiochronometers were used to calculate model ages for CRM-U100 standard reference material and two highly-enriched pieces of uranium metal from the International Technical Working Group Round Robin 3 Exercise. Results demonstrate the accuracy of the 226Ra-238U and 227Ac-235U chronometers and provide information about nuclide migration during uranium processing. (author)

  13. Trace determination of uranium and thorium in biological samples by radiochemical neutron activation analysis

    Radiochemical neutron activation analysis (RNAA) is an excellent method for determining uranium and thorium; it offers unique possibilities for their ultratrace analysis using selective radiochemical separations. Regarding the favourably sensitive nuclear characteristics of uranium and of thorium with respect to RNAA, but the different half-lives of their induced nuclides, two different approaches were used. In the first approach uranium and thorium were determined separately via 239U, 239Np and 233Pa. In the second approach these elements were 239239233 determined simultaneously in a single sample using U and/or Np and Pa. Isolation of induced nuclides was based on separation by extraction and/or anion exchange chromatography. Chemical yields were measured in each sample aliquot using added 235U, 238Np and 231Pa radioisotopic tracers. (author)

  14. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  15. Non-destructive γ spectrum analysis of polymetallic nodules from the eastern Pacific

    刘广山; 黄奕普; 蔡毅华; 陈敏

    2002-01-01

    -- Non-destructive γ spectrum analyses of 20 polymetallic nodules from the eastern Pacific were carried out. Numerous nuclides, such as 238 U, 230 Th, 226 Ra, 210 Pb, 228 Ra, 228 Th, 235 U, 227 Ac ( or 231pa) and 40K were detected. The count rates of the nuclides in the top or bottom side of nodules facing detector were measured and the ratio R of the count rates of nuclides in the top and the bottom sides was obtained. From counts and ratios, some useful information relating to the growth and movement of the nodules, the source of nuclide and relationship between those and environment can be gotten. A new method for clear distinction between the top and bottom sides of the nodule based on the R value of 226Ra or 210pb was developed. In addition, one can infer the turnover of nodules according to the R value of 230Th.

  16. Experiments to produce odd-mass neutron-deficient isotopes of superheavy elements by 48Ca ion-induced reactions

    Experimentals to produce neutron-deficient isotopes of superhevy elements in the reactions 233U (48Ca, 2n)279112, sup(231)Pa(48Ca, 3n)sup(276)111 and sup(232)Th(48Ca,2-3n)sup(277-276)110 have been carried out using a heavy ion beam of the U-300 accelerator. In these reactions, the upper limits of production cross sections and spontaneous-fission half-lives been determined to be 7x10-35 cm2, 5x10-35 cm2 and 2x10-35 cm2, and 0.05 s, 0.003 s and 0.003 s, respectively

  17. Neutron measurements for innovative fuel cycle and transmutation performed at the CEN Bordeaux-Gradignan : transfer techniques applied to the protactinium case

    Transfer reaction techniques have been used to determine neutron induced fission cross section (σn,f) of the short lived 233Pa nucleus, which is of importance for the Th-U fuel cycle for innovative reactors. The σn,f of 233Pa has been determined from the product of the fission probability of 234Pa measured in transfer reaction 232Th(3He,p) with the calculated compound nucleus formation cross section in the 233Pa+n reaction. The validity of this method has been tested with the existing data for direct neutron experiments on long-lived target nuclei 231Pa and 230Th. Transfer reaction techniques have been used too for the determination of capture cross section (σn,y) of 233Pa. This method will be extended to other highly radioactive actinides (such as 242-245Cm isotopes). (author)

  18. U-Th-Ra-Pa Disequilibria in the Kasuga Seamounts: recent "sediment" flux melting in the Mariana rear arc

    Gill, J.; Holden, P.

    2002-12-01

    Mariana volcanic front lavas define a U-Th isotope mixing line with an apparent age of 30 Ka between U-enriched "basalt fluid"-dominated Guguan and "sediment melt"-dominated Uracas in 238U-230Th equilibrium (Elliott et al., 1997). However, new results for basalts collected by dredging and diving on the shoshonitic Kasuga Seamounts, 10-20 km behind the VF, require re-interpretation of both Mariana components. Kasuga basalts are the local "sediment" extreme, reaching La/Sm = 5, Th/Nb=0.75, and eNd=3 in the most K-rich samples. Despite this extremity, their U-Th disequilibria lie along the same mixing line as for the VF, but extend to 20 percent 230Th-enrichment and (230Th)/(232Th) lower than at the intersection with the equiline. This indicates deeper melting than at the VF, and that the source's Th/U ratio was higher than the intersection. (226Ra)/(230Th) ratios extend to 3.5 even though samples have unknown eruption ages and Ba/Th is only 100, much lower than at the VF. (231Pa)/(235U) is mostly 1.7, higher than at the VF. (231Pa)/(230Th) correlates positively with excess U, consistent with recent flux melting. However, the mantle being melted is more fertile than at the VF, and the flux is more "sedimentary" apart from its disequilibria. Disequilibria in the highest-K Kasuga are most like Kick-em-Jenny, the most sediment-rich part of the Antilles.

  19. Fusion blanket benchmark experiments on a 60 cm-thick lithium-oxide cylindrical assembly

    Integral experiments on a Li2O cylindrical assembly have been carried out, using the FNS facility to provide benchmark data for verification of methods and data used in fusion neutronics research. The size of assembly was 63 cm (diameter) by 61 cm (length). Measurements included 6Li and 7Li tritium production rates ; 235U, 238U, 237Np and 232Th fission rates ; 27Al(n,α)24Na, 58Ni(n,2n)57Ni, 115In(n,n1)115mIn and 115In(n,γ)116In reaction rates. Neutron energy spectra in the assembly, as well as response rates of TLDs and PIN diodes, were also measured. Measured data are presented in tabular form together with estimated errors. A sample calculation using the DOT3.5 code is provided to facilitate the reader understanding of the experiments. Although several different measuring techniques are used in the experiment, the data are mutually consistent. This fact supports that present experimental data can be applied to the benchmark verification of methods and data. (author)

  20. Criticality safety aspects of spent fuel arrays from emerging nuclear fuel cycles

    Emerging nuclear fuel cycles: fuels with Pu or minor actinides (MA) for their self-generated recycling or transmutation in PWR or FR → reduction of radiotoxicity of HLW. The aim of work is to assess criticality (k∞) of arrays of spent nuclear fuels from these emerging fuel cycles. Procedures: Calculations of - k∞ using MCNP5 based on fresh and spent fuel compositions (infinite arrays), - spent fuel compositions using ORIGEN. Fuels considered: - commercial PWR-UO2 (R1) and -MOX (R2), [45 GWd/t] and fast reactor [100 GWd/t] (R3), - PWR self-generated Pu recycling (S1) and MA recycling (S2), FR self-generated MA recycling (S3), FR with 2% 237Np for transmutation purposes (T). Results: k∞ based on fresh and spent fuel compositions is shown. Fuels are clustered in two distinct families: - fast reactor fuels, - thermal reactor fuels; k∞ decreases when calculated on the basis of actinide and fission product inventory. In conclusions: - Emerging fuels considered resemble their corresponding commercial fuels; - k∞ decreases in all cases when calculated on the basis of spent fuel compositions (reactivity worth ∼-20%Δk/k), hence improving the effectiveness of packaging. (author)

  1. Solubility and speciation results from oversaturation experiments on neptunium, plutonium and americium in a neutral electrolyte with a total carbonate similar to water from Yucca Mountain Region Well UE- 25p No. 1

    Torretto, P.; Becraft, K.; Prussin, T.; Roberts, K.; Carpenter, S.; Hobart, D.; Nitsche, H. [Lawrence Berkeley Lab., CA (United States)

    1995-12-01

    Solubility and speciation are important in understanding aqueous radionuclide transport through the geosphere. They define the source term for transport retardation processes such as sorption and colloid formation. Solubility and speciation data are useful in verifying the validity of geochemical codes that are a part of predictive transport models. Solubility experiments will approach solution equilibrium from both oversaturation and undersaturation. In these experiments, we have approached the solubility equilibrium from oversaturation, Results are given for solubility and speciation experiments from oversaturation of {sup 237} NpO{sub 2}{sup +} {sup 239}Pu{sup 4+}, and {sup 241}Am{sup 3+}/Nd{sup 3+} in a neutral electrolyte containing a total carbonate concentration similar to groundwater from the Yucca Mountain region, Nevada, which is being investigated as a potential high-level nuclear waste disposal site, at 25{degrees}C and three pH values. In these experiments, the solubilitycontrolling steady-state solids were identified and the speciation and/or oxidation states present in the supernatant solutions were determined.

  2. Effects of ICRP Publication 30 and the 1980 Beir report on hazard assessments of high-level waste

    ICRP Publication 30 (ICRP30) gives the radiation dose to various body organs in Sv per Bq, which are readily translated into rad per curie ingested for various radioisotopes; the 1980 BEIR Report (BEIR III) gives the risk of fatal cancer per rad of dose to each body organ; the ORIGEN 2 Code gives the curies per tonne of initial uranium fuel in high level waste and in spent fuel. The product of these, Ci/tonne X rad/Ci X risk/rad gives the risk/tonne in terms of fatal cancer doses, if ingested, per tonne of uranium. When the resulting curves are compared with those calculated previously (based on ICRP Publication 2 and the 1972 BEIR Report), the hazard is less for the first century but much more thereafter. The principal sources of this greatly increased hazard is the 100-fold increase in absorption through the gut for 237Np in ICRP 30, plus the high cancer risk from radiation to liver in BEIR III. Some consequences of these changes and some scientific questions relevant to ascertaining their validity are discussed. (author)

  3. Distribution and behavior of artificial long-lived radionuclides in a sea sediment. Coastal sediment in the Irish Sea

    Over the last four decades, the Irish Sea has received controlled discharges of radioactive effluents from the Sellafield (Windscale) nuclear fuel reprocessing plant in Cumbria, UK. Enhanced levels of a range of fission, activation and transuranic elements have been detected in a variety of environmental media. Most of Pu and 241Am and about 10% of 137Cs discharged have been retained in a deposit of the fine sediment near the discharge point. The quantities of radionuclides discharged annually from Sellafield decreased by two orders of magnitude from the mid-1970s to 1980, but estimated internal and external exposure for critical group decreased by only less than one order of magnitude over this period. Redistribution of the highly contaminated marine sediment is potentially of major significance. In this paper, a review is presented of published work and our studies relating to the artificial long-lived radionuclides, 99Tc, 137Cs, 237Np, Pu isotopes and241Am, in sediments around the Irish Sea coast. The dominant mechanism (solution and/or particle transport) of supply of these radionuclides to coastal sediment is mainly discussed. (author)

  4. Enumeration of Microbial Populations in Radioactive Environments by Epifluorescence Microscopy

    Epifluorescence microscopy was utilized to enumerate halophilic bacterial populations in two studies involving inoculated, actual radioactive waste/brine mixtures and pure brine solutions. The studies include an initial set of experiments designed to elucidate potential transformations of actinide-containing wastes under salt-repository conditions, including microbially mediated changes. The first study included periodic enumeration of bacterial populations of a mixed inoculum initially added to a collection of test containers. The contents of the test containers are the different types of actual radioactive waste that could potentially be stored in nuclear waste repositories in a salt environment. The transuranic waste was generated from materials used in actinide laboratory research. The results show that cell numbers decreased with time. Sorption of the bacteria to solid surfaces in the test system is discussed as a possible mechanism for the decrease in cell numbers. The second study was designed to determine radiological and/or chemical effects of 239Pu, 243Am, 237Np, 232Th and 238U on the growth of pure and mixed anaerobic, denitrifying bacterial cultures in brine media. Pu, Am, and Np isotopes at concentrations of -5M, -6M and -4M respectively, and Th and U isotopes -3M were tested in these media. The results indicate that high actinide concentration affected both the bacterial growth rate and morphology. However, relatively minor effects from Am were observed at all tested concentrations with the pure culture

  5. Activation calculation and environmental safety analysis for fusion experimental breeder (FEB)

    An activation calculation code FDKR and decay chain data library AFDCDLIB are used to calculate the radioactivity, decay heat, dose rate and biological hazard potential (BHP) form activation products, actinides and fission products in a Fusion Experiment Breeder (FEB). The code and library are introduced briefly, and calculation results and decay curves of related hazards after one year operation with 150 MW fusion power are given. The total radioactivity inventory, decay heat and BHP are 5.74 x 1020 Bq, 8.34 MW and 4.08 x 108 km3 of air, respectively, at shutdown. Results obtained show that the first wall of FEB can meet the nuclear waste disposal criteria for the NRC 10 CFR61 Class C after a few weeks from shutdown. The inventory of important actinides for the fuel reprocessing, such as 232U and 237Np were also calculated. It was shown that their concentrations do not excess the limit value of environmental safety required. (9 refs., 4 figs., 9 tabs.)

  6. Final report on production of Pu-238 in commercial power reactors: target fabrication, postirradiation examination, and plutonium and neptunium recovery

    Considerable interest has been generated in more extensive applications of radioisotope thermoelectric generator (RTG) systems. This raises questions concerning the availability of 238Pu to supply an expanding demand. The development of much of this demand will depend upon a considerable reduction in cost of 238Pu. Two neptunia--zirconia--fuel target rods, containing four sections each of different NpO2 concentrations, were irradiated in the Connecticut Yankee Reactor for approximately one year. Following irradiation both target rods were subjected to nondestructive examination. One rod was chosen for destructive testing and analysis. Post-irradiation chemical analyses included total Pu and Np, ppM 236Pu/238Pu, and Pu isotopic abundance. The results of these analyses and of electron microprobe analysis which provided the relative Pu concentration across the pellet diameters are tabulated. It was concluded that the feasibility of all operations involved in the production of 238Pu by irradiation of 237NpO2 targets in commercial nuclear power reactors was demonstrated and that the demonstration should be extended to a pilot-scale leading to installation of a full production capacity. (U.S.)

  7. Review and Assessment of Neutron Cross Section and Nubar Covariances for Advanced Reactor Systems

    Maslov,V.M.; Oblozinsky, P.; Herman, M.

    2008-12-01

    In January 2007, the National Nuclear Data Center (NNDC) produced a set of preliminary neutron covariance data for the international project 'Nuclear Data Needs for Advanced Reactor Systems'. The project was sponsored by the OECD Nuclear Energy Agency (NEA), Paris, under the Subgroup 26 of the International Working Party on Evaluation Cooperation (WPEC). These preliminary covariances are described in two recent BNL reports. The NNDC used a simplified version of the method developed by BNL and LANL that combines the recent Atlas of Neutron Resonances, the nuclear reaction model code EMPIRE and the Bayesian code KALMAN with the experimental data used as guidance. There are numerous issues involved in these estimates of covariances and it was decided to perform an independent review and assessment of these results so that better covariances can be produced for the revised version in future. Reviewed and assessed are uncertainties for fission, capture, elastic scattering, inelastic scattering and (n,2n) cross sections as well as prompt nubars for 15 minor actinides ({sup 233,234,236}U, {sup 237}Np, {sup 238,240,241,242}Pu, {sup 241,242m,243}Am and {sup 242,243,244,245}Cm) and 4 major actinides ({sup 232}Th, {sup 235,238}U and {sup 239}Pu). We examined available evaluations, performed comparison with experimental data, taken into account uncertainties in model parameterization and made use state-of-the-art nuclear reaction theory to produce the uncertainty assessment.

  8. Neutron-emission measurements at a white neutron source

    Haight, Robert C [Los Alamos National Laboratory

    2010-01-01

    Data on the spectrum of neutrons emittcd from neutron-induced reactions are important in basic nuclear physics and in applications. Our program studies neutron emission from inelastic scattering as well as fission neutron spectra. A ''white'' neutron source (continuous in energy) allows measurements over a wide range of neutron energies all in one experiment. We use the tast neutron source at the Los Alamos Neutron Science Center for incident neutron energies from 0.5 MeV to 200 MeV These experiments are based on double time-of-flight techniques to determine the energies of the incident and emitted neutrons. For the fission neutron measurements, parallel-plate ionization or avalanche detectors identify fission in actinide samples and give the required fast timing pulse. For inelastic scattering, gamma-ray detectors provide the timing and energy spectroscopy. A large neutron-detector array detects the emitted neutrons. Time-of-flight techniques are used to measure the energies of both the incident and emitted neutrons. Design considerations for the array include neutron-gamma discrimination, neutron energy resolution, angular coverage, segmentation, detector efficiency calibration and data acquisition. We have made preliminary measurements of the fission neutron spectra from {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu. Neutron emission spectra from inelastic scattering on iron and nickel have also been investigated. The results obtained will be compared with evaluated data.

  9. Semianalytical solutions of radioactive or reactive tracer transport in layered fractured media

    In this paper, semianalytical solutions are developed for the problem of transport of radioactive or reactive tracers (solutes or colloids) through a layered system of heterogeneous fractured media with misaligned fractures. The tracer transport equations in the matrix account for (a) diffusion, (b) surface diffusion (for solutes only), (c) mass transfer between the mobile and immobile water fractions, (d) linear kinetic or equilibrium physical, chemical, or combined solute sorption or colloid filtration, and (e) radioactive decay or first order chemical reactions. Any number of radioactive decay daughter products (or products of a linear, first-order reaction chain) can be tracked. The tracer-transport equations in the fractures account for the same processes, in addition to advection and hydrodynamic dispersion. Additionally, the colloid transport equations account for straining and velocity adjustments related to the colloidal size. The solutions, which are analytical in the Laplace space, are numerically inverted to provide the solution in time and can accommodate any number of fractured and/or porous layers. The solutions are verified using analytical solutions for limiting cases of solute and colloid transport through fractured and porous media. The effect of important parameters on the transport of 3H, 237Np and 239Pu (and its daughters) is investigated in several test problems involving layered geological systems of varying complexity. 239Pu colloid transport problems in multilayered systems indicate significant colloid accumulations at straining interfaces but much faster transport of the colloid than the corresponding strongly sorbing solute species

  10. Semianalytical Solutions of Radioactive or Reactive Transport in Variably-Fractured Layered Media: 1. Solutes

    In this paper, semianalytical solutions are developed for the problem of transport of radioactive or reactive solute tracers through a layered system of heterogeneous fractured media with misaligned fractures. The tracer transport equations in the non-flowing matrix account for (a) diffusion, (b) surface diffusion, (c) mass transfer between the mobile and immobile water fractions, (d) linear kinetic or equilibrium physical, chemical, or combined solute sorption or colloid filtration, and (e) radioactive decay or first-order chemical reactions. The tracer-transport equations in the fractures account for the same processes, in addition to advection and hydrodynamic dispersion. Any number of radioactive decay daughter products (or products of a linear, first-order reaction chain) can be tracked. The solutions, which are analytical in the Laplace space, are numerically inverted to provide the solution in time and can accommodate any number of fractured and/or porous layers. The solutions are verified using analytical solutions for limiting cases of solute and colloid transport through fractured and porous media. The effect of important parameters on the transport of 3H, 237Np and 239Pu (and its daughters) is investigated in several test problems involving layered geological systems of varying complexity

  11. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  12. The use of synthetic Zn-/Ni-labeled montmorillonite colloids as a natural bentonite marker

    distribution of the colloids mobilized from the synthetic montmorillonite is comparable to the one obtained from the natural FEBEX bentonite. No colloidal attachment occurs over 3500 h (∼5 months) for both types of montmorillonite colloids on fracture filling material from Grimsel. Similar results were obtained for the FEBEX bentonite colloids. The 233U concentration is almost constant with only a slight decrease over 3500 h. This might be due to a slow sorption kinetic since a reduction of U(VI) to U(IV) is not likely on the basis of redox potentials (Eh) measured and thermodynamic calculations. Results obtained for 237Np differ from those of 233U. The 237Np concentration remaining after ultracentrifugation (UC) decreases clearly after ∼300 h for both Ni and Zn-montmorillonite. According to the experimental Eh-pH conditions, Np(V) is probably reduced to Np(IV) and might be present either as Np(IV) colloids or attached to the montmorillonite colloids, which explains the concentration decrease after UC. Unfortunately, based on the data available so far one cannot differentiate which of the two explanations holds. Nevertheless, the Np does not sorb onto the FFM as indicated by the constant concentration found in the non-UC samples, which is a clear difference compared to the FEBEX data presented in Huber et al. (2011). The 99Tc presents a concentration decrease as observed for the 237Np but delayed in time. The concentrations differ for UC and not-UC samples after 1000 h which could be explained as for the 237Np. Actually, a reduction of pertechnetate to the tetravalent 99Tc is feasible according to the experimental Eh-pH conditions. The results obtained after UC show clearly lower concentrations. This demonstrates a radionuclide colloid association in good agreement with results obtained on natural FEBEX clay colloids. Nevertheless, no clear radionuclide sorption reversibility is observed over 3500 h in contrast to the work on natural FEBEX colloids, except for Am sorbed

  13. UPTAKE OF RADIONUCLIDE METALS BY SPME FIBERS

    Duff, M; S Crump, S; Robert02 Ray, R; Keisha Martin, K; Donna Beals, D

    2006-08-28

    The Federal Bureau of Investigation (FBI) Laboratory currently does not have on site facilities for handling radioactive evidentiary materials and there are no established FBI methods or procedures for decontaminating high explosive (HE) and fire debris (FD) evidence while maintaining evidentiary value. One experimental method for the isolation of HE and FD residue involves using solid phase microextraction or SPME fibers to remove residue of interest. Due to their high affinity for organics, SPME fibers should have little affinity for most metals. However, no studies have measured the affinity of radionuclides for SPME fibers. The focus of this research was to examine the affinity of dissolved radionuclide ({sup 239/240}Pu, {sup 238}U, {sup 237}Np, {sup 85}Sr, {sup 133}Ba, {sup 137}Cs, {sup 60}Co and {sup 226}Ra) and stable radionuclide surrogate metals (Sr, Co, Ir, Re, Ni, Ba, Cs, Nb, Zr, Ru, and Nd) for SPME fibers at the exposure conditions that favor the uptake of HE and FD residues. Our results from radiochemical and mass spectrometric analyses indicate these metals have little measurable affinity for these SPME fibers during conditions that are conducive to HE and FD residue uptake with subsequent analysis by liquid or gas phase chromatography with mass spectrometric detection.

  14. Study of the uncertainty in estimation of the exposure of non-human biota to ionising radiation

    Uncertainty in estimations of the exposure of non-human biota to ionising radiation may arise from a number of sources including values of the model parameters, empirical data, measurement errors and biases in the sampling. The significance of the overall uncertainty of an exposure assessment will depend on how the estimated dose compares with reference doses used for risk characterisation. In this paper, we present the results of a study of the uncertainty in estimation of the exposure of non-human biota using some of the models and parameters recommended in the FASSET methodology. The study was carried out for semi-natural terrestrial, agricultural and marine ecosystems, and for four radionuclides (137Cs, 239Pu, 129I and 237Np). The parameters of the radionuclide transfer models showed the highest sensitivity and contributed the most to the uncertainty in the predictions of doses to biota. The most important ones were related to the bioavailability and mobility of radionuclides in the environment, for example soil-to-plant transfer factors, the bioaccumulation factors for marine biota and the gut uptake fraction for terrestrial mammals. In contrast, the dose conversion coefficients showed low sensitivity and contributed little to the overall uncertainty. Radiobiological effectiveness contributed to the overall uncertainty of the dose estimations for alpha emitters although to a lesser degree than a number of transfer model parameters

  15. Safety procedures for the electron spectroscopy of actinides at the ALS

    This is an addendum to the ALS Experimental Safety Form Renewal for the continuation of actinide microspot experiments on beamlines 7.0. There are several modifications to the previously approved. procedures. There is an increase in the amount of allowable material of the low activity isotopes 238U, 237Np, 242Pu, and 248Cm. There is also the addition of 99Tc and the activity isotopes 232Th and 243Am to the list of permissible sample materials. All of the materials are alpha-emitters with negligible gamma fields with the exception of 99Tc which is a beta-emitter. There is a series of new experiments that requires the use of a crystal cleaver in the preparation chamber of the ultraESCA end station. The beamline 7.0 ultraESCA endstation has been suitably modified to permit the safe cleave of YUPd alloy rectangular ingots. AR of the sample materials are solids. The exact nature and composition of the samples are delineated in the sample preparation section that follows. A corresponding Radiological Work Authorization (RWA) must be issued for this work at ALS since the material amounts exceed those in the Low Activity Source (LAS) guidelines in Table I and those in the Values for Exemption of Sealed Source Inventory in Table II. The preliminary date for the next run of these sample materials has been tentatively scheduled in early February 1996 and this will be with the uranium cleave alloys, not the transuranic materials

  16. International Atomic Energy Agency consultants' meeting on Analytical Quality Control Services, Vienna, 17-19 September 1990

    An International Atomic Energy Agency Consultants' Meeting on Analytical Quality Control Services was held at the Vienna International Center 17-19 September 1990. The Consultants generally conclude that the current methods of the preparation of the materials for intercomparison studies and facilities are adequate for the current Agency mission for trace element and radionuclide materials, but it is suggested that freeze-drying facilities for biological materials are needed. This is critically important for many of the collected materials uniquely important to the Agency mission. Also new equipment for automated sieving analysis be acquired for some applications. Homogeneity testing performed by AQCS includes the determination of several trace elements (radionuclides) of different concentrations of several sub-samples taken from one bottle and the results are compared with those obtained for sub-samples taken from various bottles chosen at random using one way analysis of variance. This procedure is found to be appropriate and could be used further. TC and CRP should be promoted in relation to the development of reliable analytical methods for the determination of so called ''difficult elements'' like: Mo, Al, I, F, Li, Co, Cd and Ni as well as 237Np, 226Ra, 228Ra. Intercomparison runs should stay open to the whole international community although it would be desirable to consult ''reference labs'' for the certification of selected (''difficult'') analytes. The presently employed programme for intercomparison studies for data evaluation are valuable and suitable. Refs, figs and tabs

  17. Behavior of actinides in the Integral Fast Reactor fuel cycle

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides (237Np, 240Pu, 241Am, and 243Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors' confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  18. A radiochemical analyses of metastudtite and leachates from spent fuel

    Immersion of commercial spent nuclear fuel (CSNF) in deionized water produced two novel corrosion products after a two-year contact period. Another unexpected result was that suspensions of aggregates were observed to form at the air-water interface for each of five samples. These solids were characterized, by SEM and XRD to be nearly pure metastudtite (UO4-2H2O); while the corrosion present on the surface of the fuel itself was determined to be studtite (UO4-2H2O). The occurrence of the floating phase prompted a radiochemical analysis of these solids. This chemical analysis was a unique opportunity to study the relatively pure corrosion phase for incorporation of radionuclides. The analysis indicated that high concentration of 90Sr, 137Cs, 99Tc, and that lower concentrations 237Np, 238, 239Pu and 243, 244Cm had partitioned with the air-water interface aggregates. The concentrations of 241Am were two orders of magnitude lower than the expected inventory in the suspended solids. The radiochemical analyses of the several leachate samples provide preliminary solubility data for the hydrogen peroxide leaching of CSNF and these data are compared to leaching of the same fuel in J-13 and deionized waters. The extent of fuel dissolution in these media are discussed

  19. Analysis of high burnup spent nuclear fuel by ICP-MS

    Inductively coupled plasma mass spectrometry (ICP-MS) as the primary tool for determining concentrations of a suite of nuclides in samples excised from high-burnup spent nuclear fuel rods taken from light water nuclear reactors. The complete analysis included the determination of 95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 137Cs, 143Nd, 145Nd, 148Nd, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 151Eu, 153Eu, 155Eu, 155Gd, 237Np, 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 242mAm, and 243Am. The isotopic composition of fissiogenic lanthanide elements was determined using high-performance liquid chromatography (HPLC) with ICP-MS detection. These analytical results allow the determination of fuel burn-up based on 148Nd, Pu, and U content, as well as provide input for storage and disposal criticality calculations. Results show that ICP-MS along with HPLC-ICP-MS are suitable of performing routine determinations of most of these nuclides, with an uncertainty of ±10% at the 95% confidence level. (author)

  20. Neutron-induced Fission Cross Section of 240,242Pu

    Salvador-Castiñeira, P.; Bryś, T.; Eykens, R.; Hambsch, F.-J.; Göök, A.; Oberstedt, S.; Pretel, C.; Sibbens, G.; Vanleeuw, D.; Vidali, M.

    A sensitivity analysis for the new generation of fast reactors [Salvatores (2008)] has shown the importance of improved cross section data for several actinides. Among them, the 240,242Pu(n,f) cross sections require an accuracy improvement to 1-3% and 3-5%, respectively, from the current level of 6% and 20%. At the Van de Graaff facility of the Institute for Reference Materials and Measurements (JRC-IRMM) the fission cross section of the two isotopes was measured relative to two secondary standard reactions, 237Np(n,f) and 238U(n,f), using a twin Frisch-grid ionization chamber. The secondary standard reactions were benchmarked through measurements against the primary standard reaction 235U(n,f) in the same geometry. Sample masses were determined by means of low-geometry alpha counting or/and a 2π Frisch-grid ionization chamber, with an uncertainty lower than 2%. The neutron flux and the impact of scattering from material between source and target was examined, the largest effect having been found in cross section ratio measurements between a fissile and a fertile isotope. Our 240,242Pu(n,f) cross sections are in agreement with previous experimental results and slightly lower than present evaluations. In case of the 242Pu(n,f) reaction no evidence for a resonance at En=1.1 MeV was found.

  1. Fragment-mass distributions in fission of heavy nuclei by intermediate and high-energy probes

    Deppman, Airton; Andrade-II, E. [Universidade de Sao Paulo (IF/USP), SP (Brazil). Inst. de Fisica; Menezes, J.C.M.; Garcia, F. [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Duarte, S.B.; Tavares, O.A.P. [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Rossi, P.C.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Full text: Recent experiments have shown that the multimode approach for describing the fission process leads to some compatibility with the observed results. A systematic analysis of the parameters obtained by fitting the fission-fragment mass distribution to the spontaneous and low-energy data has shown that the values for those parameters present a smooth dependence upon the nuclear mass number. In the present work it is shown that the same parameter-values obtained for low- energy fission can be used to describe high-energy fission results of fragment-mass distributions if one takes into account the appropriate distribution of the fissioning system. To calculate the fission-fragment mass distributions, Monte Carlo simulations are used. This simulation considers a two-step reaction mechanism, namely, an intranuclear cascade providing the compound nucleus followed by a mechanism of competition between particle evaporation and fission. The fission-fragment masses are obtained according to the multimode approach following the Statistical Scission Model. Simulations for fission induced by 660 MeV protons on 241Am and 237Np, and for fission of 238U induced by photons from Bremsstrahlung with end-point energies of 50 MeV and 3500 MeV have been performed, and the results have been compared with recent experimental data. (author)

  2. Contaminated nickel scrap processing

    The DOE will soon choose between treating contaminated nickel scrap as a legacy waste and developing high-volume nickel decontamination processes. In addition to reducing the volume of legacy wastes, a decontamination process could make 200,000 tons of this strategic metal available for domestic use. Contaminants in DOE nickel scrap include 234Th, 234Pa, 137Cs, 239Pu (trace), 60Co, U, 99Tc, and 237Np (trace). This report reviews several industrial-scale processes -- electrorefining, electrowinning, vapormetallurgy, and leaching -- used for the purification of nickel. Conventional nickel electrolysis processes are particularly attractive because they use side-stream purification of process solutions to improve the purity of nickel metal. Additionally, nickel purification by electrolysis is effective in a variety of electrolyte systems, including sulfate, chloride, and nitrate. Conventional electrorefining processes typically use a mixed electrolyte which includes sulfate, chloride, and borate. The use of an electrorefining or electrowinning system for scrap nickel recovery could be combined effectively with a variety of processes, including cementation, solvent extraction, ion exchange, complex-formation, and surface sorption, developed for uranium and transuranic purification. Selected processes were reviewed and evaluated for use in nickel side-stream purification. 80 refs

  3. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science

    2014-12-15

    This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  4. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    This paper aims to investigate 232Th/233U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. 232Th/235U/238U oxide mixture was considered as fuel in the core, when the mass fraction of 232Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of 238U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the 232Th, 233U, 238U, 237Np, 239Pu, 241Am and 244Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  5. Group neutron fission and radiative-capture cross-sections for transactinides

    A comparison is made between evaluations of radiative-capture and fission cross-sections for the isotopes 236U, 237Np, 238Pu, 241Am, 243Am, 242Cm and 244Cm, and group cross-sections for use in fast-reactor calculations are recommended. Group cross-sections obtained from the HEDL graphical data (evaluation for ENDF/B-V) are shown for 234U, 236Pu, 237Pu, 242Pu, 244Pu, sup(242m)Am, 241Cm, 243Cm and 248Cm. Group cross-sections for 32 isotopes from the ENDL-76 library files are also given. In choosing recommended cross-sections, account was taken of the extent of agreement with experimental data where these are available, the extent to which the cross-sections are documented and the extent to which they have been calculated from a theoretical model. The reliability of evaluations is discussed. An attempt is made to evaluate the error in single-group cross-sections averaged over a typical fast-reactor spectrum. Conclusions are drawn from a study of the literature on the current status of experimental and theoretical research on transactinide cross-sections, and from the spread of the different evaluation data. Finally, the situation with respect to the integral experiments which can be used for correcting transactinide cross-sections is discussed. (author)

  6. ANSI/ANS-8.15-1981(R87): Nuclear criticality control of special actinide elements

    The American National Standard, open-quotes Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactotorsclose quotes American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1-1983(R88) provides guidance for the nuclides 233U, 235U, and 239Pu. These three nuclides are of primary interest in out-of-reactor criticality safety since they are the most commonly encountered in the vast majority of operations. However, some operations can involve nuclides other than 233U, 235U, and 239Pu in sufficient quantities that their effect on criticality safety could be of concern. ANSI/ANS-8.15-1981(R87) open-quotes Nuclear Criticality Control of Special Actinide Elements,close quotes provides guidance for 15 such nuclides. The standard was approved for use on November 9, 1981. When it received its first 5-yr review, no changes were made, and it was reaffirmed effective October 30, 1987. The standard was again reviewed and reaffirmed without changes in December 1995. The next 5-yr review of the standard is due in December 2000. The affected nuclides are 237Np, 238Pu, 240Pu, 242Pu, 241Am, 243Am, 244Cm, 239Pu, 241Pu, 242mAm, 243Cm, 245Cm, 247Cm, 249Cf, and 251Cf

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

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

    2015-07-23

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

  8. Management of disused long lived sealed radioactive sources (LLSRS)

    The document provides advice the sealed source users and the national waste management organizations with the technical know-how on the management of disused and spent long lived sealed radioactive sources (LLSRS) and with the particular guidelines required for handling, conditioning for storage, and storage of these sources. The guidance is intended to assist in establishing compliance with the present standards, requirements, and adopted practices. It also provides background material for any possible technical assistance to developing countries and serves as a reference for technical staff involved with IAEA programmes on the subject. Because of the historic nature of many of the sources under this category and the lack of well developed technical procedures recognized on the international level, this publication can serve as a basis for establishing future handling and conditioning procedures. The LLSRS addressed in this publication are primarily those containing radionuclides having half-lives greater than 30 years. These sources may contain long lived alpha-emitters, mainly 238Pu, 239Pu, 237Np, 241Am, 226Ra; beta-emitters: 14C, and 63Ni and could be neutron sources such as PuBe, RaBe and AmBe

  9. Proceedings of public hearings: plutonium and the other transuranium elements, Washington, D.C., December 10--11, 1974. Volume 1

    The Environmental Protection Agency embarked on a program to evaluate the environmental impact of the transuranium elements and to consider whether further guidelines or standards are needed to assure adequate protection of the general ambient environment and of the public health from potential contamination of the environment by radionuclides of these elements. Public hearings were held in Washington, D. C., and Denver, Colorado, to gather information regarding the public and social implications of plutonium utilization; the factors involved in the balancing of costs vs benefits; dosimetry, health, and environmental effects; environmental levels and pathways; applications using plutonium; and control and cleanup technology. The proceedings of the hearing in Washington, D. C., Dec. 10-11, 1974, are presented. Data are included on current and potential sources of transuranium elements in the environment; animal studies on the tissue distribution of 233U, 237Np, 238Pu, 239Pu, 241Am, 244Cm, 249Bk, 252Cf, and 253Es and pathological effects of body burdens of these radionuclides; and data on the health status of personnel known to have body burdens of 238Pu or 239Pu acquired during acute or chronic exposure, many of them over 30 years previously. It is pointed out that the lack of demonstrable biological effects of Pu in man provides presumptive evidence that the radiation protection standards in effect are adequate. (U.S.)

  10. Investigation of the mobilization and sorption characteristics of selected radionuclides at natural and technical barriers under MAW final storage conditions

    The final report summarizes the results of various research areas concerned with possible disposal-site accidents involving the destruction of containers containing radioactive waste by salt brine. The physico-chemical data obtained are intended to serve model calculations for probable risks. Prior attention is directed to the behaviour of 237Np, 129I, 135Cs and 79Se as long-lived nuclides, examined with 233U, 131I, 137Cs and 75Se as tracer nuclides. An initial research concern is with the characterisation of proximity zones during MAW emplacement in chambers and holes (pH, Eh, concentrations, carbonate concentrations, carrier colloids) in appropriate solution systems. The second research concern consists in the examination of the solubility and sorption behaviour of the elements involved, in the form of their compounds in the relevant systems. The third concern is with the sorption of the above mentioned elements in corrosion products of cement and of the materials of the tanks, as well as in salt grit and waste material. (RB)

  11. Use of boron filters for determining reactor spectra by means of threshold detectors

    The possibility of spectrum estimation and dose determination by means of a threshold detector system with boron and cadmium filters has been studied for reactor spectra. Measured spectra are presented in the form of step curves recorded by detectors consisting of 235U, 238U in Cd or of 235U, 237Np, 238U, 32S in a boron sphere (1.2 gcm-210B). Cross sections and effective energy thresholds of the detectors behind the boron filter were taken from the papers of Keirim-Markus et al. The detectors chosen allow to determine the neutron flux density with sufficient accuracy in the following energy ranges: 0.1 ... 0.4 eV, 0.4 ... 400 eV, 400 eV ... 0.56 MeV, 0.56 ... 1.4 MeV, 1.4 ... 2.8 MeV, 2.8 ... 10 MeV. The results of the experiments performed at the Dubna IBR-30 reactor are compared with spectra obtained from the activation method and the SAND code. (author)

  12. Systematic study of actinide and pre-actinide fission modes

    Andrade-II, E; Deppman, A; Bernal-Castillo, J L; Balabekyan, A R; Demekhina, N A; Adam, J; Garcia, F; Guzmán, F

    2016-01-01

    In this work, we present new experimental data on mass distribution of fission fragments from $^{241}$Am proton-induced fission at $660$ MeV measured at the LNR Phasotron (JINR). The systematic analysis of several measured fragment mass distributions from different fission reactions available in the literature is also presented. The proton-induced fission of $^{241}$Am, $^{237}$Np and $^{238}$U at 26.5, 62.9 and 660 MeV was studied. The proton-induced fission of $^{232}$Th was studied at 26.5, 62.9 and 190 MeV. The fission of $^{208}$Pb also by a proton was investigated at 190, 500 and 1000 MeV. The fission of $^{197}$Au was studied for 190 and 800 MeV protons. Bremsstrahlung reactions with maximum photon energies of 50 and 3500 MeV were studied for $^{232}$Th and $^{238}$U. The framework of the Random Neck Rupture Model was applied in the analysis. The roles of the neutron excess and of the so called fissility parameter were also investigated.

  13. A concept for quantitative NDA measurements of advanced reprocessing product materials

    As new reprocessing methods for spent nuclear fuel are developed, such as the uranium extraction (UREX) process, methods using nondestructive assay (NDA) techniques must also be developed to allow for quantitative measurements of product materials. Currently developed NDA techniques cannot directly quantify materials containing U, Np, Pu, and Am. This research investigates the ability to quantify these actinides in an oxide form using neutron multiplicity measurements. This technique assumes that the isotopic composition of the sample is known, either through gamma spectroscopy or other means. This measurement technique is based on performing three different neutron measurements and analyzing their neutron multiplicity response. The first is a passive measurement of the product material to determine the effective plutonium-240 (240Pueff) content, self multiplication (M), and alpha-neutron reaction rate (α). The second is an active, AmLi (α, n) source, measurement of the product material to determine the effective 235U content. The third is an active, AmB (α, n) source, measurement of the product material to determine the effective 237Np content. The quantity of Am in the sample can be determined from α. Simulated results using Monte Carlo N-Particle eXtended (MCNPX) version 2.6 will illustrate the viability of this technique and its practical limitations. (author)

  14. Progress report. P9

    China Evaluated Nuclear Data Library, version 3 (CENDL-3) was accomplished in 2000. CENDL-3 contains about 206 nuclides. Among them, the data of 161 nuclei will be newly or reevaluated: Fissile nuclei 15 ( 233-239 U, 237 Np, 238-242 Pu, 241Am , 242 Am); Structure material nuclei 34 (Natural elements Ni, Cu, Zr, Hf, Pb and their isotopes, 23 Na, Nat Si); Light nuclei 3 ( 6Li, 7 Li, 9Be); Fission product nuclei 109. The benchmark testing for CENDL-3 is being carried out. Several problems in physics and format have been found on major fissile nuclides, and are being improved. CINDA activities comprised compilation od 118 entries from the works in communication of Nuclear Data Progress in 1999-2000. Two young staff members of CNDC have started the EXFOR compilation, and fished 16 entries measured in China. 'Communication of Nuclear Data Progress' (CNDP) has been published for 24 issues by CNDC and Atomic Energy Press science 1989, and it has been distributed by the IAEA Nuclear Data Section as an NDC document

  15. Presentations and documents submitted to the 27. meeting of the Working Party on International Nuclear Data Evaluation Co-operation (WPEC), NEA Headquarters, 21-22 May 2015

    The NEA's nuclear data evaluation co-operation activities involve the following evaluation projects: ENDF (United States), JENDL (Japan), ROSFOND/BROND (Russia), JEFF (other Data Bank member countries) and CENDL (China) in close co-operation with the Nuclear Data Section of the International Atomic Energy Agency (IAEA). The working party was established to promote the exchange of information on nuclear data evaluations, measurements, nuclear model calculations, validation, and related topics, and to provide a framework for co-operative activities between the participating projects. The working party assesses nuclear data improvement needs and addresses these needs by initiating joint evaluation and/or measurement efforts. This document brings together the available documents and presentations relative to this meeting: the agenda, the Summary record of the previous meeting held on May 2014, the Reports (slides) on experimental activities from Europe, Japan, USA, Russia and China, the Brief progress reports from the evaluation projects (ENDF, JEFF, JENDL, ROSFOND, CENDL, IAEA, TENDL), the presentation from Subgroup 39 (Methods and approaches to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data files) and from Subgroup 41 (Improving nuclear data accuracy of 241Am and 237Np capture cross-sections). The document ends with a preliminary proposal for a New Subgroup 42 (Thermal Scattering Kernel S(α,β): Measurement, Evaluation and Application)

  16. Maximization of burning and/or transmutation (B/T) capacity in coupled spectrum reactor (CSR) by fuel and core adjustment

    A conceptual design of burning and/or transmutation (B/T) reactor, based on a modified conventional 1150 MWe-PWR system, consisted of two core regions for thermal and fast neutrons, respectively, was proposed herein for the treatments of minor actinides (MA). In the outer region 237Np, 241Am, and 243Am burned by thermal neutrons, while in the inner region 244Cm was burned mainly by fast neutrons. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio. The maximization of B/T capacity in CSR were done by, first, increasing the radius of the inner region. Second, reducing the coolant to fuel volume ratio, and third, choosing a suitable B/T fuel type. The result of the calculations showed that the equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute up to 808 kg of MA in a single reactor core effectively and safely. (author)

  17. Synthesis of Superheavy Nuclei in 48CA-INDUCED Reactions

    Oganessian, Yu. Ts.; Utyonkov, V. K.; Lobanov, Yu. V.; Abdullin, F. Sh.; Polyakov, A. N.; Sagaidak, R. N.; Shirokovsky, I. V.; Tsyganov, Yu. S.; Voinov, A. A.; Gulbekian, G. G.; Bogomolov, S. L.; Gikal, B. N.; Mezentsev, A. N.; Iliev, S.; Subbotin, V. G.; Sukhov, A. M.; Subotic, K.; Zagrebaev, V. I.; Vostokin, G. K.; Itkis, M. G.; Moody, K. J.; Patin, J. B.; Shaughnessy, D. A.; Stoyer, M. A.; Stoyer, N. J.; Wilk, P. A.; Kenneally, J. M.; Landrum, J. H.; Wild, J. F.; Lougheed, R. W.

    2008-11-01

    Thirty-four new nuclides with Z = 104-116, 118 and N = 161-177 have been synthesized in the complete-fusion reactions of 238U, 237Np, 242,244Pu, 243Am, 245,248Cm, and 249Cf targets with 48Ca beams. The masses of evaporation residues were identified through measurements of the excitation functions of the xn-evaporation channels and from cross bombardments. The decay properties of the new nuclei agree with those of previously known heavy nuclei and with predictions from different theoretical models. A discussion of self-consistent interpretations of all observed decay chains originating from the parent isotopes 282,283112, 282113, 286-289114, 287,288115, 290-293116, and 294118 is presented. Decay energies and lifetimes of the neutron-rich superheavy nuclei as well as their production cross sections indicate a considerable increase in the stability of nuclei with an increasing number of neutrons, which agrees with the predictions of theoretical models concerning the decisive dependence of the structure and radioactive properties of superheavy elements on their proximity to the nuclear shells with N = 184 and Z = 114.

  18. Synthesis, Decay Properties, and Identification of Superheavy Nuclei Produced in 48CA-INDUCED Reactions

    Oganessian, Yu. Ts.; Utyonkov, V. K.; Lobanov, Yu. V.; Abdullin, F. Sh.; Polyakov, A. N.; Sagaidak, R. N.; Shirokovsky, I. V.; Tsyganov, Yu. S.; Voinov, A. A.; Iliev, S.; Subbotin, V. G.; Sukhov, A. M.; Gulbekian, G. G.; Bogomolov, S. L.; Gikal, B. N.; Mezentsev, A. N.; Subotic, K.; Zagrebaev, V. I.; Itkis, M. G.; Moody, K. J.; Henderson, R. A.; Patin, J. B.; Shaughnessy, D. A.; Stoyer, M. A.; Stoyer, N. J.; Wilk, P. A.; Kenneally, J. M.; Landrum, J. H.; Wild, J. F.; Lougheed, R. W.

    2008-04-01

    Thirty-four new nuclides with Z = 104-116, 118 and N = 161-177 have been synthesized in the complete-fusion reactions of 238U, 237Np, 242,244Pu, 243Am, 245,248Cm, and 249Cf targets with 48Ca beams. The masses of evaporation residues were identified through measurements of the excitation functions of the xn-evaporation channels and from cross bombardments. The decay properties of the new nuclei agree with those of previously known heavy nuclei and with predictions from different theoretical models. A discussion of self-consistent interpretations of all observed decay chains originating from the parent isotopes 282,283112, 282113, 286-289114, 287,288115, 290-293116, and 294118 is presented. Decay energies and lifetimes of the neutron-rich superheavy nuclei as well as their production cross sections indicate a considerable increase in the stability of nuclei with the approach to the theoretically predicted nuclear shells with N = 184 and Z = 114.

  19. Nuclear Reactions Used For Superheavy Element Research

    Stoyer, Mark A.

    2008-04-01

    Some of the most fascinating questions about the limits of nuclear stability are confronted in the heaviest nuclei. How many more new elements can be synthesized? What are the nuclear and chemical properties of these exotic nuclei? Does the "Island of Stability" exist and can we ever explore the isotopes inhabiting that nuclear region? This paper will focus on the current experimental research on the synthesis and characterization of superheavy nuclei with Z>112 from the Dubna/Livermore collaboration. Reactions using 48Ca projectiles from the U400 cyclotron and actinide targets (233,238U, 237Np, 242,244Pu, 243Am, 245,248Cm, 249Cf) have been investigated using the Dubna Gas Filled Recoil Separator in Dubna over the last 8 years. In addition, several experiments have been performed to investigate the chemical properties of some of the observed longer-lived isotopes produced in these reactions. Some comments will be made on nuclear reactions used for the production of the heaviest elements. A summary of the current status of the upper end of the chart of nuclides will be presented.

  20. Photofission Analysis for Fissile Dosimeters Dedicated to Reactor Pressure Vessel Surveillance

    Bourganel Stéphane

    2016-01-01

    Full Text Available Fissile dosimeters are commonly used in reactor pressure vessel surveillance programs. In this paper, the photofission contribution is analyzed for in-vessel 237Np and 238U fissile dosimeters in French PWR. The aim is to reassess this contribution using recent tools (the TRIPOLI-4 Monte Carlo code and latest nuclear data (JEFF3.1.1 and ENDF/B-VII nuclear libraries. To be as exhaustive as possible, this study is carried out for different configurations of fissile dosimeters, irradiated inside different kinds of PWR: 900 MWe, 1300 MWe, and 1450 MWe. Calculation of photofission rate in dosimeters does not present a major problem using the TRIPOLI-4® Monte Carlo code and the coupled neutron-photon simulation mode. However, preliminary studies were necessary to identify the origin of photons responsible of photofissions in dosimeters in relation to the photofission threshold reaction (around 5 MeV. It appears that the main contribution of high enough energy photons generating photofissions is the neutron inelastic scattering in stainless steel reactor structures. By contrast, 137Cs activity calculation is not an easy task since photofission yield data are known with high uncertainty.

  1. Measurement of cross-sections of fission reactions induced by neutrons on actinides from the thorium cycle at n-TOF facility

    In the frame of innovating energy source system studies, thorium fuel cycle reactors are considered. Neutron induced fission cross section on such cycle involved actinides play a role in scenario studies. To feed them, data bases are built with experimental results and nuclear models. For some nuclei, they are not complete or in disagreement. In order to complete these data bases, we have built an original set up, consisting in an alternation of PPACs (Parallel Plate Avalanche Chamber) and ultra - thin targets, which we installed on n-TOF facility. We describe detectors, set up, and the particular care brought to target making and characterization. Fission products in coincidence are detected with precise time measurement and localization with delay line read out method. We contributed, within the n-TOF collaboration, to the CERN brand new intense spallation neutron source characterization, based on time of flight measurement, and we describe its characteristics and performances. We were able to measure such actinide fission cross sections as 232Th, 234U, 233U, 237Np, 209Bi, and natPb relative to 235U et 238U standards, using an innovative acquisition system. We took advantage of the lame accessible energy field, from 0.7 eV to 1 GeV, combined with the excellent energy resolution in this field. Data treatment and analysis advancement are described to enlighten performance and limits of the obtained results. (author)

  2. Rapid determination of actinides and {sup 90}Sr in river water

    Habibi, A., E-mail: azza.habibi@irsn.fr [IRSN/PRP-ENV/STEME/LMRE, Rue du belvédère, Bâtiment 501, Bois des rames, 91400 Orsay (France); Boulet, B. [IRSN/PRP-ENV/STEME/LMRE, Rue du belvédère, Bâtiment 501, Bois des rames, 91400 Orsay (France); Gleizes, M. [IRSN/PRP-ENV/STEME, 31 rue de l' écluse, 78116 Le Vésinet (France); Larivière, D. [Laboratoire de radioécologie, Département de chimie, Université Laval, 1045 Avenue de la médecine, G1V 0A6 Québec (Canada); Cote, G. [PSL Research University, Chimie ParisTech CNRS, Institut de Recherche de Chimie Paris, 11 rue Pierre et Marie Curie, 75005 Paris (France)

    2015-07-09

    Highlights: • A new method to separate six actinides and {sup 90}Sr was developed. • The method was applied successfully to river water samples. • The separation and the measure take about seven hours. • The method permits to reach high yields. - Abstract: Nuclear accidents occurred in latest years highlighted the difficulty to achieve, in a short time, the quantification of alpha and beta emitters. Indeed, most of the existing methods, though displaying excellent performances, can be very long, taking up to several weeks for some radioisotopes, such as {sup 90}Sr. This study focuses on alpha and beta radioisotopes which could be accidentally released from nuclear installations and which could be measured by inductively coupled plasma mass spectrometer (ICP-MS). Indeed, a new and rapid separation method was developed for {sup 234,235,236,238}U, {sup 230,232}Th, {sup 239,240}Pu, {sup 237}Np, {sup 241}Am and {sup 90}Sr. The main objective was to minimize the duration of the separation protocol by the development of a unique radiochemical procedure with elution media compatible with ICP-MS measurements. Excellent performances were obtained with spiked river water samples. These performances are characterized by total yields exceeding 80% for all monitored radionuclides, as well as good reproducibility (RSD ≤ 10%, n = 12). The proposed radiochemical separation (including counting time) required less than 7 h for a batch of 8 samples.

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

    Pham, M.K. [International Atomic Energy Agency (IAEA), Marine Environment Laboratory (MEL), MC 98000 (Monaco)]. E-mail: m.pham@iaea.org; Sanchez-Cabeza, J.A.; Povinec, P.P.; Gastaud, J.; La Rosa, J.; Lee, S.-H.; Liong Wee Kwong, L.; Oregioni, B.; Wyse, E. [International Atomic Energy Agency (IAEA), Marine Environment Laboratory (MEL), MC 98000 (Monaco); Arnold, D. [Physikalisch-Technische Bundesanstalt, Braunschweig, 38116 Germany (Germany); Benmansour, M. [Centre National de l' Energie, des Sciences et des Techniques Nucleaires (CNESTEN), B.P. 1382, R.P.10001, Rabat (Morocco); Bojanowski, R. [Institute of Oceanology, Polish Academy of Sciences, PL-81-712 Sopot (Poland); Carvalho, F.P. [Instituto Tecnologico e Nuclear, Departamento de Proteccao Radiologica e Seguranca Nuclear, P-2685-953 Sacavem (Portugal); Kim, C.K. [Department of Radiological Environmental Assessment, Korea Institute of Nuclear Safety, Yo-song, Taejon 305-600, Korea (Korea); Esposito, M. [Laboratorio di Ingegneria Nucleare, Universita di Bologna, 40136 Bologna (Italy); Gasco, C.L. [CIEMAT-DIAE, Radioecologia del Medio Acuatico, 28040 Madrid (Spain); Ham, G.J. [National Radiological Protection Board, Chilton, Didcot, Oxon, OX11 0RA (United Kingdom); Hegde, A.G. [Environmental Survey Laboratory, Bhabha Atomic Research Center, Tarapur Atomic Power Station, Maharashtra 401 504 (India); Holm, E. [Department of Medical Radiation Physics, Lund University Hospital, 22185 Lund (Sweden); Jaskierowicz, D. [Lab. d' Analyses de Surveillance et d' Expertise de la Marine, Base Navale de Cherbourg, 50115 Cherbourg (France); Kanisch, G. [Federal Research Centre for Fisheries, Institute of Fisheries Ecology, 20539 Hamburg (Germany); Llaurado, M. [Lab. de Radiologia Ambiental, Dept. de Quimica Analitica, Facultat de Quimica, Universitat de Barcelona, 08028 Barcelona (Spain); Le Petit, G. [Commissariat a l' Energie Atomique, DASE/SRCE, 91680 Bruyeres-le-Chatel (France); Maruo, Y. [and others

    2006-10-15

    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 ({sup 4}K, {sup 137}Cs, {sup 232}Th, {sup 234}U, {sup 235}U, {sup 238}U, {sup 238}Pu, {sup 239+24}Pu and {sup 241}Am) were certified for this material. Information on massic activities with 95% confidence intervals is given for six other radionuclides ({sup 9}Sr, {sup 21}Pb({sup 21}Po), {sup 226}Ra, {sup 239}Pu, {sup 24}Pu {sup 241}Pu). Less frequently reported radionuclides ({sup 99}Tc, {sup 129}I, {sup 228}Th, {sup 23}Th and {sup 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.

  4. Lagoon sediment radioactivity in Polynesian French nuclear test sites

    In 1996, at the request of the French government, IAEA conducted an international study to assess the radiological situation of the atolls of Mururoa and Fangataufa. Results obtained by French scientists in the previous years were provided for the study. In this paper, some new results, such as 240Pu/239Pu activity ratio, are reported. 27 top-layer sediments were sampled in 1993 and 1995 in Mururoa and Fangataufa lagoons. Radioactivity measurements of the major man-made γ-emitters (60Co, 125Sb, 137Cs and 155Eu), of 90Sr, 99Tc and 239Pu, together with activity ratios (238Pu/239Pu, 240Pu/239Pu, 241Am/239Pu and 237Np/241Am) are given. These results are compared with previous measurements, in particular the radiological situations established in years 1984-90. Radiological maps, 155Eu concentration , 239,240Pu concentration and 238Pu/239,240Pu activity ratio, for sediment radioactivity in Mururoa and Fangataufa lagoons are shown. Results of two cores sampled in 1996 during the IAEA study are also reported. An assessment of total inventories is given for both atolls. (author)

  5. Radionuclide adsorption distribution coefficients measured in Hanford sediments for the low level waste performance assessment project

    Preliminary modeling efforts for the Hanford Site's Low Level Waste-Performance Assessment (LLW PA) identified 129I, 237Np, 79Se, 99Tc, and 234,235,238U as posing the greatest potential health hazard. It was also determined that the outcome of these simulations was very sensitive to the parameter describing the extent to which radionuclides sorb to the subsurface matrix, i.e., the distribution coefficient (Kd). The distribution coefficient is a ratio of the radionuclide concentration associated with the solid phase to that in the liquid phase. The objectives of this study were to (1) measure iodine, neptunium, technetium, and uranium Kd values using laboratory conditions similar to those expected at the LLW PA disposal site, and (2) evaluate the effect of selected environmental parameters, such as pH, ionic strength, moisture concentration, and radio nuclide concentration, on Kd values of selected radionuclides. It is the intent of these studies to develop technically defensible Kd values for the PA. The approach taken throughout these studies was to measure the key radio nuclide Kd values as a function of several environmental parameters likely to affect their values. Such an approach provides technical defensibility by identifying the mechanisms responsible for trends in Kd values. Additionally, such studies provide valuable guidance regarding the range of Kd values likely to be encountered in the proposed disposal site

  6. Semianalytical solutions of radioactive or reactive tracer transport in layered fractured media

    Moridis, G.J.; Bodvarsson, G.S.

    2001-10-10

    In this paper, semianalytical solutions are developed for the problem of transport of radioactive or reactive tracers (solutes or colloids) through a layered system of heterogeneous fractured media with misaligned fractures. The tracer transport equations in the matrix account for (a) diffusion, (b) surface diffusion (for solutes only), (c) mass transfer between the mobile and immobile water fractions, (d) linear kinetic or equilibrium physical, chemical, or combined solute sorption or colloid filtration, and (e) radioactive decay or first order chemical reactions. Any number of radioactive decay daughter products (or products of a linear, first-order reaction chain) can be tracked. The tracer-transport equations in the fractures account for the same processes, in addition to advection and hydrodynamic dispersion. Additionally, the colloid transport equations account for straining and velocity adjustments related to the colloidal size. The solutions, which are analytical in the Laplace space, are numerically inverted to provide the solution in time and can accommodate any number of fractured and/or porous layers. The solutions are verified using analytical solutions for limiting cases of solute and colloid transport through fractured and porous media. The effect of important parameters on the transport of {sup 3}H, {sup 237}Np and {sup 239}Pu (and its daughters) is investigated in several test problems involving layered geological systems of varying complexity. {sup 239}Pu colloid transport problems in multilayered systems indicate significant colloid accumulations at straining interfaces but much faster transport of the colloid than the corresponding strongly sorbing solute species.

  7. DEVELOPMENT OF AN IMPROVED TITANATE-BASED SORBENT FOR STRONTIUM AND ACTINIDE SEPARATIONS UNDER STRONGLY ALKALINE CONDITIONS

    Hobbs, D.; Peters, T.; Taylor-Pashow, K.; Fink, S.

    2010-02-18

    High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove {sup 134,137}Cs, {sup 90}Sr, and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. Separation processes at SRS include the sorption of {sup 90}Sr and alpha-emitting radionuclides onto monosodium titanate (MST) and caustic side solvent extraction of {sup 137}Cs. The MST and separated {sup 137}Cs is encapsulated along with the sludge fraction of high-level waste (HLW) into a borosilicate glass waste form for eventual entombment at a federal repository. The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes {sup 238}Pu, {sup 239}Pu, and {sup 240}Pu; {sup 237}Np; and uranium isotopes, {sup 235}U and {sup 238}U. This paper describes recent results evaluating the performance of an improved sodium titanate material that exhibits increased removal kinetics and capacity for {sup 90}Sr and alpha-emitting radionuclides compared to the current baseline material, MST.

  8. Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction 237Np(n,f)F.P. 1.2 The reaction is useful for measuring neutrons with energies from approximately 0.7 to 6 MeV and for irradiation times up to 30 to 40 years. 1.3 Equivalent fission neutron fluence rates as defined in Practice E 261 can be determined. 1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E 261. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 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.

  9. Leaching sensitivity to geologic environmental parameters. Task 3. Characterization of radioactive waste forms. A series of final reports (1985-89). No 14

    This report presents the results obtained in the optional part of the 'Repository system simulation test' (RSST). The results of the compulsory part have been reported and evaluated in another publication (EUR 12544). The effects of the oxidation-reduction potential and carbonate concentration on the alterability and containment properties of R7T7 glass were investigated under semi-integral conditions, i.e. by leaching glass samples doped with 239Pu and 237Np in the presence of environmental materials (sand, clay and granite) in three different test media: . a carbonated medium obtained by adding sodium bicarbonate; . an oxidative medium with a 2-bar oxygen partial pressure; . a reducing medium with a 2-bar partial pressure of argon and 5% H2 additive. Under the semi-integral test conditions, the normalized Np and Pu mass losses varied by less than an order of magnitude for the different test media, but the following general tendencies were observed: NL(Np)CO2 > NL(Np)O2 > NL(Np)H2 and NL(Pu)O2 > NL(Pu)CO2 > NL(Pu)H2. A large fraction of the released activity was incorporated by the environmental materials present in the conditioning cell, notably by the clay. The actinide concentrations in the interstitial water and the amounts fixed on the granite were low

  10. Distribution coefficient values describing iodine, neptunium, selenium, technetium, and uranium sorption to Hanford sediments. Supplement 1

    Burial of vitrified low-level waste (LLW) in the vadose zone of the Hanford Site is being considered as a long-term disposal option. Regulations dealing with LLW disposal require that performance assessment (PA) analyses be conducted. Preliminary modeling efforts for the Hanford Site LLW PA were conducted to evaluate the potential health risk of a number of radionuclides, including Ac, Am, C, Ce, Cm, Co, Cs, Eu, 1, Nb, Ni, Np, Pa, Pb, Pu, Ra, Ru, Se, Sn, Sr, Tc, Th, U, and Zr (Piepho et al. 1994). The radionuclides, 129I, 237Np, 79Se, 99Tc, and 234,235,238U, were identified as posing the greatest potential health hazard. It was also determined that the outcome of these simulations were very sensitive to the parameter describing the extent to which radionuclides sorbed to the subsurface matrix, described as a distribution coefficient (Kd). The distribution coefficient is a ratio of the radionuclide concentration associated with the solid phase to that in the liquid phase. The literature-derived Kd values used in these simulations were conservative, i.e., lowest values within the range of reasonable values used to provide an estimate of the maximum health threat. Thus, these preliminary modeling results reflect a conservative estimate rather than a best estimate of what is likely to occur. The potential problem with providing only a conservative estimate is that it may mislead us into directing resources to resolve nonexisting problems

  11. Determination of actinides in environmental and biological samples using high-performance chelation ion chromatography coupled to sector-field inductively coupled plasma mass spectrometry.

    Truscott, J B; Jones, P; Fairman, B E; Evans, E H

    2001-08-31

    High-performance chelation ion chromatography, using a neutral polystyrene substrate dynamically loaded with 0.1 mM dipicolinic acid, coupled with sector-field inductively coupled plasma mass spectrometry has been successfully used for the separation of the actinides thorium, uranium, americium, neptunium and plutonium. Using this column it was possible to separate the various actinides from each other and from a complex sample matrix. In particular, it was possible to separate plutonium and uranium to facilitate the detection of the former free of spectral interference. The column also exhibited some selectivity for different oxidation states of Np, Pu and U. Two oxidation states each for plutonium and neptunium were found, tentatively identified as Np(V) and Pu(III) eluting at the solvent front, and Np(IV) and Pu(IV) eluting much later. Detection limits were 12, 8, and 4 fg for 237Np, 239Pu, and 241Am, respectively, for a 0.5 ml injection. The system was successfully used for the determination of 239Pu in NIST 4251 Human Lung and 4353 Rocky Flats Soil, with results of 570+/-29 and 2939+/-226 fg g(-1), respectively, compared with a certified range of 227-951 fg g(-1) for the former and a value of 3307+/-248 fg g(-1) for the latter. PMID:11589474

  12. Trace, ultratrace and isotope analysis of long-lived radionuclides by laser ablation inductively coupled plasma mass spectrometry

    Laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) has become established as a very efficient and sensitive technique for the analysis of solids. For the determination of long-lived radionuclides in solid nuclear waste or contaminated environmental samples LA-ICP-MS is the method of choice. The capability of LA-ICP-MS for measurements on long-lived radionuclides in non-conducting concrete matrix, which is a very common matrix in waste packages will be investigated. Of special interest are the limits of detection of long-lived radionuclides, which are compared for two different types of mass spectrometer coupled to a commercial laser ablation system. The limits of detection of long-lived radionuclides investigated in concrete matrix are determined in the low pg g-1 range in quadrupole LA-ICP-MS and in double-focusing sector field LA-ICP-MS. The main problem in the quantification of analytical results is that no suitable standard reference materials are available. Therefore synthetic laboratory standards (concrete matrix doped with long-lived radionuclides, such as 99Tc, 232Th, 233U, 235U, 237Np, 238U) were investigated by LA-ICP-MS. Different calibration procedures - the correction of analytical results with experimentally determined relative sensitivity coefficients (RSCs), the use of calibration curves and solution calibration by coupling LA-ICP-MS with an ultrasonic nebulizer - were applied for the determination of long-lived radionuclides, especially for Th and U in different solid samples. (orig.)

  13. Isotope ratio measurement of uranium in safeguards environmental samples by inductively-coupled plasma mass spectrometry (ICP-MS)

    In order to measure isotope ratio of uranium in safeguards environmental samples with ICP-MS precisely, production of polyatomic ions of IrAr, PtAr and AuAr was measured and mass bias of ICP-MS is investigated by using isotopic standards of uranium and lead. The intensities of IrAr, PtAr and AuAr relative to the atomic ions were found to be 1.8 x 10-6, 1.6 x 10-5 and 4.1 x 10-5, respectively. The production of 193Ir40Ar is too small to interfere with the measurement of 233U, if the concentration of Ir is the same level as that of 233U. However, there is possibility that the presence of Pt and Au interferes with the measurement of minor isotopes of uranium and 237Np. On the other hand, the mass biases of 235U/238U and 208Pb/206Pb were measured with the parameter of 238U16O/238U. Since unexpected change of the mass bias during measurements causes frequently erroneous results, the monitoring of 238U16O/238U is effective for the precise isotope ratio measurement. (author)

  14. Actinide behavior in the Integral Fast Reactor. Final project report

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides (237Np, 240Pu, 241Am, and 243Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  15. Anthropogenic radionuclides in the environment

    Hu, Q; Weng, J; Wang, J

    2007-11-15

    Studies of radionuclides in the environment have entered a new era with the renaissance of nuclear energy and associated fuel reprocessing, geological disposal of high-level nuclear wastes, and concerns about national security with respect to nuclear non-proliferation. This work presents an overview of anthropogenic radionuclide contamination in the environment, as well as the salient geochemical behavior of important radionuclides. We first discuss the following major anthropogenic sources and current development that contribute to the radionuclide contamination of the environment: (1) nuclear weapons program; (2) nuclear weapons testing; (3) nuclear power plants; (4) commercial fuel reprocessing; (5) geological repository of high-level nuclear wastes, and (6) nuclear accidents. Then, we summarize the geochemical behavior for radionuclides {sup 99}Tc, {sup 129}I, and {sup 237}Np, because of their complex geochemical behavior, long half-lives, and presumably high mobility in the environment. Biogeochemical cycling and environment risk assessment must take into account speciation of these redox-sensitive radionuclides.

  16. Activation calculation and waste management for a fusion experimental breeder, FEB-E

    The engineering outline design of the fusion experimental breeder, FEB-E is reported briefly. Using an activation calculation code FDKR and its associated data library AF-DCDLIB to calculate the radioactivity, decay heat, waste disposal rating and biological hazard potential from activation products, actinides and fission products in the FEB-E. The codes and libraries used in calculation are introduced briefly, and calculation results and decay curves of related hazards after the shutdown of one-year operation of the FEB-E are given. The activation features of five candidate structural materials were also evaluated for FEB-E design. Detailed calculation and analyses of waste disposal rating and remote maintenance rating for all long-lived radionuclides were performed to identify the safety, environmental and radioactive waste characteristics of the FEB-E design. Results obtained show that the total radioactivity inventory, decay heat and BHP at shutdown are 5.74x1013MBq, 8.34 MW and 4.08x108km3 of air for 316SS structure material, respectively. The inventory of actinides important for the fuel reprocessing, such as 232U and 237Np was also investigated. It was shown that their low concentrations in FEB-E appear to be manageable

  17. BCR certified reference materials for reactor neutron dosimetry

    A series of reference materials intended for use as activation or fission monitors for neutron fluence rate measurements has been prepared by the Joint Research Centre of the European Commission. Certification has been carried out by expert European laboratories and distribution of the certified reference materials (CRMs) is through the BCR programme of the Commission. The list (18 CRMs) includes materials to cover the complete energy spectrum, and suitable for different irradiation times. Fission monitors are 238UO2 or 237NpO2 in the form of microspheres. Activation monitors are high purity metals (Ni, Cu, Al, Fe, Nb, Rh, or Ti), certified for interfering trace impurities, or dilute aluminium-based alloys, where aluminium is chosen as a suitable matrix for reducing the neutron self-shielding effect. Newly certified materials are IRMM-530R Al-0.1%Au, replacing the exhausted IRMM-530 material, used as comparator for k0-standardization, and three new Al-Co alloys (0.01-1%Co). Two others, in the process of certification are Al-0.1%Ag and Al-2%Sc for thermal and epithermal fluence rate measurements. Other candidate reference materials currently being certified are two uranium-doped glass intended for dosimetry by the fission-track technique. (author)

  18. Analysis of environmental friendliness of DUPIC fuel cycle

    Some properties of irradiated DUPIC fuels are compared with those of other fuel cycles. It was indicated that the toxicity of the DUPIC option based on 1 GWe-yr is much smaller than those of other fuel cycle options, and is just about half the order of magnitude of other fuel cycles. From the activity analysis of 99Tc and 237Np, which are important to the long-term transport of fission products stored in geologic media, the DUPIC option, was being contained only about half of those other options. It was found from the actinide content estimation that the MOX option has the lowest plutonium arising based on 1 GWe-year and followed by the DUPIC option. However, fissile Pu content generated in the DUPIC fuel was the lowest among the fuel cycle options. From the analysis of radiation barrier in proliferation resistance aspect, the fresh DUPIC fuel can play a radiation barrier part, better than CANDU spent fuels as well as fresh MOX fuel. It is indicated that the DUPIC fuel cycle has the excellent resistance to proliferation, compared with an existing reprocessing option and CANDU once-through option. In conclusions, DUPIC fuel cycle would have good properties on environmental effect and proliferation resistance, compared to other fuel cycle cases

  19. Calculation of fission reactions within the MM-RNR model

    In the past fission-fragment properties and cross-sections for 235,238U(n, f), 237Np(n, f), 239Pu(n, f) and 252Cf(SF) have been investigated. The interpretation of the experimental data in the frame of the multi-modal random neck-rupture (MM-RNR) model has been incorporated into the most recent evaluation exercise on neutron-induced fission cross-section and prompt-neutron emission data in the actinide region of the chart of nuclides. The three most dominant fission modes were considered, the two asymmetric standard I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode, namely. Except for 252Cf, de-convoluted modal fission cross-sections as well as prompt neutron multiplicity and spectra have been calculated in the energy range from 0.01 MeV to 5.5 MeV in excellent agreement with experimental data. In addition, the obtained fission-mode branching ratios allow the calculation of fission-fragment yield and energy distributions where no experimental data exist. Most recently the first ever-direct measurement of the neutron-induced fission cross-section of 233Pa has been performed at IRMM. The subsequent evaluation suggests a radical revision of today's evaluated data files. (author)

  20. PERFORMANCE OF THE SAVANNAH RIVER SITE COULOMETER FOR NEPTUNIUM PROCESSACCOUNTABILITY AND NEPTUNIUM OXIDE PRODUCT CHARACTERIZATION

    Holland, M; Patterson Nuessle, P; Sheldon Nichols, S; Joe Cordaro, J; George Reeves, G

    2008-06-04

    The Savannah River Site's (SRS) H-Area B-Line (HB-Line) nuclear facility is processing neptunium solutions for stabilization as an oxide. The oxide will eventually be reprocessed and fabricated into target material and the 237Np irradiated to produce {sup 238}Pu in support of National Aeronautics and Space Administration space program missions. As part of nuclear materials accountability, solution concentrations were measured using a high-precision controlled-potential coulometer developed and manufactured at the SRS for plutonium accountability measurements. The Savannah River Site Coulometer system and measurement methodology for plutonium meets performance standards in ISO 12183-2005, 'Controlled-Potential Coulometric Assay of Plutonium'. The Department of Energy (DOE) does not produce or supply a neptunium metal certified reference material, which makes qualifying a measurement method and determining accuracy and precision difficult. Testing and performance of the Savannah River Site Coulometer indicates that it can be used to measure neptunium process solutions and dissolved neptunium oxide without purification for material control and accountability purposes. Savannah River Site's Material Control and Accountability organization has accepted the method uncertainty for accountability and product characterization measurements.

  1. Disposal of radioactive waste into clay layers the most natural option

    Among the geological formations suitable for the disposal of radioactive waste, the clay formations provide outstanding opportunities : impermeable for water, self-healing, strongly absorbing for ions, widespread in nature. The self-healing properties of large clay deposits have been demonstrated by their auto-sealing and plastic response to tectonic stress and magmatic intrusion. The discovery of fossil trees preserved after geologic periods of burial in clay is one of the most dramatic illustrations of their entombment ability. The physicochemical and hydrologic characteristics of the Boom clay are very favorable for the confinement of migrating radionuclides within the layer. Except for the extremely long half-lives (237Np, 129I,...) no radionuclide can escape from the clay body. The effects of heat, metal corrosion, material interaction and biochemical degradation on the natural properties of the clay layer are discussed in some detail and related to the natural properties of the clay formation which have to stay unaltered for geologic periods. The first Safety Assessment Report, established by NIRAS-ONDRAF in close collaboration with SCK-CEN, has been submitted to a multi-disciplinary task force which is to advise the Belgian Government on the suitability of the Boom clay layer below the Nuclear Research site of Mol as a potential host formation for nuclear waste coming from the electronuclear program. 13 refs., 2 figs., 1 tab

  2. Development of a TES microcalorimeter for spectroscopic measurement of LX-rays emitted by transuranium elements

    A phase transition edge sensor (TES) microcalorimeter was developed for the energy-dispersive measurement of LX-ray photons emitted by transuranium elements. The phase transition temperature of the TES was designed to be 200 mK using a bilayer structure of Au of 120 nm thickness and Ti of 50 nm thickness. A Au layer of 5.0 μm thickness was deposited on the Au/Ti bilayer to achieve an absorption efficiency of 50% and counting rate of 100 counts per second in the detection of LX-ray photons with energy from 10 to 20 keV. The TES microcalorimeter was operated for the detection of LX-ray photons emitted by 241Am, 238Pu, and 239Pu sources. A decay time constant of 180 μs for the detection signal pulses allowed the TES microcalorimeter to operate with a counting rate higher than 100 counts per second. The achieved energy resolution was 50 eV for the full width at half maximum of a peak corresponding to a 237Np Lβ1 X-ray of 17.75 keV. (author)

  3. Validity of the generalized Brink-Axel hypothesis in $^{238}$Np

    Guttormsen, M; Görgen, A; Renstrøm, T; Siem, S; Tornyi, T G; Tveten, G M

    2016-01-01

    We have analyzed primary $\\gamma$-ray spectra of the odd-odd $^{238}$Np nucleus extracted from $^{237}$Np($d,p\\gamma$)$^{238}$Np coincidence data measured at the Oslo Cyclotron Laboratory. The primary $\\gamma$ spectra cover an excitation-energy region of $0 \\leq E_i \\leq 5.4$ MeV, and allowed us to perform a detailed study of the $\\gamma$-ray strength as function of excitation energy. Hence, we could test the validity of the generalized Brink-Axel hypothesis, which, in its strictest form, claims no excitation-energy dependence on the $\\gamma$ strength. In this work, using the available high-quality $^{238}$Np data, we show that the $\\gamma$-ray strength function is to a very large extent independent on the initial and final states. Thus, for the first time, the generalized Brink-Axel hypothesis has been experimentally verified for $\\gamma$ transitions between states in the quasi-continuum region, not only for specific collective resonances, but also for the full strength below the neutron separation energy. B...

  4. Np Analysis in IAT-Samples Containing <10 Microgram Pu

    A method for the determination of neptunium to plutonium in safeguards samples containing less than 10 microgram Pu is presented. The chemical treatment and the optimized measurement conditions for gamma spectrometry are reported. This method is based on thermal ionization mass spectrometry (TIMS) after chemical treatment and separation and was validated with mixtures of U, Pu and Np certified reference materials and using the 237Np standard addition method, followed by separation of the waste fraction and gamma spectrometric analysis. The highest sensitivity, precision and accuracy in the determination of the Np:Pu ratio at microgram levels of Pu is achieved by evaluating 241Pu and 233Pa after measuring the adsorbent with a well-type gamma detector 3 weeks after chemical treatment. The repeatability of determining the Np:Pu ratio is estimated to be 5%, the maximum uncertainty as determined from comparing the 4 measurement modes is within ± 10% for samples containing 3 μg Pu, while being within ± 20% for 0.4 μg Pu. (authors)

  5. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of 232Th and 237Np, measured in GFR-like lattices. (authors)

  6. Evaluation of radiation exposure due to radioactive releases into the biosphere from a salt dome repository

    Potential health consequences of hypothetical releases from a salt dome repository planned in the Federal Republic of Germany have been evaluated taking into account the following parameters: (a) contamination of near-surface groundwater by transport of radionuclides with groundwater movement from the salt dome, (b) irrigation of agricultural areas and radionuclide uptake by plants, (c) intake of activity by man through food contaminated by use of water for irrigation, through drinking water and fish consumption, and (d) internal exposure for members of the population. For quantitative assessment of long-lived radionuclide concentrations in food crops a time-dependent irrigation model was developed and used in this project. It takes into account agricultural practices used in the northern part of the Federal Republic of Germany. For the important long-lived nuclides potential, individual doses were calculated for members of the local population. The uncertainties of this approach are discussed with reference to 237Np. Conditions of agricultural production in the far future are a more uncertain factor than the variability of the biological parameters considered. It is doubtful whether, in general, the accuracy of such calculations can permit any quantitative mathematical procedure for the optimization of radiation protection measures in waste disposal options. (author)

  7. Studies of heavy radioactive nuclides transmutation using relativistic particles at the JINR (Dubna)

    We overview briefly the investigations of the process of transmutation of long-lived fission products into short-lived or stable nuclides, which are conducted for several years at the synchrophasotron of the Joint Institute for Nuclear Research (JINR), Dubna, using the beams of protons and 12C ions of 3.67 GeV and 18 GeV energy, respectively. As a result of these experiments it has been found that the measured transmutation rates for these nuclides irradiated with the help of rather simple and versatile experimental arrangement allows to estimate the transmutation rate capacity for a 10 m A -1.5 GeV proton accelerator coupled to a Pb target as: 4% of 129I into 130Xe for one year and 6% of 237Np into 238Np during one month. Current methodical and computational problems arising in the relevant works are also discussed. Some remarks concerning similar investigations conducted in other laboratories (in particular, at CERN [3]) are made, too

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

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

  9. Single-pass continuous-flow leach test of PNL 76-68 glass: some selected Bead Leach I results

    A single-pass continuous-flow leach test of PNL 76-68 glass beads (7 mm dia) was concluded after 420 days of uninterrupted operation. Variables included in the experimental matrix were flow-rate, leachant composition, and temperature. Analysis was conducted on all leachate samples for 237Np and 239Pu as well as a number of nonradioactive elements. Results indicated that flow-rate and leachant systematically affected the leach rate, but only slightly. Temperature effects were significant. Plutonium leach rate was lower at higher temperature suggesting that Pu sorption onto the beads was enhanced at the higher temperature. The range of leach rates for all analyzed elements (except Pu), at both temperatures, at all three flow rates, and with all three leachant compositions varied over only three orders of magnitude. The range of variables used in this experiment covered those expected in many proposed repository environments. The preliminary interpretation of the results aPPh3 also reacted with Mn2(CO)10 and Cp2Mo2(CO)6 to give a variety of products at room temperature. A radical mechanism was suggested

  10. Plan and progress of a cooperative research program on field migration test between CIRP and JAERI

    As the main parts of a cooperative research project on safety assessment method for shallow land disposal of low level radioactive wastes between the China Institute for radiation Protection (CIRP) and the Japan Atomic Energy Research Institute (JAERI), field migration tests investigating the migration of radionuclides 90Sr, 237Np and 238Pu and stable elements Sr, Ce, and Nd in both aerated zone and aquifer are carried out. 3H and Br are used as tracers to obtain the water flow velocity in the aquifer and the movement of moisture water in the aerated zone for radioactive and non-radioactive tests, respectively. The aerated zone migration tests are carried out with media including soil, bentonite, cement block and mortar in both natural rainfall condition and artificial rainfall condition. The aquifer migration tests are carried out with assemblies (Type I), experimental frame (Type II) and injection tube (Type III), respectively, in and Underground Research Facility (URF). The aerated zone migration tests were started in May, 1997, while the aquifer migration tests started in August 1997. (author)

  11. Anticipating Potential Waste Acceptance Criteria for Defense Spent Nuclear Fuel

    Rechard, R.P.; Lord, M.E.; Stockman, C.T. [Sandia National Labs., Albuquerque, NM (United States). Nuclear Waste Management Center; McCurley, R.D. [New Mexico Univ., Albuquerque, NM (United States). New Mexico Engineering Research Institute

    1997-12-31

    The Office of Environmental Management of the U.S. Department of Energy is responsible for the safe management and disposal of DOE owned defense spent nuclear fuel and high level waste (DSNF/DHLW). A desirable option, direct disposal of the waste in the potential repository at Yucca Mountain, depends on the final waste acceptance criteria, which will be set by DOE`s Office of Civilian Radioactive Waste Management (OCRWM). However, evolving regulations make it difficult to determine what the final acceptance criteria will be. A method of anticipating waste acceptance criteria is to gain an understanding of the DOE owned waste types and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent fuel. Preliminary results from such an analysis indicate that releases of 99Tc and 237Np from commercial spent fuel exceed those of the DSNF/DHLW; thus, if commercial spent fuel can meet the waste acceptance criteria, then DSNF can also meet the criteria. In large part, these results are caused by the small percentage of total activity of the DSNF in the repository (1.5%) and regulatory mass (4%), and also because commercial fuel cladding was assumed to provide no protection.

  12. SOLID PHASE MICROEXTRACTION SAMPLING OF HIGH EXPLOSIVE RESIDUES IN THE PRESENCE OF RADIONUCLIDES AND RADIONUCLIDE SURROGATE METALS

    Duff, M; S Crump, S; Robert02 Ray, R; Donna Beals, D

    2007-04-13

    The Federal Bureau of Investigation (FBI) Laboratory currently does not have on site facilities for handling radioactive evidentiary materials and there are no established FBI methods or procedures for decontaminating high explosive (HE) evidence while maintaining evidentiary value. One experimental method for the isolation of HE residue involves using solid phase microextraction or SPME fibers to remove residue of interest. Due to their high affinity for organics, SPME fibers should have little affinity for most metals. However, no studies have measured the affinity of radionuclides for SPME fibers. The focus of this research was to examine the affinity of dissolved radionuclide ({sup 239/240}Pu, {sup 238}U, {sup 237}Np, {sup 85}Sr, {sup 133}Ba, {sup 137}Cs, {sup 60}Co and {sup 226}Ra) and stable radionuclide surrogate metals (Sr, Co, Ir, Re, Ni, Ba, Cs, Nb, Zr, Ru, and Nd) for SPME fibers at the exposure conditions that favor the uptake of HE residues. Our results from radiochemical and mass spectrometric analyses indicate these metals have little measurable affinity for these SPME fibers during conditions that are conducive to HE residue uptake with subsequent analysis by liquid or gas phase chromatography with mass spectrometric detection.

  13. Measurement of cross-sections of fission reactions induced by neutrons on actinides from the thorium cycle at n-TOF facility; Mesures de sections efficaces de fission induite par neutrons sur des actinides du cycle du thorium a n-TOF

    Ferrant, L

    2005-09-01

    In the frame of innovating energy source system studies, thorium fuel cycle reactors are considered. Neutron induced fission cross section on such cycle involved actinides play a role in scenario studies. To feed them, data bases are built with experimental results and nuclear models. For some nuclei, they are not complete or in disagreement. In order to complete these data bases, we have built an original set up, consisting in an alternation of PPACs (Parallel Plate Avalanche Chamber) and ultra - thin targets, which we installed on n-TOF facility. We describe detectors, set up, and the particular care brought to target making and characterization. Fission products in coincidence are detected with precise time measurement and localization with delay line read out method. We contributed, within the n-TOF collaboration, to the CERN brand new intense spallation neutron source characterization, based on time of flight measurement, and we describe its characteristics and performances. We were able to measure such actinide fission cross sections as {sup 232}Th, {sup 234}U, {sup 233}U, {sup 237}Np, {sup 209}Bi, and {sup nat}Pb relative to {sup 235}U et {sup 238}U standards, using an innovative acquisition system. We took advantage of the lame accessible energy field, from 0.7 eV to 1 GeV, combined with the excellent energy resolution in this field. Data treatment and analysis advancement are described to enlighten performance and limits of the obtained results. (author)

  14. Photofission Analysis for Fissile Dosimeters Dedicated to Reactor Pressure Vessel Surveillance

    Bourganel, Stéphane; Faucher, Margaux; Thiollay, Nicolas

    2016-02-01

    Fissile dosimeters are commonly used in reactor pressure vessel surveillance programs. In this paper, the photofission contribution is analyzed for in-vessel 237Np and 238U fissile dosimeters in French PWR. The aim is to reassess this contribution using recent tools (the TRIPOLI-4 Monte Carlo code) and latest nuclear data (JEFF3.1.1 and ENDF/B-VII nuclear libraries). To be as exhaustive as possible, this study is carried out for different configurations of fissile dosimeters, irradiated inside different kinds of PWR: 900 MWe, 1300 MWe, and 1450 MWe. Calculation of photofission rate in dosimeters does not present a major problem using the TRIPOLI-4® Monte Carlo code and the coupled neutron-photon simulation mode. However, preliminary studies were necessary to identify the origin of photons responsible of photofissions in dosimeters in relation to the photofission threshold reaction (around 5 MeV). It appears that the main contribution of high enough energy photons generating photofissions is the neutron inelastic scattering in stainless steel reactor structures. By contrast, 137Cs activity calculation is not an easy task since photofission yield data are known with high uncertainty.

  15. Leaching of actinide elements from simulated fuel debris into seawater

    For the prediction of the leaching behavior of actinide elements contained in the fuel debris that has arisen from the severe accident in Fukushima Daiichi Nuclear Power Station (NPS), a simulated fuel debris consisting of UO2 - ZrO2 solid solution doped with 137Cs, 237Np, 236Pu and 241Am tracers was synthesized, and agitated leaching tests were conducted for the simulated fuel debris in seawater. The synthesized simulated fuel debris was immersed and shaken in natural seawater collected at a coast 11 km away from Fukushima Daiichi NPS. The brief leaching test conditions were T = 25°C and solid-liquid ratio = 4 g/l, and the test duration was up to 31 days. The ratio of tracers leached into seawater from the simulated fuel debris by the agitated leaching test for 4 days was evaluated to be 0.09% for U, 0.01% for Np, 0.01% for Pu, 0.01% for Am and 35.39% for Cs by the α or γ spectrometry of the soluble fraction. The leaching of actinides from the real fuel debris in reactor units 1 - 3 in Fukushima Daiichi NPS is expected to be suppressed in comparison with that from normal light water reactor spent fuel. (author)

  16. Numerical methods to analyze alpha spectra and application to the study of neptunium 237 and neptunium 236

    A set of numerical methods to analize alpha spectra measured with semiconductor detectors are presented. The methods can be divided in two groups, the first being based in the X2 minimization ands the second in the use of the Fourier Transform. The methods based in the minimization of X2 can, in turn, be divided according to the model used to fit the spectra. Some of them use a monoenergetic line for the intercomparison with the other peaks in the same spectrum. The others take into account the analytical function developed to represent an alpha line. Both allow the determination of positions and areas of the components, as well as the uncertainties of the results. The Fast Fourier Transform is applied to the second group of methods, which include the smoothing of experimental data, and the deconvolution of spectra. Examples are given of the application of these methods to real spectra. The alpha spectra of 237Np and 236Np are studied by using some of the methods described in this work. (Author)

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

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

    2015-07-01

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

  18. The effect of concentration and chemical form on the gastrointestinal absorption of neptunium

    Gastrointestinal absorption of neptunium at low mass concentrations as 239Np nitrate (0.5-1 ng Np; 2-5 μg 1-1) has been found to be 0.03, 0.02 and 0.18% in adult rats, hamsters and rabbits, respectively. Administration as 239Np bicarbonate increased uptake in the rat to 0.15% but had no significant effect on uptake in either the hamster or the rabbit. Absorption in the rat was also increased to 0.14% for 239Np citrate but not for either 239Np phytate (0.04%) or 239Np incorporated into rat liver (0.01%). The fasting of rats for 8 hours increased their subsequent absorption of 239Np as the bicarbonate to 0.25%. The absorption of neptunium at high mass concentrations as 237Np nitrate (0.5 mg Np, 5 g 1-1) was increased to 0.26% in the rat compared with the value of 0.03% for 239Np (0.5 ng, 5 μg 1-1) but a similar effect of concentration was not observed in the hamster. The results support the use of a value of absorption of 0.1% instead of 1% in calculations of annual limits on intake of radioisotopes of neptunium by workers and in estimates of radiation doses resulting from the ingestion of contaminated food and water by adult members of the public. (author)

  19. Biological pathways and chemical behavior of plutonium and other actinides in the environment

    The principal long-lived actinide elements that may enter the environment from either U or Pu fuel cycles are Pu, Am, Cm, and Np. Approximately 25% of the alpha activity estimated to be released to the atmosphere from the LMFBR fuel cycle will be contributed by 241Am, 242Cm, and 244Cm. The balance of the alpha activity will come from Pu isotopes. Activities of 242Cm, 244Cm, 241Am, 243Am, and 237Np in waste may exceed concentrations of Pu isotopes in waste after various periods of decay. Thorium and uranium isotopes may also be released by operations of the thorium fuel cycle. Environmental actinides are discussed under the following headings: sources of man-made actinide elements; pathways of exposure; environmental chemistry of actinides; uptake of actinides by plants; distribution of actinides in components of White Oak Lake; entry of actinides into terrestrial food chains; relationship between chemical behavior and uptake of actinides by organisms; and behavior of Pu in freshwater and marine food chains

  20. Fission of Oriented Nuclei by Low Energy Neutrons. RCN Report

    This report describes the study of the angular distribution of α-particles and of fission fragments originating from neutron capture in heavy nuclei, which are aligned at low temperatures by the method of hyperfine interaction. The results of the measurements with the target nucleus 233U with neutrons in the energy range from 0 to 2000 eV can be interpreted with the Bohr-theory of transition states at the deformation barrier for nuclear fission. The relatively invariant behaviour of the anisotropy in the angular distribution of fission fragments as a function of neutron energy indicates that the available fission channels are strongly mixed. For neutron resonances with spin and parity 2+ 2 to 3 channels are open and for 3+ resonances 1 to 2. The group structure in the subthreshold fission cross section of 237Np has been explained by the double-humped deformation barrier proposed by Strutinsky. The implication of this interpretation is that all the resonances in one group have the same spin. The resonances in the first group at 40 eV agree consistently with the fission channel (2+,2). The groups at higher neutron energies up to 2000 eV correspond mainly with the channels (2+,2) and (3+,2). (author)

  1. Century of radiochemistry. History and future

    Several periods in radiochemistry development history are observed. In the report the sources of radionuclide income into environment are examined including long-lived transuranic elements on the different stages of full nuclear fuel cycle. Radioactive substance contamination analysis is given for different regions of Russia from natural and man-caused sources. Potential danger of long-lived radionuclide and transuranic element presence in the wastes of nuclear fuel cycle plants is shown. Data related with the sequences of nuclear weapon testing on the proving grounds near Semipalatinsk, and Novaya Zemlya are presented. The modern radioecological situation around the reprocessing plant 'Mayak', which was constructed more than 40 years ago for the production of plutonium for military purposes, is overviewed. The following topics are considered: lake Karachay; artificial water reservoirs contaminated by radionuclides; solid radioactive wastes and their vitrification. Some new approaches, methods and tools developed at the Vernadsky Institute of Russian Academy of Sciences for determination of different radionuclides in various environmental samples from the impact zone of the facility are discussed. The data on distribution, occurrence forms and migration processes of 90Sr, 137Cs, 237Np, 239Pu, and 241Am in aquatic and terrestrial ecosystems are presented. (author)

  2. Interim salt disposition program macrobatch 6 tank 21H qualification monosodium titanate and cesium mass transfer tests

    Savannah River National Laboratory (SRNL) performed experiments on qualification material for use in the Interim Salt Disposition Program (ISDP) Batch 6 processing. This qualification material was a set of six samples from Tank 21H in October 2012. This sample was used as a real waste demonstration of the Actinide Removal Process (ARP) and the Extraction-Scrub-Strip (ESS) tests process. The Tank 21H sample was contacted with a reduced amount (0.2 g/L) of MST and characterized for strontium and actinide removal at 0 and 8 hour time intervals in this salt batch. 237Np and 243Am were both observed to be below detection limits in the source material, and so these results are not reported in this report. The plutonium and uranium samples had decontamination factor (DF) values that were on par or slightly better than we expected from Batch 5. The strontium DF values are slightly lower than expected but still in an acceptable range. The Extraction, Scrub, and Strip (ESS) testing demonstrated cesium removal, stripping and scrubbing within the acceptable range. Overall, the testing indicated that cesium removal is comparable to prior batches at MCU

  3. Heavy element radionuclides (Pu, Np, U) and {sup 137}Cs in soils collected from the Idaho National Engineering and Environmental Laboratory and other sites in Idaho, Montana, and Wyoming

    Beasley, T.M.; Rivera, W. Jr. [Dept. of Energy, New York, NY (United States). Environmental Measurements Lab.; Kelley, J.M.; Bond, L.A. [Pacific Northwest National Lab., Richland, WA (United States); Liszewski, M.J. [Bureau of Reclamation (United States); Orlandini, K.A. [Argonne National Lab., IL (United States)

    1998-10-01

    The isotopic composition of Pu in soils on and near the Idaho National Engineering and Environmental Laboratory (INEEL) has been determined in order to apportion the sources of the Pu into those derived from stratospheric fallout, regional fallout from the Nevada Test Site (NTS), and facilities on the INEEL site. Soils collected offsite in Idaho, Montana, and Wyoming were collected to further characterize NTS fallout in the region. In addition, measurements of {sup 237}Np and {sup 137}Cs were used to further identify the source of the Pu from airborne emissions at the Idaho Chemical Processing Plant (ICPP) or fugitive releases from the Subsurface Disposal Area (SDA) in the Radioactive Waste Management Complex (RWMC). There is convincing evidence from this study that {sup 241}Am, in excess of that expected from weapons-grade Pu, constituted a part of the buried waste at the SDA that has subsequently been released to the environment. Measurements of {sup 236}U in waters from the Snake River Plain aquifer and a soil core near the ICPP suggest that this radionuclide may be a unique interrogator of airborne releases from the ICPP. Neptunium-237 and {sup 238}Pu activities in INEEL soils suggest that airborne releases of Pu from the ICPP, over its operating history, may have recently been overestimated.

  4. Measured solubilities and speciations of neptunium, plutonium, and americium in a typical groundwater (J-13) from the Yucca Mountain region

    Solubility and speciation data are important in understanding aqueous radionuclide transport through the geosphere. They define the source term for transport retardation processes such as sorption and colloid formation. Solubility and speciation data are useful in verifying the validity of geochemical codes that are part of predictive transport models. Results are presented from solubility and speciation experiments of 237NpO2+, 239Pu4+, 241Am3+/Nd3+, and 243Am3+ in J-13 groundwater (from the Yucca Mountain region, Nevada, which is being investigated as a potential high-level nuclear waste disposal site) at three different temperatures (25 degree, 60 degree, and 90 degree C) and pH values (5.9, 7.0, and 8.5). The solubility-controlling steady-state solids were identified and the speciation and/or oxidation states present in the supernatant solutions were determined. The neptunium solubility decreased with increasing temperature and pH. Plutonium concentrations decreased with increasing temperature and showed no trend with pH. The americium solutions showed no clear solubility trend with increasing temperature and increasing pH

  5. LOS ALAMOS NEUTRON SCIENCE CENTER CONTRIBUTIONS TO THE DEVELOPMENT OF FUTURE POWER REACTORS

    GAVRON, VICTOR I. [Los Alamos National Laboratory; HILL, TONY S. [Los Alamos National Laboratory; PITCHER, ERIC J. [Los Alamos National Laboratory; TOVESSON, FREDERIK K. [Los Alamos National Laboratory

    2007-01-09

    The Los Alamos Neutron Science Center (LANSCE) is a large spallation neutron complex centered around an 800 MeV high-currently proton accelerator. Existing facilities include a highly-moderated neutron facility (Lujan Center) where neutrons between thermal and keV energies are produced, and the Weapons Neutron Research Center (WNR), where a bare spallation target produces neutrons between 0.1 and several hundred MeV.The LANSCE facility offers a unique capability to provide high precision nuclear data over a large energy region, including that for fast reactor systems. In an ongoing experimental program the fission and capture cross sections are being measured for a number of minor actinides relevant for Generation-IV reactors and transmutation technology. Fission experiments makes use of both the highly moderated spallation neutron spectrum at the Lujan Center, and the unmoderated high energy spectrum at WNR. By combininb measurements at these two facilities the differential fission cross section is measured relative to the {sup 235}U(n,f) standard from subthermal energies up to about 200 MeV. An elaborate data acquisition system is designed to deal with all the different types of background present when spanning 10 energy decades. The first isotope to be measured was {sup 237}Np, and the results were used to improve the current ENDF/B-VII evaluation. Partial results have also been obtained for {sup 240}Pu and {sup 242}Pu, and the final results are expected shortly. Capture cross sections are measured at LANSCE using the Detector for Advanced Neutron Capture Experiments (DANCE). This unique instrument is highly efficient in detecting radiative capture events, and can thus handle radioactive samples of half-lives as low as 100 years. A number of capture cross sections important to fast reaction applications have been measured with DANCE. The first measurement was on {sup 237}Np(n,{gamma}), and the results have been submitted for publication. Other capture

  6. Sources and transport of anthropogenic radionuclides in the Ob River system, Siberia

    Cochran, J. Kirk; Moran, S. Bradley; Fisher, Nicholas S.; Beasley, Thomas M.; Kelley, James M.

    2000-06-01

    The potential sources of anthropogenic radionuclides to the Ob River system of western Siberia include global stratospheric fallout, tropospheric fallout from atomic weapons tests and releases from production and reprocessing facilities. Samples of water, suspended and bottom sediments collected in 1994 and 1995 have been used to characterize the sources and transport of 137Cs, Pu isotopes, 237Np and 129I through the system. For the radionuclides that associate with particles, isotope ratios provide clues to their sources, providing any geochemical fractionation can be taken into account. Activity ratios of 239,240Pu/ 137Cs in suspended sediments are lower than the global fallout ratio in the Irtysh River before its confluence with the Ob, comparable to fallout in the central reach of the Ob, and greater than the fallout values in the lower Ob and in the Taz River. This pattern mirrors the downriver decrease in dissolved organic carbon (DOC) concentrations. Laboratory adsorption experiments with Ob River sediment and water show that Kd values for Am (and presumably other actinides) are depressed by two orders of magnitude in the presence of Ob DOC concentrations, relative to values measured in DOC-free Ob water. Iodine and cesium Kd values show little or no (less than a factor of 2) dependence on DOC. Mixing plots using plutonium isotope ratios (atom ratios) show that Pu in suspended sediments of the Ob is a mixture of stratospheric global fallout at northern latitudes, tropospheric fallout from the former Soviet Union test site at Semipalatinsk and reprocessing of spent fuel at Tomsk-7. Plutonium from Semipalatinsk is evident in the Irtysh River above its confluence with the Tobal. Suspended sediment samples taken in the Ob above its confluence with the Irtysh indicate the presence of Pu derived from the Tomsk-7 reprocessing facilities. A mixing plot constructed using 237Np/ 239Pu vs. 240Pu/ 239Pu shows similar mixtures of stratospheric and tropospheric fallout

  7. Determination of actinide elements in environmental samples by ICP-MS

    Methods for the determination of the actinide elements in water, biological, soil and sediment samples have been developed using on-line solid phase extraction and high performance liquid chromatography (HPLC) coupled with inductively coupled plasma mass spectrometry (ICP-MS). Initial applications utilised a commercially available resin, namely TRU-Spec resin, for efficient removal of the matrix prior to elution of uranium and thorium analytes. Comparative analyses of reference materials and natural water samples from Plymouth and Dartmoor demonstrated significant improvement in precision and speed of analysis by using TRU-Spec coupled to ICP-MS compared with alpha spectrometry. Further applications of the TRU-Spec resin for the determination of the transuranic actinide elements neptunium, plutonium and americium, resulted in the successful determination of 239Pu and 217Np in biological reference materials. Detection limits were 700, 850, and 600 attograms (ag) for 237Np, 233Pu, and 241Am, respectively, for a 0.5 mI sample injection, and better than 200 ag g-1 with 50 ml pre-concentration when sector field (SF) ICP-MS was used. A method for the selective sequential elution of uranium and plutonium was also developed to facilitate the determination of 239Pu without interference due to the 238U1H+ polyatomic ion, caused by high concentrations of 238U in sediment samples. Investigations were performed into the use of a polymeric substrate, which was dynamically coated with chelating dyes such as xylenol orange and 4-(2-pyridylazo) resorcinol, and a silica substrate coated with permanently bonded iminodiacetic acid. The latter was used for the successful determination of uranium and thorium in certified reference material waters. However, the column was found to have a high affinity for iron, making it unsuitable for the determination of the actinides in soil and sediment samples. Subsequently, a polystyrene substrate which was dynamically coated with dipicolinic acid

  8. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  9. Neutron reaction cross section data for advanced nuclear applications

    Full text of publication follows: Worldwide major research efforts are currently being carried out in order to develop a new concept of nuclear power generation, so-called accelerator driven systems (ADS) for energy production and transmutation of radioactive nuclear waste. A suggested approach is the energy amplifier (EA), which is a sub-critical reactor using a powerful proton accelerator and a spallation reaction as neutron source. Since the EA is based on the thorium-uranium fuel cycle, where the natural resources of the main fuel thorium are estimated to last for hundred thousands of years, it is considered to provide clean and almost inexhaustible nuclear energy. Apart from necessary new technical developments, the realization of these concepts depends strongly on the availability of accurate nuclear reaction data. In particular, precise knowledge about cross sections for fission, neutron capture and scattering is required for the nuclides involved in the Th-U fuel cycle. Among the first priority isotopes the IAEA had pointed out 231Pa and 233Pa. The latter one, 233Pa, is of specific interest, since it plays an important role as an intermediate nucleus in the formation of the fissile 233U from the fertile 232Th. With its half life of 27.0 days for β-decay, 233Pa is not a 'long-lived' nucleus, but it still requires careful attention in the design and operation of thorium-fueled reactors. When a thorium-fueled reactor is stopped, the present amount of 233Pa will continue to decay into 233U, leading to an increase in reactivity, which may even cause criticality. This mechanism is known as 'protactinium effect' and is proportional to the power level of the reactor. Also the precise knowledge of the fission cross section of 231Pa (above 1 b for fast neutrons) is essential for simulations of the balance of nuclei in and, thus, the reactivity behavior of the reactor. We present recent cross section data from direct, energy resolved measurements of the neutron

  10. Deep circulation changes in the South Atlantic since the Last Glacial Maximum from Nd isotope and multi-proxy records

    Wei, R.; Abouchami, W.; Zahn, R.; Masque, P.

    2016-01-01

    We report down-core sedimentary Nd isotope (εNd) records from two South Atlantic sediment cores, MD02-2594 and GeoB3603-2, located on the western South African continental margin. The core sites are positioned downstream of the present-day flow path of North Atlantic Deep Water (NADW) and close to the Southern Ocean, which makes them suitable for reconstructing past variability in NADW circulation over the last glacial cycle. The Fe-Mn leachates εNd records show a coherent decreasing trend from glacial radiogenic values towards less radiogenic values during the Holocene. This trend is confirmed by εNd in fish debris and mixed planktonic foraminifera, albeit with an offset during the Holocene to lower values relative to the leachates, matching the present-day composition of NADW in the Cape Basin. We interpret the εNd changes as reflecting the glacial shoaling of Southern Ocean waters to shallower depths combined with the admixing of southward flowing Northern Component Water (NCW). A compilation of Atlantic εNd records reveals increasing radiogenic isotope signatures towards the south and with increasing depth. This signal is most prominent during the Last Glacial Maximum (LGM) and of similar amplitude across the Atlantic basin, suggesting continuous deep water production in the North Atlantic and export to the South Atlantic and the Southern Ocean. The amplitude of the εNd change from the LGM to Holocene is largest in the southernmost cores, implying a greater sensitivity to the deglacial strengthening of NADW at these sites. This signal impacted most prominently the South Atlantic deep and bottom water layers that were particularly deprived of NCW during the LGM. The εNd variations correlate with changes in 231Pa/230Th ratios and benthic δ13C across the deglacial transition. Together with the contrasting 231Pa/230Th: εNd pattern of the North and South Atlantic, this indicates a progressive reorganization of the AMOC to full strength during the Holocene.

  11. Partitioning of the rare earths and actinides between R7T7 nuclear glass alteration products and solution according to disposal conditions; Partage des terres rares et des actinides entre solution et produits d`alteration du verre nucleaire type R7T7 en fonction des conditions de stockage

    Menard, O.

    1995-10-25

    The alteration of nuclear glass by water is liable to release radionuclides into the environment. Determining the release kinetics of these elements and their aqueous chemical forms are therefore essential steps in establishing the safety of a geological repository site. Leach tests were conducted with a nonradioactive specimen of the French ``R7T7`` light water containment glass spiked with U and Th, and with two R7T7 specimens spiked with {sup 237}Np and {sup 239}Pu, respectively. The alteration solution compositions were representative of deep groundwater and contained carbonate, sulfate, phosphate, fluorine and chlorine ions. The release of U, Th, Np and Pu, as well as of the rare earths La, Ce and Nd were monitored by ICP mass spectrometry and by {alpha} spectrometry. Scanning and transmission electron microscopic examination of the nonradioactive altered glass surfaces was also performed to assess the partitioning balance for the rare earths, U and Th between the glass alteration products and solution. The mobility of these elements depends on two competing mechanisms. The rare earths and thorium are incorporated in the alteration products (gel); the retention process is assumed to involve chemisorption or coprecipitation, enhanced in the gel layer by the presence of phosphate ions in particular. Conversely, the aqueous species in the alteration solutions (mainly anions) form complexes with the actinides and rare earths; this phenomenon is particularly evident with U and Np. The presence of carbonate ions favors this mobility. Plutonium differs from U and Np in that it is adsorbed mainly on colloids formed by glass dissolution, the principal factors governing its chemical evolution in solution. (author). refs., 122 figs., 185 tabs.

  12. Basalt-radionuclide distribution coefficient determinations. FY-1979 annual report

    Experimental radionuclide distribution coefficients (Kd') were determined for Pomona, Flow E, Umtanum basalts, and secondary mineralization associated with Pomona basalt at 230, 600 and 1500C. Radionuclides used were 75Se, 85Sr, 99Tc, 125I, 135Cs, 226Ra, 237Np, 238U, 241Am, and 241Pu. Solution oxygen contents were controlled by the basalt/groundwater system (Eh = 600 to 700 mV), and were high (8.2 to 8.4 mg/l) at 230C. Oxygen contents and pH changed little in contact with basalt. The effects of temperature changes on radionuclide Kd' results varied depending upon the radionuclide involved, solution-solid reactions, and the relationship of the radionuclide to these reactions. For example, cesium Kd' values decreased from 3100 ml/g for Umtanum basalt at 230C to 120 ml/g at 1500C. At the same time, strontium Kd' values increased for Umtanum basalt from 105 ml/g at 230C to complete removal at 1500C and 40 days. Radionuclide adsorption coefficient measurements at higher temperatures and pressures were made in addition to the 230C, solution-solid contact time-conditional Kd (Kd') measurements. These include Kd' measurements with Umtanum basalt, Pomona basalt, Flow E basalt and secondary mineralization and radioisotopes of americium, cesium, iodine, neptunium, plutonium, radium, selenium, strontium, technetium and uranium. The additional temperatures involved were 600C, 1500C, and 3000C. At 1500C, argon pressures of 6.9, 13.8, 20.7, and 27.6 MPa will be used to ascertain the effects of pressure changes on Kd' values. So far only the 6.9 MPa argon pressure has been investigated. The upper temperature of 2500C is where thermal breakdown of dioctahedral smectites (secondary mineralization) begins

  13. Radionuclide reactions with groundwater and basalts from Columbia River basalt formations

    Chemical reactions of radionuclides with geologic materials found in Columbia River basalt formations were studied. The objective was to determine the ability of these formations to retard radionuclide migration from a radioactive waste repository located in deep basalt. Reactions that can influence migration are precipitation, ion-exchange, complexation, and oxidation-reduction. These reactions were studied by measuring the effects of groundwater composition and redox potential (Eh) on radionuclide sorption on fresh basalt surfaces, a naturally altered basalt, and a sample of secondary minerals associated with a Columbia River basalt flow. In addition, radionuclide sorption isotherms were measured for these materials and reaction kinetics were determined. The radionuclides studied were 137Cs, 85Sr, 75Se, /sup 95m/Tc, 237Np, 241Am, 226Ra and 237Pu. The Freundlich equation accurately describes the isotherms when precipitation of radionuclides does not occur. In general, sorption increased in the order: basalt < altered basalt < secondary minerals. This increase in sorption corresponds to increasing surface area and cation exchange capacity. The Eh of the system had a large effect on technetium, plutonium, and neptunium sorption. Technetium(VII), Pu(VI), and Np(V) are reduced to Tc(IV), Pu(IV), and Np(IV), respectively, under Eh conditions expected in deep basalt formations. The kinetics of radionuclide sorption and basalt-groundwater reactions were observed over a period of 18 weeks. Most sorption reactions stabilized after about four weeks. Groundwater composition changed the least in contact with altered basalt. Contact with secondary minerals greatly increased Ca, K, and Mg concentrations in the groundwater

  14. Effects of physico-chemical properties of actinide oxides on tumour induction after inhalation exposure in rats

    Fritsch, P.; Dudoignon, N.; Ramounet, B.; Guezingar-Liebard, F.; Matton, S.; Lizon, C.; Massiot, P. [CEA, DSV/DRR/SRCA/LRT, Bruyeres le Chatel (France). Laboratoire de Radiotoxicologie

    2000-07-01

    This review described results obtained with new methods in authors' laboratory to measure dissolution parameters of inhaled actinide oxides and the distribution of {alpha}-delivered dose within lungs in relation to their tumor induction in rats. The oxides were industrial PuO{sub 2} (>50% of {alpha}, due to {sup 238}Pu), {sup 237}NpO{sub 2} and 2 different (U, Pu)O{sub 2} containing about 5% industrial Pu. The aerodynamic median activity diameter of their aerosols with authors' specific device measured with a cascade impactor were similar to each other (1.7-3.6 {mu}m {sigma}g). Their chemical composition was characterized using scanning transmission electron microscopy by energy dispersive X-ray spectrometry on alveolar macrophages which had phagocytosed them. Measurements of dissolution parameters after inhalation exposure in the rat and after in vitro incubation respectively revealed that the f{sub r} values were in the range of 2 x 10{sup -2}-1 x 10{sup -4}, indicating oxides behaved as a type S compound, and that the values were quite different from those above, suggesting S{sub s} should be considered as a variable for dose calculation, depending on time after inhalation. An autoradiographic method using solid tract detector and lung frozen sections revealed that aggeregations involving interstitial macrophages were often associated with fibrosis and/or preneoplastic lesions, which was explainable of a threshold in the dose-effect relationship for lung cancer occurrence. Results showed that, on inhalation of the oxides, risk assessment for lung tumor induction at low doses can not be extrapolated from that at high doses. (K.H.)

  15. Crystallography and magnetic properties of transuranium element oxygen compounds (Np, Pu and Am)

    This paper includes: 1) The brief description of the experimental techniques used for analyzing very small quantities of solid radioactive compounds (differential thermal micro-analyses, diffraction of X rays, magnetic susceptibility and Moessbauer resonance). 2) The methods of synthesis of the ternary oxides of transuranic elements at oxidation degrees III (Pu2MoO6, Pu2WO6, Pu2(WO4)3, Am2MoO6, Am2WO6, Am2(MoO4)3 and Am2(WO4)3) and at degree IV (Np(VO3)4, Np(MoO4)2, Np(WO4)2 and Pu(MoO4)2). The drawing up of liquid-solid balance diagrams enabled the field of stability of molybdate (or tungstate) systems of alkaline transuranic - mobybdates (or tungstates) to be clarified. 3) The study of the structural properties of the identified phases. These results taken as a whole made it possible to establish a ''comparative crystal - chemistry'' of the oxigenated phases of Np, Pu and Am with those of the thorium and uranium actinide elements and with the rare earths of adjacent ionic radius. 4) The Moessbauer resonance study of 237Np in the solid solution Usub(1-x)Npsub(x)O2 (0 < x <= 1). The analysis of the results obtained is the subject of Part IV of the manuscript. 5) The effects of alpha self-irradiation mentioned in the last part show that it is necessary to gather fresh information on the condensed phases containing heavy radioactive atoms

  16. Enumeration of microbial populations in radioactive environments by epifluorescence microscopy

    Epifluorescence microscopy was utilized to enumerate halophilic bacterial populations in two studies involving inoculated, actual waste/brine mixtures and pure brine solutions. The studies include an initial set of experiments designed to elucidate potential transformations of actinide-containing wastes under salt-repository conditions, including microbially mediated changes. The first study included periodic enumeration of bacterial populations of a mixed inoculum initially added to a collection of test containers. The contents of the test containers are the different types of actual radioactive waste that could potentially be stored in nuclear waste repositories in a salt environment. The transuranic waste was generated from materials used in actinide laboratory research. The results show that cell numbers decreased with time. Sorption of the bacteria to solid surfaces in the test system is discussed as a possible mechanism for the decrease in cell numbers. The second study was designed to determine radiological and/or chemical effects of 239Pu, 243Am, 237Np, 232Th and 238U on the growth of pure and mixed anaerobic, denitrifying bacterial cultures in brine media. Pu, Am, and Np isotopes at concentrations of ≤1x10-6 M , ≤5x10-6 M and ≤5x10-4M respectively, and Th and U isotopes ≤4x10-3M were tested in these media. The results indicate that high concentrations of certain actinides affected both the bacterial growth rate and morphology. However, relatively minor effects from Am were observed at all tested concentrations with the pure culture

  17. Fabrication and characterization of MCC approved testing material: ATM-9 glass

    The Materials Characterization Center ATM-9 glass is designed to be representative of glass to be produced by the Defense Waste Processing Facility at the Savannah River Plant, Aiken, South Carolina. ATM-9 glass contains all of the major components of the DWPF glass and corresponds to a waste loading of 29 wt %. The feedstock material for this glass was supplied by Savannah River Laboratory, Aiken, SC, as SRL-165 Black Frit to which was added Ba, Cs, Md, Nd, Zr, as well as 99Tc, depleted U, 237Np, 239+240Pu, and 243Am. The glass was produced under reducing conditions by the addition of 0.7 wt % graphite during the final melting process. Three kilograms of the glass were produced from April to May of 1984. On final melting, the glass was formed into stress-annealed rectangular bars of two sizes: 1.9 x 1.9 x 10 cm and 1.3 x 1.3 x 10 cm. Seventeen bars of each size were made. The analyzed composition of ATM-9 glass is listed. Examination by optical microscopy of a single transverse section from one bar showed random porosity estimated at 0.36 vol % with nominal pore diameters ranging from approx. 5 μm to 200 μm. Only one distinct second phase was observed and it was at a low concentraction level in the glass matrix. The phase appeared as spherical metallic particles. X-ray diffraction analysis of this same sample did not show any diffraction peaks from crystalline components, indicating that the glass contained less than 5 wt % of crystalline devitrification products. The even shading on the radiograph exposure indicated a generally uniform distribution of radioactivity throughout the glass matrix, with no distinct high-concentration regions

  18. Neutron cross section standards and instrumentation

    1992-09-01

    This report from the National Institute of Standards and Technology contains a summary of the accomplishments of the Neutron Cross Section Standards and Instrumentation Project during the second year of a three-year interagency agreement. This program includes a broad range of data measurements and evaluations. An emphasis has been focused on the (sup 10)B cross sections where serious discrepancies in the nuclear data base remain. In particular, there are important problems with the interpretation of the helium gas production associated with diagnostic measurements of interest in nuclear technology. The enhanced use of this isotope for medical treatment is also of significance. New measurements of neutron reaction cross sections for (sup 10)B are in progress in collaboration with scientists at the Oak Ridge National Laboratory. New experiments are in progress on the important dosimetry standards (sup 237)Np(n,f) and (sup 239)Pu(n,f) below 1 MeV neutron energy. In addition, new measurements of charged-particle production in basic biological elements for medical applications are underway. Further measurements are planned or in progress in collaborations which include fission fragment energy and angular distributions, and neutron energy spectra and angular distributions from neutron-induced fission. Also measurements of angular distributions of neutrons from scattering on protons, and determinations of capture cross section of gold are planned for a later time. Data evaluation will shift to include a unified international effort to motivate new measurements and evaluations. In response to the requests of the measurement community, NIST is beginning the formation of a national depository for fissionable isotope mass standards. This action will preserve for future measurements the valuable and irreplaceable critical samples whose masses and composition have been carefully determined and documented over the past 30 years of the nuclear program.

  19. Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

    P. D. Wheatley (INEEL POC); R. P. Rechard (SNL)

    1998-09-01

    The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the U.S. Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF). The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW). A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP). A performance assessment (PA) was chosen as the method of analysis. The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr. Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis. Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria. This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (1.5%) and mass (3.8%) of waste in the potential repository. Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF.

  20. Neutron-induced fission cross section of 240Pu from 0.5 MeV to 3 MeV

    Salvador-Castiñeira, P.; Bryś, T.; Eykens, R.; Hambsch, F.-J.; Göök, A.; Moens, A.; Oberstedt, S.; Sibbens, G.; Vanleeuw, D.; Vidali, M.; Pretel, C.

    2015-07-01

    240Pu has recently been pointed out by a sensitivity study of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) to be one of the isotopes whose fission cross section lacks accuracy to meet the upcoming needs for the future generation of nuclear power plants (GEN-IV). In the High Priority Request List (HPRL) of the OECD, it is suggested that the knowledge of the 240Pu(n ,f ) cross section should be improved to an accuracy within 1-3 %, compared to the present 5%. A measurement of the 240Pu cross section has been performed at the Van de Graaff accelerator of the Joint Research Center (JRC) Institute for Reference Materials and Measurements (IRMM) using quasi-monoenergetic neutrons in the energy range from 0.5 MeV to 3 MeV. A twin Frisch-grid ionization chamber (TFGIC) has been used in a back-to-back configuration as fission fragment detector. The 240Pu(n ,f ) cross section has been normalized to three different isotopes: 237Np(n ,f ) , 235U (n ,f ) , and 238U (n ,f ) . Additionally, the secondary standard reactions were benchmarked through measurements against the primary standard reaction 235U (n ,f ) in the same geometry. A comprehensive study of the corrections applied to the data and the associated uncertainties is given. The results obtained are in agreement with previous experimental data at the threshold region. For neutron energies higher than 1 MeV, the results of this experiment are slightly lower than the ENDF/B-VII.1 evaluation, but in agreement with the experiments of Laptev et al. (2004) as well as Staples and Morley (1998).

  1. In situ and laboratory migration experiments through boom clay at Mol

    Physico-chemical characterization and migration studies in the Boom clay, envisaged as a potential host sediment for high level waste disposal in Belgium, were started some 15 years ago. A synthesis study of this experimental work has recently been conducted to compile all available data. From a comparison of the available migration data and the data requirements as derived from the performance assessment studies PAGIS (1988) and PACOMA (1991) the new migration programme (1991-1995) was defined. The critical radionuclides, both with relation to dose rates to man and to missing or unreliable migration data, turned out to be 14 C, 99 Tc. 135 Cs and 237 Np. A second group of radionuclides was found to be possibly critical: 79 Se, 93 Zr, 107 Pd, U-, Am-, Cm-, and Pu-isotopes. This report concentrates on the experimental results as obtained from the migration experiments started in the previous migration programme. Some of the reported radionuclides e.g. 90 Sr) have lost their critical character and will not be further studied within the new programme. New experimental data from laboratory tests have become available for Np, Cs, Sr and C (as HC03-) and the first results on the migration of organic molecules dissolved in the interstitial Boom clay water are reported. The hydraulic parameters (the hydraulic conductivity K and the storage coefficient So) were calculated from both laboratory percolation experiments and in situ piezometric measurements. Conclusions concerning Boom clay anisotropy are drawn. Finally, a short description of the ongoing in situ HTO injection experiment is given and the experimental data are analyzed and discussed. 10 refs., 4 figs., 1 tab

  2. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

    Arzu Alpan, F.; Kulesza, Joel A.

    2016-02-01

    This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237Np(n,f) in the cavity capsule using BUGLE-B7.

  3. Water Sorption and Radiolysis Studies for Neptunium Oxides

    Icenhour, A.S.

    2004-02-03

    Plans are to convert the {sup 237}Np that is currently stored as a nitrate solution at the Savannah River Site to NpO{sub 2} and then ship it to the Y-12 National Security Complex in Oak Ridge for interim storage. This material will serve as feedstock for the {sup 238}Pu production program, and some will be periodically shipped to the Oak Ridge National Laboratory (ORNL) for fabrication into targets. The safe storage of this material requires an understanding of the radiolysis of moisture that is sorbed on the oxides, which, in turn, provides a basis for storage criteria (namely, moisture content). A two-component experimental program has been undertaken at ORNL to evaluate the radiolytic effects on NpO{sub 2}: (1) moisture uptake experiments and (2) radiolysis experiments using both gamma and alpha radiation. These experiments have produced two key results. First, the water uptake experiments demonstrated that the 0.5 wt % moisture limit that has been typically established for similar materials (e.g., uranium and plutonium oxides) cannot be obtained in a practical environment. In fact, the uptake in a typical environment can be expected to be at least an order of magnitude lower than the limit. The second key result is the establishment of steady-state pressure plateaus as a result of the radiolysis of sorbed moisture. These plateaus are the result of back reactions that limit the overall pressure increase and H{sub 2} production. These results clearly demonstrate that 0.5 wt % H{sub 2}O on NpO{sub 2} is safe for long-term storage--if such a moisture content could even be practically reached.

  4. Large scale production of Pu-238 to 'denature' weapons-grade plutonium. Rev. 0

    The concept of open-quote denaturing close-quote weapons-grade or reactor-grade plutonium through the addition of Pu-238 has been proposed recently by Dr. Mel Coops of Lawrence Livermore National Laboratory. He indicates that on the order of 2-3 wt.% plutonium-238 (Pu-238) may be sufficient to render the plutonium impractical for weapons use due to excessive heat generation. This would significantly reduce or eliminate proliferation concerns for use of plutonium in commercial nuclear fuels. This report presents a high-level assessment of the potential for producing large quantities of Pu-238 for the purpose of open-quote denaturing close-quote the weapons-grade or reactor-grade plutonium that may be used in commercial reactors as mixed-oxide (MOX) fuels. Pu-238 production can be done by two methods. The traditional method employed in the U.S. was to irradiate targets of neptunium-237 (Np-237), this isotope being produced from neutron capture in U-236. Recycling enriched uranium fuels in the defense production reactor fuel cycle increased the U-236 content, increasing the subsequent yield of Np-237. Production by this method thus involves significant reactor irradiation time, and reprocessing of fuel to recover uranium and Np-237 for fabrication into new fuel and targets, respectively. The second method would be to irradiate targets of Am-241, recovered as a decay product from aged plutonium. This method again would require processing of plutonium or spent fuel to recover Am-241. Either method would require processing of targets after irradiation, and a facility for fabricating targets and processing the plutonium to a final pure oxide form

  5. OSMOSE program : statistical review of oscillation measurements in the MINERVE reactor R1-UO2 configuration.

    Stoven, G.; Klann, R.; Zhong, Z.; Nuclear Engineering Division

    2007-08-28

    The OSMOSE program is a collaboration on reactor physics experiments between the United States Department of Energy and the France Commissariat Energie Atomique. At the working level, it is a collaborative effort between the Argonne National Laboratory and the CEA Cadarache Research Center. The objective of this program is to measure very accurate integral reaction rates in representative spectra for the actinides important to future nuclear system designs, and to provide the experimental data for improving the basic nuclear data files. The main outcome of the OSMOSE measurement program will be an experimental database of reactivity-worth measurements in different neutron spectra for the heavy nuclides. This database can then be used as a benchmark to verify and validate reactor analysis codes. The OSMOSE program (Oscillation in Minerve of isotopes in Eupraxic Spectra) aims at improving neutronic predictions of advanced nuclear fuels through oscillation measurements in the MINERVE facility on samples containing the following separated actinides: {sup 232}Th, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm, and {sup 245}Cm. The first part of this report provides an overview of the experimental protocol and the typical processing of a series of experimental results which is currently performed at CEA-Cadarache. In the second part of the report, improvements to this technique are presented, as well as the program that was created to process oscillation measurement results from the MINERVE facility in the future.

  6. Comparative Measures of Radionuclide Containment in the Crystalline Geophere

    A probabilistic model for assessing the capacity of a fractured crystalline rock volume to contain radionuclides is developed. The rock volume is viewed as a network of discrete fractures through which radionuclides are transported by flowing water. Diffusive mass transfer between the open fractures and the stagnant water in the pore space of the rock matrix allow radionuclides access to mineral grains where physical and chemical processes - collectively known as sorption - can retain radionuclides. A stochastic Lagrangian framework is adopted to compute the probability that a radionuclide particle will be retained by the rock, i.e., the probability that it will decay before being released from the rock volume. A dimensionless quantity referred to as the 'containment index' is related to this probability and proposed as a suitable measure for comparing different rock volumes; such a comparative measure may be needed, for example, in a site selection program for geological radioactive waste disposal. The probabilistic solution of the transport problem is based on the statistics of two Lagrangian variables: τ, the travel time of an imaginary tracer moving with the flowing water, and β, a suitably normalized surface area available for retention. Statistics of τ and β may be computed numerically using site-specific discrete fracture network simulations. Fracture data from the well-characterized Aespoe Hard Rock Laboratory site in southern Sweden are used to illustrate the implementation of the proposed containment index for six radionuclides (126Sn, 129I, 135Cs, 237Np, 239Pu, and 79Se). It is found that fractures of small aperture imply prolonged travel times and hence long tails in both beta and tau. This, in turn, enhances retention and is favorable from a safety assessment perspective

  7. Working Party on International Nuclear Data Evaluation Co-operation (WPEC). Presentations and documents submitted to the 27. meeting, NEA Headquarters, 21-22 May 2015

    The NEA's nuclear data evaluation co-operation activities involve the following evaluation projects: ENDF (United States), JENDL (Japan), ROSFOND/BROND (Russia), JEFF (other Data Bank member countries) and CENDL (China) in close co-operation with the Nuclear Data Section of the International Atomic Energy Agency (IAEA). The working party was established to promote the exchange of information on nuclear data evaluations, measurements, nuclear model calculations, validation, and related topics, and to provide a framework for co-operative activities between the participating projects. The working party assesses nuclear data improvement needs and addresses these needs by initiating joint evaluation and/or measurement efforts. The 27. Meeting of the WPEC was the occasion to review the experimental activities, the evaluation projects and the Status of subgroups. This document brings together the available documents of the meeting: 1 - The Proposed agenda, the list of participants and the Summary record of the previous meeting, May 2014 (Report NEA-SEN-NSC-WPEC--2014-2); 2 - The Reports on experimental activities: Europe (NEA DB), Japan, USA, Russia, China; 3 - Some Brief progress reports from the evaluation projects: ENDF, JEFF, JENDL, BROND/ROSFOND, CENDL, IAEA, and TENDL; 4 - The Status of some subgroups: Subgroup 35 (Scattering angular distribution in the fast energy range); Subgroup 36 (Reporting and usage of experimental data for evaluation in the resolved resonance region); Subgroup 37 (Improved fission product yield evaluation methodologies); Subgroup 38 (A modern nuclear database structure beyond the ENDF format); Subgroup 39 (Methods and approaches to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data files); Subgroup 40 (Collaborative International Evaluated Library Organisation (CIELO) Pilot Project); Subgroup 41 (Improving nuclear data accuracy of 241Am and 237Np capture cross-sections); Subgroup C (High Priority

  8. Conceptual study of the future nuclear fuel cycle system for the extended LWR age

    A large scale integrated fuel cycle facility (IFCF) is assumed for the future nuclear fuel cycle in the extended LWR age. Spent MOX fuels are reprocessed mixed with UOX in a centralized reprocessing plant. The reprocessing plant separates long-lived nuclides as well as Pu. Nitric acid solutions of those products are fed directly to MOX fabrication process which is incorporated with reprocessing. MOX pellets are made by sphere-cal process. Two process concepts are made as advanced reprocessing incorporated with partitioning (ARP) which has the function of long-lived nuclides recovery. One is a simplified Purex combined with partitioning. Extractable long-lived nuclides, 237Np and 99Tc, are assumed to be recovered in main flow stream of the improved Purex process. The other process concept is made aiming at recovering all TRU nuclides in reprocessing to meet with TRU recycle requirement in the long future. A concept of the future fuel cycle system is made by combining integrated fuel cycle facility and very high burnup LWRs (VHBR). The reactor concept of VHBRs has been proposed to improve Pu recycle economy in the future. Highly enriched MOX fuel are loaded in the full core of reactor in order to increase reactivity for the burnup. Fuel cycle indices such as Pu isotopic composition change, spent fuel integration, nuclide transmutation effect are estimated by simulating the Pu recycling in the system of VHBR and ARP. It is concluded that Pu enrichment of MOX fuel can be kept less than 20 % through multi-recycle. Reprocessing MOX fuels with UOX shows a favorable effect for keeping Pu reactivity high enough for VHBR. Integration of spent MOX fuel can be reduced by Pu recycle. Transmutation of Np is feasible by containing Np into MOX fuel. (author)

  9. Gut-related studies of radionuclide toxicity

    This project is concerned with the behavior of radioactive materials that mayu be ingested as a consequence of a reactor accident, unavoidable occupational exposure, or after release to the environment and incorporation into the food chain. Current emphasis is on evaluating hazards from ingested actinides as a function of animal age, species, nutrition, and diet, or chemicophysical state of the actinide. We are also concerned with the behavior of actinides that are inhaled and pass through the gastrointestinal (GI) tract after clearance from the lungs. Recent experiments showed that adult swine absorbed more 238Pu nitrate than had previously been indicated in studies with 239Pu nitrate, three times more than is absorbed by rats. Absorption of 238Pu by rats on a vitamin-D-deficient diet was about 10 times higher than absorption by rats on a balanced diet. Studies on the effect of chemical form on actinide absorption showed that the citrate forms of 238Pu, 241Am and 244Cm were transported in higher quantities than the nitrate forms across the intestine. Citrate had no effect on the 237Np transport, but the mass of isotope administered was found to be important. Absorption by neonates was inversely related to the mass of neptunium gavaged, in contrast to the effect of mass on neptunium absorption by adult rats. Organic binding of 238Pu in liver tissue, in situ, resulted in decreased absorption by adult or neonatal rats. These results demonstrate that animal age, species and nutritional state are important factors in determining GI absorption of actinide compounds. Chemical form and oxidation state also influence transport. These effects vary with animal age and with the actinide in question

  10. Working Party on International Nuclear Data Evaluation Co-operation (WPEC). Presentations and documents submitted to the 28. meeting, OECD Headquarters, Conference Centre, Paris, France, 9-13 May 2016

    The NEA's nuclear data evaluation co-operation activities involve the following evaluation projects: ENDF (United States), JENDL (Japan), ROSFOND/BROND (Russia), JEFF (other Data Bank member countries) and CENDL (China) in close co-operation with the Nuclear Data Section of the International Atomic Energy Agency (IAEA). The working party was established to promote the exchange of information on nuclear data evaluations, measurements, nuclear model calculations, validation, and related topics, and to provide a framework for co-operative activities between the participating projects. The working party assesses nuclear data improvement needs and addresses these needs by initiating joint evaluation and/or measurement efforts. The 28. Meeting of the WPEC was the occasion to review the experimental activities, the evaluation projects and the Status of subgroups. This document brings together the available documents of the meeting: 1 - The Reports on experimental activities: Europe (NEA DB), Japan, USA, China; 2 - Some Brief progress reports from the evaluation projects: ENDF, JEFF, JENDL, BROND/ROSFOND, CENDL, IAEA, TENDL; 3 - The Status of subgroups: Subgroup 37 (Improved fission product yield evaluation methodologies); Subgroup 38 (A modern nuclear database structure beyond the ENDF format); Subgroup 39 (Methods and approaches to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data files); Subgroup 40 (Collaborative International Evaluated Library Organisation (CIELO) Pilot Project); Subgroup 41 (Improving nuclear data accuracy of 241Am and 237Np capture cross-sections); Subgroup 42 (Thermal Scattering Kernel S(a,b): Measurement, Evaluation and Application); Subgroup C (High Priority Request List - HPRL); New Subgroups were proposed and presented: 43 - Code infrastructure to support a general nuclear database structure; WPEC long-term sub-group proposal: International standard for a general nuclear database structure

  11. Solar r-process-constrained actinide production in neutrino-driven winds of supernovae

    Goriely, S.; Janka, H.-Th.

    2016-04-01

    Long-lived radioactive nuclei play an important role as nucleo-cosmochronometers and as cosmic tracers of nucleosynthetic source activity. In particular nuclei in the actinide region like thorium, uranium, and plutonium can testify to the enrichment of an environment by the still enigmatic astrophysical sources that are responsible for the production of neutron-rich nuclei by the rapid neutron-capture process (r-process). Supernovae and merging neutron-star (NS) or NS-black hole binaries are considered as most likely sources of the r-nuclei. But arguments in favour of one or the other or both are indirect and make use of assumptions; they are based on theoretical models with remaining simplifications and shortcomings. An unambiguous observational determination of a production event is still missing. In order to facilitate searches in this direction, e.g. by looking for radioactive tracers in stellar envelopes, the interstellar medium or terrestrial reservoirs, we provide improved theoretical estimates and corresponding uncertainty ranges for the actinide production (232Th, 235, 236, 238U, 237Np, 244Pu, and 247Cm) in neutrino-driven winds of core-collapse supernovae. Since state-of-the-art supernova models do not yield r-process viable conditions -but still lack, for example, the effects of strong magnetic fields- we base our investigation on a simple analytical, Newtonian, adiabatic and steady-state wind model and consider the superposition of a large number of contributing components, whose nucleosynthesis-relevant parameters (mass weight, entropy, expansion time scale, and neutron excess) are constrained by the assumption that the integrated wind nucleosynthesis closely reproduces the solar system distribution of r-process elements. We also test the influence of uncertain nuclear physics.

  12. Evaluation of the alveolar macrophage role in the pulmonary distribution of actinide oxides

    Actinide oxide inhalation is potentially a risk during the fuel fabrication process in the electronuclear industry. These particles can induce pulmonary lesions. The alveolar macrophage play an important role in the particle sequestration and transport but the actinide toxicity towards these cells is not well known. The aim of this work was to characterize the evolution of particle localisation in lungs after inhalation and to evaluate the role of macrophages in the lesion histo-genesis. We have used of a solid track detector to visualise alpha dose distribution within lung tissue. After 237NpO2, MOX or PuO2 inhalation by rats, different kinetics of clearance were observed for the sub-pleural and peri-bronchial areas compared to the others alveolar areas. For initial lung burdens that alter the lung clearance, particle aggregates were observed. Their kinetic and localisation vary depending on the aerosol, for a same global dose delivered to the lungs. This could be due to the different specific alpha activities of the particles and to the particle number deposited in the lung to obtain a similar burden but it could be also due to a chemical toxicity of neptunium higher than that of the others actinides. The flow cytometry methods developed allow us to measure apoptosis, phagocytosis and free radicals generation. After addition of soluble uranium to the culture medium, similar results were obtained using either alveolar macrophages extracted from rats or a macrophage cell line. This work confirms that alveolar macrophages are involved in the aggregate formation which induces heterogeneous dose distribution within the different lung tissues. (author)

  13. Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity [O1] - Fundamentals of P3 Mechanism and Methodology Development for Plutonium Categorization

    'Protected Plutonium Production (P3)' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MAs). Doping the small amount of MAs such as 237Np or 241Am with large neutron capture cross-section into the uranium fuel to enhance the production of 238Pu or 242Pu, which have high spontaneous fission neutron source or also high decay heat to makes the process of the nuclear weapon manufacture and maintenance technologically difficult, can be effective for improving the isotopic barrier of proliferation resistance of the plutonium in thermal reactors. Super weapon grade plutonium could be produced in the blanket of a conventional FBR. However, by increasing the 238Pu or 242Pu ratio in the total plutonium by MAs doping into the fresh blanket, the protected plutonium with high proliferation-resistance can be bred. A new evaluation function, 'attractiveness', defined as a ratio of potential of fission yield to the technological difficulties of nuclear explosive device, has been proposed to evaluate the proliferation resistance of Pu based on the nuclear material property for Plutonium Categorization. In the conference, the fundamentals of P3 mechanism by transmutation of MA, and the comparison of the 'attractiveness' of the Pu produced in advanced reactors based on P3 mechanism and in the conventional reactors will be presented. Instead of the geological disposal or just their burning of MAs by the fission reaction, they should be treated as valuable fertile materials to enhance the proliferation resistance of plutonium produced in the thermal and fast breeder reactors for peace and sustainable prosperity in future. Acknowledgement: Some parts of this work have been supported by the Ministry of Education, Culture, Sports, Science and Technology in Japan. (authors)

  14. Protected Plutonium Production (P3) by transmutation of minor actinides for peace and sustainable prosperity

    'Protected Plutonium Production (P3)' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MAs). Doping the small amount of MAs such as 237Np or 241Am with large neutron capture cross-section into the uranium fuel to enhance the production of 238Pu or 242Pu, which have high spontaneous fission neutron source or also high decay heat to makes the process of the nuclear weapon manufacture and maintenance technologically difficult,can be effective for improving the isotopic barrier of proliferation resistance of the plutonium in thermal reactors. Super weapon grade plutonium could be produced in the blanket of a conventional FBR. However, by increasing the 238Pu or 242Pu ratio in the total plutonium by MAs doping into the fresh blanket, the protected plutonium with high proliferation-resistance can be bred. A new evaluation function, 'attractiveness', defined as a ratio of potential of fission yield to the technological difficulties of nuclear explosive device, has been proposed to evaluate the proliferation resistance of Pu based on the nuclear material property for Plutonium Categorization. The new evaluation function of attractiveness is applied for assessing the existing plutonium criteria as summarized in the following, (a) weapon grade plutonium (b) plutonium with 30% fraction of 240Pu (c) plutonium with 6% fraction of 238Pu (d) plutonium exempt from safeguards. Since both proliferation resistant plutonium compositions (b) and (c) give almost the same value of attractiveness, plutonium is categorized by following well accepted terminology, weapon grade, usable, practically unusable and exempt as shown. It is concluded based on the new evaluation function 'Attractiveness' that P3 mechanism by the transmutation of MA is very effective to improve the proliferation resistance of plutonium. In the conference, the fundamentals of P3 mechanism by transmutation of MA, and the comparison of the 'attractiveness' of

  15. Production of photofission fragments and study of their nuclear structure

    The fission fragments of heavy nuclei (Z > 90) are neutron-rich isotopes of the elements from Zn (Z = 30) to Nd (Z = 60) with a neutron number of 45 - 90. The large neutron excess in the fission fragments under study could lead to the essential change in their structure and radioactive decay characteristics. This change will manifest itself in the appearance of new magic numbers of protons or neutrons, new regions of deformation, of new islands of isomerism. The high energy of β-decay can result in the new, much rarer, modes of radioactive decay (for example, β2n or βα). Study of the nuclear structure of fission fragments is one of the main directions of the DRIBs project, being developed in the Flerov Laboratory of Nuclear Reactions, JINR. The aim of this project is the production of the intense beams of accelerated radioactive nuclei in a wide range of Z and A and the study of their properties. The neutron-rich nuclei of medium mass numbers will be produced in the photofission reactions on the electron accelerator microtron MT-25. The parameters of this microtron allow to produce the high yield of photofission fragments (up to 1011s-1 at the irradiation of thick uranium target by Bremsstrahlung. The first experiments using photofission fragments were performed. The independent yields of Kr (A = 87 - 93) and Xe (A = 137 - 143) fragments at the photofission of different heavy nuclei 232 Th, 238 U, 237 Np, 244 Pu were measured. These data allows to refer about yields of the most neutron-rich isotopes. The rare mode decay, emission of delayed neutron pair (β2n), was observed at the photofission of 238 U. It is, probably, 136 Sb, the intensity of β 2n branch is about 10-3. Other experiments using neutron, gamma and laser spectroscopy methods are planned. (authors)

  16. Annual progress report on nuclear data 1992

    Hansen, H.H. [ed.

    1993-06-01

    This is the 1992 annual report on nuclear data from the Central Bureau for Nuclear Measurements, Geel (Belgium). Work on standard neutron cross sections included {sup 235}U(n,f)/H(n,n) with Frisch gridded ionization chambers and using octacosanol samples. Mass, energy, and angular distribution of fission fragments for {sup 237}Np(n,f) from 0.5 to 5.5 MeV neutron energy. Alpha decay probabilities of {sup 239}Pu. In the area of nuclear data for fission technology, a measurement on the normalization of the {sup 239}Pu fission cross sections was performed. Parameters for 384 resonances in {sup 58}Ni and 350 resonances in {sup 60}Ni have been analyzed up to 1 MeV and 800 KeV, respectively. In the field of nuclear data for fusion technology, double differential neutron emission cross sections for {sup 9}Be(n,2n) for incident neutron energies between 0. 6 and 11.1 MeV have been reported. Extensive measurements of the neutron decay cross sections of {sup 207}Pb have been made. In the radionuclide metrology subproject contributions were made by the preparation of low energy x-ray standard sources, measurements of K- shell fluorescence yields, standardization of a {sup 152}Eu solution, evaluation of the second EUROMET intercomparison of {sup 192}Ir brachytherapy sources, and low level measurements on volcanic rock, archeological ceramics, soil and river sediments. Work was also reported in neutron metrology, major facilities upgrades, radiation physics, and support for a number of PhD projects.

  17. Neutron cross section standards and instrumentation

    This report from the National Institute of Standards and Technology contains a summary of the accomplishments of the Neutron Cross Section Standards and Instrumentation Project during the second year of a three-year interagency agreement. This program includes a broad range of data measurements and evaluations. An emphasis has been focused on the 10B cross sections where serious discrepancies in the nuclear data base remain. In particular, there are important problems with the interpretation of the helium gas production associated with diagnostic measurements of interest in nuclear technology. The enhanced use of this isotope for medical treatment is also of significance. New measurements of neutron reaction cross sections for 10B are in progress in collaboration with scientists at the Oak Ridge National Laboratory. New experiments are in progress on the important dosimetry standards 237Np(n,f) and 239Pu(n,f) below 1 MeV neutron energy. In addition, new measurements of charged-particle production in basic biological elements for medical applications are underway. Further measurements are planned or in progress in collaborations which include fission fragment energy and angular distributions, and neutron energy spectra and angular distributions from neutron-induced fission. Also measurements of angular distributions of neutrons from scattering on protons, and determinations of capture cross section of gold are planned for a later time. Data evaluation will shift to include a unified international effort to motivate new measurements and evaluations. In response to the requests of the measurement community, NIST is beginning the formation of a national depository for fissionable isotope mass standards. This action will preserve for future measurements the valuable and irreplaceable critical samples whose masses and composition have been carefully determined and documented over the past 30 years of the nuclear program

  18. Leaching of actinide-doped nuclear waste glass in a tuff-dominated system

    A laboratory leaching test has been performed as part of a project to evaluate the suitability of tuff rocks at Yucca Mountain, Nevada, as a site for a high-level nuclear waste repository. Glass samples were placed in water inside tuff vessels, and then the tuff vessels were placed in water inside Teflon containers. Glass-component leach rates and migration through the tuff were measured for samples of the ATM-8 actinide glass, which is a PNL 76-68 based glass doped with low levels of 99Tc, 237Np, 238U, and 239Pu to simulate wastes. Disc samples of this glass were leached at 900C for 30, 90, and 183 days inside tuff vessels using a natural groundwater (J-13 well-water) as the leachant. At the end of each leaching interval, the J-13 water present inside and outside the rock vessel was analyzed for glass components in solutions. Boron, molybdenum, and technetium appear to migrate through the rock at rates that depend on the porosity of each vessel and the time. The actinide elements were found only in the inner leachate. Normalized elemental mass loss values for boron, molybdenum, and technetium were calculated using concentrations of the inner and outer leachates and assuming a negligible retention on the rock. The maximum normalized release was 2.3 g/m2 for technetium. Boron, molybdenum, technetium, and neptunium were released linearly with respect to each other, with boron and molybdenum released at about 85% of the technetium rate, and neptunium at 5 to 10% of the technetium rate. Plutonium was found at low levels in the inner leachate but was strongly sorbed on the steel and Teflon supports. Neptunium was sorbed to a lesser extent. 8 refs., 6 figs., 6 tabs

  19. Water Sorption and Radiolysis Studies for Neptunium Oxides

    Plans are to convert the 237Np that is currently stored as a nitrate solution at the Savannah River Site to NpO2 and then ship it to the Y-12 National Security Complex in Oak Ridge for interim storage. This material will serve as feedstock for the 238Pu production program, and some will be periodically shipped to the Oak Ridge National Laboratory (ORNL) for fabrication into targets. The safe storage of this material requires an understanding of the radiolysis of moisture that is sorbed on the oxides, which, in turn, provides a basis for storage criteria (namely, moisture content). A two-component experimental program has been undertaken at ORNL to evaluate the radiolytic effects on NpO2: (1) moisture uptake experiments and (2) radiolysis experiments using both gamma and alpha radiation. These experiments have produced two key results. First, the water uptake experiments demonstrated that the 0.5 wt % moisture limit that has been typically established for similar materials (e.g., uranium and plutonium oxides) cannot be obtained in a practical environment. In fact, the uptake in a typical environment can be expected to be at least an order of magnitude lower than the limit. The second key result is the establishment of steady-state pressure plateaus as a result of the radiolysis of sorbed moisture. These plateaus are the result of back reactions that limit the overall pressure increase and H2 production. These results clearly demonstrate that 0.5 wt % H2O on NpO2 is safe for long-term storage--if such a moisture content could even be practically reached

  20. The effect of sample stability on the determination of radioactivity for various radionuclides by liquid scintillation counting

    For measuring a sample stored for a long period of time using liquid scintillation counting (LSC), it is necessary to study the long-term stability of the sample. The effect of sample stability on the determination of radioactivity for 241Am, 90Sr/90Y, 137Cs, 147Pm, 237Np/233Pa, and 3H by LSC has been investigated. The variation of quench level over time can be an indication of sample stability. If the variation in a sample is little, the effect of sample stability on the determination of the above radionuclides can be neglected. Otherwise, the sample stability will have impact not only on the counting efficiency (especially for low energy β emitters), but also on the results of α/β discrimination. For studying the stability of a sample, special attention should be paid to the radionuclides with chemical form apt to be adsorbed, because the quench level of a sample cannot be reflected by the quench index SQP(E) alone when significant physical quench exists. Shaking a sample stored for a long period of time and checking the LSC spectra can give the information on physical quench in the sample. In the range of this study, OptiPhase Hisafe 3 has much better quench resistance than Ultima Gold AB. - Highlights: • The variation of quench level over time can be an indication of sample stability. • The sample stability has impact on α/β discrimination as well as on the counting efficiency. • Special attention should be paid to the radionuclides with chemical form apt to be adsorbed. • Shaking a sample and checking the LSC spectra can give the information on physical quench. • Significant physical quench usually leads to a new peak in the LSC spectrum of an α emitter

  1. Advantages of Irradiated DUPIC Fuels from the Perspective of Environmental Impact

    This study compares some properties of irradiated Direct Use of Spent Pressurized Water Reactor (PWR) Fuel In Canada Deuterium Uranium reactor (CANDU) (DUPIC ) fuels with properties of other fuel cycles. The properties include the radiotoxicity, decay heat, activity, and actinide content embedded in various spent fuels or high-level wastes, which could be measures of the effectiveness of waste management. From radiotoxicity analysis of fuel cycles, the toxicity of the DUPIC option based on 1 GW(electric).yr is much smaller than those of other fuel cycle options such as the PWR once-through mode, mixed oxide fuel recycling mode, and CANDU once-through mode. The analysis shows that the value is just about half the order of magnitude of other fuel cycles until decayed to a level below the toxicity of initial ore. This means that the DUPIC option could have an indirect benefit on the environmental effects of long-term spent-fuel disposal. From total activity analysis of various fuel cycle options, the activity per metric ton heavy metal of spent fuel is the lowest in natural uranium CANDU fuel, but in the case of activity based on 1 GW(electric).yr, the DUPIC option has the smallest activity. In the meanwhile, from the activity analysis of 99Tc and 237Np, which are important to the long-term transport in geologic media, the DUPIC option was being contained in only about half of those other options. In conclusion, compared to other fuel cycle cases, the irradiated DUPIC fuels would have good properties from the perspective of environmental effects

  2. Advantages of dry-processed oxide fuels from an environmental impact perspective

    This study compares environmental impacts of dry-processed oxide fuels. DUPIC (Direct Use of Spent PWR Fuel In CANDU) fuels, with those of other fuel cycles. For this, some properties of irradiated fuels first were chosen as indirect indicators of environmental impact of fuel cycle. The properties include the radio-toxicity, decay heat, activity and actinide content embedded in various spent fuels or high level wastes which could be measures for the effectiveness of waste management. From radio-toxicity analysis of fuel cycle, the toxicity of the DUPIC option based on 1 GWe-yr is much smaller than those of other fuel cycle options such as PWR (Pressurized Water Reactor) of the once-through mode and MOX (Mixed Oxide) fuel recycling mode and CANDU (Canadian Deuterium Uranium) with once-through mode. It showed that the value is just about half the order of magnitude of other fuel cycles until decayed to a level below toxicity of initial ore. It means that the DUPIC option could have an indirect benefit on the environmental effects of long term spent fuel disposal. From total activity analysis of various fuel cycle options, the activity per metric ton heavy metal of spent fuel is the lowest in natural uranium CANDU fuel. but, in the case of activity based on 1 GWe-yr, the DUPIC option has the smallest activity. In the mean while, from the activity analysis of 99Tc and 237Np, which are important to the long-term transport in geologic media, the DUPIC option was being contained only about half of those other options. In conclusions, the irradiated DUPIC fuels would have good properties in environmental effect aspect, compared to other fuel cycle cases

  3. Evaluated cross section data from Russian reactor dosimetry file

    New updates to the RRDF-98 library were presented. Problems with the covariance matrix in previously submitted evaluations have been eliminated by using extended precision, which resolved the issue of negative Eigenvalues in some of the covariance matrices. A new 237Np(n, f) evaluation was also provided; this file does not yet have a File 1 comments section, but includes the data required to finalize the contents of the IRDF-2002 library. Several example cases were presented where excessive scatter in the experimental data had been resolved by carefully tracing the standards used by the experimenter and re-normalizing the data using the current best estimates of the reference standard cross sections. This renormalization required that Zolotarev track down the experimental details, often contacting the actual experimenters since details were not provided in their written documentation. Cases were shown where the resulting re-normalization procedure dramatically collapsed the spread in the experimental data. Some of Zolotorev's tables have the latest RRDF-98 contributions labelled as IRDF-2002 evaluations. Since the final IRDF-2002 library contents will not be decided until the end of this meeting, these contributions will be relabeled Updated RRDF-98. Zsolnay expressed the views of the whole meeting when she thanked Zolotarev for his extensive contributions to this DDP, and for his quick response to requests to resolve the issues that arose from the reviews. The issue of adding cumulative fission yields to the IRDF-2002 library was discussed since these data are part of the database required in using the new dosimetry cross sections. Trkov noted that the IAEA has an on-going project that addresses this need, but the results will not be available within the timeframe of the IRDF-2002 release. The issue of the addition of fission yields to the IRDF-2002 library was set aside for consideration as part of any future revision

  4. Radiation Effects on Materials in the Near-Field of Nuclear Waste Repository

    Site restoration activities at DOE facilities and the permanent disposal of nuclear waste generated at DOE facilities involve working with and within various types and levels of radiation fields. Once the nuclear waste is incorporated into a final form, radioactive decay will decrease the radiation field over geologic time scales, but the alpha-decay dose for these solids will still reach values as high as 1018 alpha-decay events/gm in periods as short as 1,000 years. This dose is well within the range for which important chemical (e.g., increased leach rate) and physical (e.g., volume expansion) changes may occur in crystalline ceramics. Release and sorption of long-lived actinides (e.g., 237Np) can provide a radiation exposure to backfill materials, and changes in important properties (e.g., cation exchange capacity) may occur. The objective of this research program is to evaluate the long term radiation effects in the materials in the near-field of a nuclear waste repository with accelerated experiments in the laboratory using energetic particles (electrons, ions and neutrons). The proposed target materials for study include zeolites, clays because they are important materials in the near-field expected to retard the migration of radionuclides by absorption or ion exchange processes. Due to the moving of both PI and the co-PI of this program from the University of New Mexico to the University of Michigan in mid-1997, the program schedule has been delayed for several months and now with a DOE approved new ending date of November 14, 1999. Thus, this report is an annual report covering the period from June 1, 1998 to May 30, 1999 , although the initial award was made in mid-1996. The final report will be submitted by the end of 1999

  5. Simultaneous photon and neutron interrogation using an electron accelerator in order to quantify actinides in encapsulated radioactive wastes

    Measuring out alpha emitters, such as (234,235,236,238U 238,239,240,242,244Pu, 237Np 241,243Am...), in solid radioactive waste, allows us to quantify the alpha activity in a drum and then to classify it. The SIMPHONIE (SIMultaneous PHOton and Neutron Interrogation Experiment) method, developed in this Ph.D. work, combines both the Active Neutron Interrogation and the Induced Photofission Interrogation techniques simultaneously. Its purpose is to quantify in only one measurement, fissile (235U, 239,241Pu...) and fertile (236,238U, 238,240Pu...) elements separately. In the first chapter of this Ph.D. report, we present the principle of the Radioactive Waste Management in France. The second chapter deals with the physical properties of neutron fission and of photofission. These two nuclear reactions are the basis of the SIMPHONIE method. Moreover, one of our purposes was to develop the ELEPHANT (ELEctron PHoton And Neutron Transport) code in view to simulate the electron, photon and neutron transport, including the (γ, n), (γ, 2n) and (γ, f) photonuclear reactions that are not taken into account in the MCNP4 (Monte Carlo N-Particle) code. The simulation codes developed and used in this work are detailed in the third chapter. Finally, the fourth chapter gives the experimental results of SIMPHONIE obtained by using the DGA/ETCA electron linear accelerators located at Arcueil, France. Fissile (235U, 239Pu) and fertile (238U) samples were studied. Furthermore, comparisons between experimental results and calculated data of photoneutron production in tungsten, copper, praseodymium and beryllium by using an electron LINear Accelerator (LINAC) are given. This allows us to evaluate the validity degree of the ELEPHANT code, and finally the feasibility of the SIMPHONIE method. (author)

  6. Partitioning of the rare earths and actinides between R7T7 nuclear glass alteration products and solution according to disposal conditions

    The alteration of nuclear glass by water is liable to release radionuclides into the environment. Determining the release kinetics of these elements and their aqueous chemical forms are therefore essential steps in establishing the safety of a geological repository site. Leach tests were conducted with a nonradioactive specimen of the French ''R7T7'' light water containment glass spiked with U and Th, and with two R7T7 specimens spiked with 237Np and 239Pu, respectively. The alteration solution compositions were representative of deep groundwater and contained carbonate, sulfate, phosphate, fluorine and chlorine ions. The release of U, Th, Np and Pu, as well as of the rare earths La, Ce and Nd were monitored by ICP mass spectrometry and by α spectrometry. Scanning and transmission electron microscopic examination of the nonradioactive altered glass surfaces was also performed to assess the partitioning balance for the rare earths, U and Th between the glass alteration products and solution. The mobility of these elements depends on two competing mechanisms. The rare earths and thorium are incorporated in the alteration products (gel); the retention process is assumed to involve chemisorption or coprecipitation, enhanced in the gel layer by the presence of phosphate ions in particular. Conversely, the aqueous species in the alteration solutions (mainly anions) form complexes with the actinides and rare earths; this phenomenon is particularly evident with U and Np. The presence of carbonate ions favors this mobility. Plutonium differs from U and Np in that it is adsorbed mainly on colloids formed by glass dissolution, the principal factors governing its chemical evolution in solution. (author). refs., 122 figs., 185 tabs

  7. Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

    The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the U.S. Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF). The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW). A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP). A performance assessment (PA) was chosen as the method of analysis. The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr. Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis. Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria. This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (1.5%) and mass (3.8%) of waste in the potential repository. Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF

  8. Partitioning and recovery of neptunium from high level waste streams of PUREX origin using 30% TBP

    237Np is one of the longest-lived nuclides among the actinides present in the high level waste solutions of reprocessing origin. Its separation, recovery and transmutation can reduce the problem of long term storage of the vitrified waste to a great extent. With this objective, the present work was initiated to study the extraction of neptunium into TBP under the conditions relevant to high level waste, along with uranium and plutonium by oxidising it to hexavalent state using potassium dichromate and subsequently recovering it by selective stripping. Three types of simulated HLW solutions namely sulphate bearing (SB), with an acidity of ∼ 0.3 M and non-sulphate wastes originating from the reprocessing of fuels from pressurised heavy water reactor (PHWR) and fast breeder reactor (FBR) with acidities of 3.0 M HNO3 were employed in these studies. The extraction of U(VI), Np(VI) and Pu(VI) was very high for PHWR- and FBR-HLW solutions, whereas for the SB-HLW solution, these values were less but reasonably high. Quantitative recovery of neptunium and plutonium was achieved using a stripping solution containing 0.1 M H2O2 and 0.01 M ascorbic acid at an acidity of 2.0 M. Since, cerium present in the waste solutions is expected to undergo oxidation in presence of K2Cr2O7, its extraction behaviour was also studied under similar conditions. Based on the results, a scheme was formulated for the recovery of neptunium along with plutonium and was successfully applied to actual high level waste solution originating from the reprocessing of research reactor fuels. (author). 19 refs., 2 figs., 17 tabs

  9. The state of the art on nuclides separation in high level liquid wastes by Truex process

    For the advancement of the back-end of nuclear fuel cycle, novel CMPO RUEX process was studied for separating minor actinides from fission products in high level liquid waste using real radioactive solutions from PUREX experiments, so as to support PNC's actinides recycling program using fast reactor. The present PUREX process was also studied to improve the separation of 237Np, 106Ru and 99Tc, the most interfering-natured nuclides in both PUREX and TRUEX processes, by utilizing electrochemistry-based salt-free methods which can eliminate the secondary radioactive waste. The state of the art of separation technologies are described by summarizing the extraction behaviors of nuclides in recent hot counter-current runs using CMPO RUEX process with mild salt-free stripping reagents. The degradation and regeneration characteristics of CMPO/TBP/n-dodecane mixture solvent were also simulated by semi-hot experiments. Several experiments to separate minor actinides and lanthanides from the TRUEX mixture product using aqueous amino-poly-carboxylate complexant, DTPA, resulted in reasonable MA/Ln separation profiles in multiple mixer-settler stages and allowed a unique separation flowsheet adaptable to the TRUEX process to be proposed. Application of electrochemistry to assist both solvent extraction processes, e.g., 'anodic oxidation' to destroy PUREX and TRUEX solvent waste in the presence of electron transfer mediator Age2+ or 'cathodic reduction' for electrolytic extraction of Pd2+, RuNO3+ and 99TcO4- from 3 M nitric acid medium is under study. (authors)

  10. Intercomparison of Lepricon (SN) and Tripoli (MC) calculated and measured neutron reaction rates in the ex-vessel cavity of SLB1

    One of the key-pieces of the PWR safety is the pressure vessel. Therefore the surveillance of the degradation of the vessel steel mechanical properties due to neutron exposure is crucial problem. This surveillance programme is based on the correlation of the neutron dosimetry and the mechanical tests results obtained from the samples loaded in the surveillance capsules. These correlations are then extrapolated to the pressure vessel position using the lead factors obtained by means of transport codes based on SN or Monte Carlo methods. The general political tendency of the utilities towards PLIM (Plant LIfe Management) and the latest changes in the fuel loading management necessitate a neutron surveillance dosimetry more accessible but as accurate as the one of the surveillance capsules. One answer to this problem in the ex-vessel dosimetry loaded between the pressure vessel and the concrete biological shielding. The emerging question then concerns the validity of the neutron transport calculation tools to analyze the ex-vessel dosimetry whereas they were qualified for in-vessel dosimetry analysis. To answer this question an ex-vessel neutron dosimetry validation programme was conducted in the cavity of the French 900 MWe PWR SLB1. Different threshold detectors were loaded in the cavity among them: Cu, Ni, Fe, 238U, 237Np. In this paper we will report on the comparison of the measured and the calculated ones using the LEPRICON code system in association with ELXSIR library and the TRIPOLI Monte Carlo code associated to an ENDF/B-VI based library. (authors)

  11. Set up of an innovative methodology to measure on-line the incineration potential of minor actinides under very high neutron sources in the frame of the future prospects of the nuclear waste transmutation; Mise au point d'une methodologie innovante pour la mesure du potentiel d'incineration d'actinides mineurs sous des sources tres intenses de neutrons, dans la perspective de transmutation des dechets nucleaires

    Fadil, M

    2003-03-01

    This work deals generally with the problem of nuclear waste management and especially with the transmutation of it to reduce considerably its radiotoxicity potential. The principal objective of this thesis is to show the feasibility to measure on-line the incineration potential of minor actinides irradiated under very high neutron flux. To realize this goal, we have developed fission micro-chambers able to operate, for the first time in the world, in saturation regime under a severe neutron flux. These new chambers use {sup 235}U as an active deposit. They were irradiated in the high flux reactor at Laue-Langevin Institute in Grenoble. The measurement of the saturation current delivered by these chambers during their irradiation for 26 days allowed to evaluate the burn-up of {sup 235}U. We have determined the neutron flux intensity of 1,6 10{sup 15} n.cm{sup -2}.s{sup -1} in the bottom of the irradiation tube called 'V4'. The relative uncertainty of this value is less than 4 %. This is for the first time that such high neutron flux is measured with a fission chamber. To confirm this result, we have also performed independent measurements using gamma spectroscopy of irradiated Nb and Co samples. Both results are in agreement within error bars. Simple Deposit Fission Chambers (SDFC) as above were the reference of the new generation of fission chambers that we have developed in the framework of this thesis: Double Deposit Fission Chambers (DDFC). The reference active deposit was {sup 235}U. The other deposit was the actinide that we wanted to study (e.g. {sup 237}Np and {sup 241}Am). At the end of the thesis, we present some suggestions to ameliorate the operation of the DDFC to be exploited in other transmutation applications in the future. (author)

  12. Part A: Radioactive SON 68 18 17 L1C2A2Z1 glass interaction with environmental materials. Part B: Investigation of irradiation damage in glass specimens by thermoluminescence

    This paper is divided into two parts. Part A: Radioactive doped glass with 237Np and 239Pu (R7T7 glass) was leached in contact with different environmental materials: smectite, illite, bentonite, sand, granite, Boom clay and French salt. The corrosion test results confirmed the significant role of these materials in glass alteration. The best results were obtained with bentonite, which not only limited glass corrosion by supplying silicon to the solution, but also reduced the quantity of actinides in solution by fixing them on clay folia. Granite and sand did not result in increased corrosion compared with double- distilled water: the actinide retention factor in the alteration film formed in contact with these materials was not lower than in double distilled water, and they appear better suited for fixing 239Pu and 241Am. The poorest results were obtained with Boom clay, not only because of increase glass corrosion but also because of the presence of humic acids and organic compounds that lead to the formation of complexes maintaining a large fraction of the actinides in solution. Bench-scale experiments showed that glass is only slightly altered in salt and granite media provided they are not implemented with a clay engineered barrier of the type used in these performing tests. A very small amount of clay in the granite is sufficient to result in significantly higher corrosion. Part B: Physical radiation effects in radioactive waste glasses was investigated by means of thermoluminescence (TL). After a description of radiation damage and the TL properties of a glass, the report discusses the detection and description of 'alpha' radiation damage. A radiation damage mechanism based on irradiation-induced displacement of oxygen atoms is proposed and partial reversibility is demonstrated

  13. Updating of the performance assessments of the geological disposal of high-level and medium-level wastes in the Boom clay formation

    The objective of this report is to assess the performance of disposed medium-level and alpha-bearing waste in a geological repository in the Boom clay formation at the Mol site (Belgium). The results of this study are based on calculations that are in agreement with recent information on the Belgian nuclear programme and the corresponding waste arising. The applied methodology consists of two consecutive steps: (1) a scenario analysis in which relevant scenarios, leading to the exposure of man to radiation are selected, and (2) a consequence analysis in which potential radiological consequences of the exposure are evaluated. The scenario, selected in this study, is designated as the normal evolution scenario and comprises a normal evolution scenario in which the present conditions are assumed to last infinitely. The scenario is extended with a climatic change, a secondary glaciation effects, and a faulting scenario. The applied consequence analysis consists in deterministic and stochastic calculations, which are are complementary. Three pathways of radionuclides to man were considered: (1) the discharge of contaminated groundwater into rivers or (2) into agricultural soils, and (3) the sinking of a water well into the aquifer that overlies the host formation. Calculations indicate that most radionuclides decay within the first metres of the clay barrier. The fission and activation products 14C, 129I, 79Se, 99Tc, 107Pd, 93Zr, and 135Sr as well as some actinides of the 237Np and uranium decay series can however reach the biosphere. The maximum dose rates for high-level waste, fuel cladding, medium-level, and iodine waste were calculated. Deterministic calculations indicate that the maximum dose rates are attained via the water well pathway for the case of the climatic change scenario. (A.S.)

  14. Study of calculated and measured time dependent delayed neutron yields. [TX, for calculating delayed neutron yields; MATINV, for matrix inversion; in FORTRAN for LSI-II minicomputer

    Waldo, R.W.

    1980-05-01

    Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of /sup 232/U, /sup 237/Np, /sup 238/Pu, /sup 241/Am, /sup 242m/Am, /sup 245/Cm, and /sup 249/Cf were studied for the first time. The delayed neutron emission from /sup 232/Th, /sup 233/U, /sup 235/U, /sup 238/U, /sup 239/Pu, /sup 241/Pu, and /sup 242/Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from /sup 232/Th to /sup 252/Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables.

  15. Fabrication and characterization of MCC approved testing material - ATM-12 glass

    Wald, J.W.

    1985-10-01

    The Materials Characterization Center (MCC) Approved Testing Material ATM-12 is a borosilicate glass that incorporates elements typical of high-level waste (HLW) resulting from the reprocessing of commercial nuclear reactor fuels. The composition has been adjusted to match that predicted for HLW type 76-68 glass at an age of 300 y. Radioactive constituents contained in this glass include depleted uranium, {sup 99}Tc, {sup 237}Np, {sup 239}Pu, and {sup 241}Am. The glass was produced by the MCC at the Pacific Northwest Laboratory (PNL). ATM-12 glass ws produced from July to November of 1984 at the request of the Nevada Nuclear Waste Site Investigations (NNWSI) Program and is the third in a series of glasses produced for NNWSI. Most of the glass produced was in the form of cast bars; special castings and crushed material were also produced. Three kilograms of ATM-12 glass were produced from a feedstock melted in a nitrogen-atmosphere glove box at 1150{sup 0}C in a platinum crucible, and formed into stress-annealed rectangular bars and the special casting shapes requested by NNWSI. Bars of ATM-12 were nominally 1.9 x 1.9 x 10 cm, with an average mass of 111 g each. Nineteen bars and 37 special castings were made. ATM-12 glass has been provided to the NNWSI Program, in the form of bars, crushed powder and special castings. As of August 1985 approximately 590 g of ATM-12 is available for distribution. Requests for materials or services related to this glass should be directed to the Materials Characterization Center Program Office, PNL.

  16. Determination of 51Cr and 241Am X-ray and gamma-ray emission probabilities per decay

    Full text: In this paper results of X-ray and gamma-ray emission probabilities per decay of 51Cr and 241Am are presented. The measurements were carried out by means of an HPGe planar spectrometer. The radionuclide 51Cr decays with a half-life of 27.7 days by electron capture process populating the excited levels of 51V, which emits X-rays between 4 and 6 keV and a gamma ray of 320 keV. The radionuclide 241Am decays with a half-life of 157,850 days by alpha emission, populating the excited levels of 237Np, which emits X-rays between 11 and 20 keV and two main gamma rays of 26 and 59 keV. The choice of 51Cr was made due to the need of more results since there are two discrepant sets of values for the 320 keV emission probability per decay in the literature. The choice of 241Am was made because, although this radionuclide be considered a primary standard for spectrometers calibration using its well-known 59 keV gamma ray emission probability, the 26 keV gamma ray and the X-rays emission probabilities present large discrepancies in the literature. The activity of 51Cr and 241Am samples was determined in a 4-coincidence counting system. The same sources were measured in an HPGe planar spectrometer with Be window, suitable for measurements in low energy range. The HPGe spectrometer was calibrated in a well defined geometry by means of 54Mn, 55Fe, 57Co, 133Ba, 152Eu, 166mHo and 241Am (59 keV) sources, previously standardized in a coincidence system. The MCNP Monte Carlo code was used for simulation of the spectrometer calibration curve for the selected geometry, and compared with the experimental one

  17. Study of calculated and measured time dependent delayed neutron yields

    Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of 232U, 237Np, 238Pu, 241Am, /sup 242m/Am, 245Cm, and 249Cf were studied for the first time. The delayed neutron emission from 232Th, 233U, 235U, 238U, 239Pu, 241Pu, and 242Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from 232Th to 252Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables

  18. Transmutation of high level wastes in a fusion-driven transmuter (FDT)

    This study presents the transmutations of both the minor actinides (MAs: 237Np, 241Am, 243Am and 244Cm) and the long-lived fission products (LLFPs: 99Tc, 129I and 135Cs), discharged from high burn-up PWR-MOX spent fuel, in a fusion-driven transmuter (FDT) and the effects of the MA and LLFP volume fractions on their transmutations. The blanket configuration of the FDT is improved by analyzing various sample blanket design combinations with different radial thicknesses. Two different transmutation zones (TZMA and TZFP which contain the MA and LLFP nuclides, respectively) are located separately from each other. The volume fraction of the MA is raised from 10 to 20% stepped by 2%. The MAs are cladded with the graphite (10%) and cooled with the high-pressured helium gas for nuclear heat transfer. The volume fraction of helium is reduced from 80 to 70% depending on that of MA. Furthermore, the volume fraction of graphite is raised from 10 to 80% stepped by 5% to slow down the energy of neutrons entering into the TZFP while the volume fraction of LLFP is reduced from 80 to 10% depending on the graphite volume fraction. The calculations are performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 5 MW/m2 to estimate neutronic parameters and transmutation characteristics per D-T fusion neutron. The transmutation rates of the LLFP nuclides increase linearly with the increase of volume fractions of the MA, and the 99Tc nuclide among them has the highest transmutation rate

  19. Solar r-process-constrained actinide production in neutrino-driven winds of supernovae

    Goriely, S.; Janka, H.-Th.

    2016-07-01

    Long-lived radioactive nuclei play an important role as nucleo-cosmochronometers and as cosmic tracers of nucleosynthetic source activity. In particular, nuclei in the actinide region like thorium, uranium, and plutonium can testify to the enrichment of an environment by the still enigmatic astrophysical sources that are responsible for the production of neutron-rich nuclei by the rapid neutron-capture process (r-process). Supernovae and merging neutron-star (NS) or NS-black hole binaries are considered as most likely sources of the r-nuclei. But arguments in favour of one or the other or both are indirect and make use of assumptions; they are based on theoretical models with remaining simplifications and shortcomings. An unambiguous observational determination of a production event is still missing. In order to facilitate searches in this direction, e.g. by looking for radioactive tracers in stellar envelopes, the interstellar medium or terrestrial reservoirs, we provide improved theoretical estimates and corresponding uncertainty ranges for the actinide production (232Th, 235, 236, 238U, 237Np, 244Pu, and 247Cm) in neutrino-driven winds of core-collapse supernovae. Since state-of-the-art supernova models do not yield r-process viable conditions - but still lack, for example, the effects of strong magnetic fields - we base our investigation on a simple analytical, Newtonian, adiabatic and steady-state wind model and consider the superposition of a large number of contributing components, whose nucleosynthesis-relevant parameters (mass weight, entropy, expansion time-scale, and neutron excess) are constrained by the assumption that the integrated wind nucleosynthesis closely reproduces the Solar system distribution of r-process elements. We also test the influence of uncertain nuclear physics.

  20. Influence of gamma irradiation on the deterioration of reactor pressure vessel materials and on reactor dosimetry measurements. Final report

    Radiation embrittlement of pressure vessel steel in mixed neutron-gamma fields is mostly determined by neutrons, but in some cases also by gamma-radiation. Depending on the reactor type, gamma radiation can influence evaluations of lead factors of surveillance specimens, effect the interpretation of results of irradiation experiments and finally, it can result in changed pressure vessel lifetime evaluations. The project aimed at the evaluation of the importance of gamma radiation for RPV steel damage for several types of light-water reactors. Absolute neutron and gamma fluence rate spectra had been calculated for the Russian PWR types VVER-440 and two core loading variants of VVER-1000, for a German 1300 MW PWR and a German 900 MW BWR. Based on the calculated spectra several flux integrals and radiation damage parameters were derived for the region of the azimuthal flux maxima in the mid-planes for different radial positions between core and biological shield, especially, at the inner and outer surfaces of the PV walls, at the (1/4)-PV-thickness and at the surveillance positions. Fissionable materials are often used as activation detectors for neutron fluence measurements. To get the real value the analysis demands to take into account the gamma induced fissions. Therefore, the part of these fissions in the total number of fissions was estimated for the detector reactions 237Np(n,f) and 238U(n,f) in the calculated neutron/gamma fields. It has been found that considerable corrections of the neutron fluence measurements can be necessary, especially in case of 238U(n,f). Most of the calculations were performed using a three-dimensional synthesis of 2D/1D-flux distributions obtained by the SN-code DORT with the BUGLE-96T group cross-section library. (orig.)