The Oklo and Bangombé natural fossil fission reactors formed ca. 2 Ga ago in the Franceville basin, Gabon. The response of uraninite in the natural reactors to different geological conditions has implications for the disposal of the UO2 in spent nuclear fuel. Uraninite and galena from two reactor zones, RZ16 at Oklo and RZB at Bangombé, were studied to clarify the chronology and effect of alteration events on the reactor zones. In addition, ion microprobe U-Pb analysis of zircons from a dolerite dyke in the Oklo deposit were completed to better constrain the age of the dyke, and thereby testing the link between the dyke and an important alteration event in the reactor zones. The analyzed uraninite from RZ16 and RZB contains ca. 6 wt% PbO, indicating a substantial loss of radiogenic Pb. Transmission electron microscopy showed that microscopic uraninite grains in the reactor zones consist of mainly defect-free nanocrystalline to microcrystalline aggregates. However, the nanocrystalline regions have elevated Si contents and lower Pb contents than coarser uraninite crystallites. Single stage model ages of large, millimeter-sized galena grains at both RZ16 and RZB correlate well with the age of the Oklo dolerite dyke, 860 ± 39 Ma (2?). Thus, the first major Pb loss from uraninite occurred at both Oklo and Bangombé during regional extension and the intrusion of a dyke swarm in the Franceville basin, ˜860 890 Ma ago. Uraninite Pb isotopes from RZ16 and RZB give lower ages of ca. 500 Ma. These ages agree with the “chemical” ages of the uraninite, and show that an ancient Pb loss occurred after the intrusion of the dolerite dykes. The presence of nanocrystallites in the reactor uraninite indicates internal recrystallization, which may have occurred around 500 Ma, resulting in the 6wt% PbO uraninite. It is suggested that leaching by fluid interaction triggered by the Pan-African orogeny was important during this second Pb-loss event. Thus, there are indications that uraninite at both the Oklo and Bangombé natural reactors has experienced at least two ancient episodes of Pb loss associated with internal recrystallization. These recrystallization events have occurred without significantly depleting the 2 Ga fission products compatible with the uraninite structure.

The possibility of continuous nuclear fission chain reactions during four gigayears up to the present in the interior of the Earth is studied. Natural fast reactor in the form of lakes could be formed as a result of precipitation of uranium from the liquid layer on the solid Earth core. Mechanism of uranium concentration in the Earth core is described. In such layers a chain nuclear reaction with new fissile nuclides breeding could proceed. Some characteristics of georeactor have been calculated. Its operation in a pulse mode as it was in case of a natural nuclear reactor in Oklo (Gabon) is most probable.

One of the main part of he CEC project `Oklo-Natural Analogue` is devoted to present time migration studies. This part comprises hydrogeology, groundwater chemistry, isotope geochemistry and modelling. Two sites are being investigated: the less perturbed reactor zone of the Oklo mine (OK84 in the southern mine extension of Okelobondo) at around 400 meters depth and the Bangombe reactor zone, sited in a shallow environment 30 km south of Oklo. The present contribution aims to define regional hydrogeology and hydro-chemistry boundary conditions for the modelling exercise, to assess the present day water-rock interaction in the vicinity of reactor zones, to gather information on the geochemical conditions which allowed the preservation of reactor zones for two billions years, to estimate the uranium migration from the reactor zone in using a natural marker (the depleted {sup 235}U/{sup 238}U ratio resulting from the fission) and to compare these data with predictive modelling. Based on the hydrogeological conceptual modelling, we have sampled waters in recharge areas, discharge areas above and below reactors, and in major local aquifers. We have been able to reconstruct the evolution of the groundwaters, in a way which is consistent with the hydrogeology, using major elements and environmental isotopes. (author) 24 refs.

Two billion years ago, the increase of oxygen in atmosphere and the high {sup 235}U/{sup 238}U uranium ratio (> 3%) made possible the occurrence of natural nuclear reactors on Earth. These reactors are considered to be a good natural analogue for nuclear waste disposal. Their preservation during such a long period of time is mainly due to the geological stability of the site, the occurrence of clays surrounding the reactors and acting as an impermeable shield, and the occurrence of organic matter that maintained the environment in reducing conditions, favourable for the stability of uraninite. Hydrogeochemical studies and modelling have shown the complexity of the geochemical system at Oklo and Bangombe (Gabon) and the lack of precise data about uranium and fission products retention and migration mechanisms in geological environments. (author)

Lutetium thermometry has been used to analyze Oklo natural nuclear reactor zones but leads to widely varying and puzzling predictions for the temperatures TO, which in turn impacts Oklo bounds on the time variation of the fine structure constant ?. We revisit results for reactor zone RZ10 in light of new astrophysical measurements of the isomer branching ratio B^g in ^175Lu neutron capture at 5 and 25 keV. We recalculate predictions for TO as a function of B^g using realistic models of the Oklo neutron flux. We find TO= 100 ±30 C using a new value of B^g, in contrast to 350 energy. Lutetium thermometry can be applicable to analyses of Oklo reactor data, but a better measurement of B^g with thermal neutrons is needed to confirm the reliability of temperature predictions.

Lutetium thermometry has been used to analyze Oklo natural nuclear reactor zones but leads to widely varying and puzzling predictions for the temperatures $T_O$ which in turn impacts bounds on time variation of the fine structure constant $\\alpha$. We revisit results for reactor zone RZ10 in light of new measurements of the isomer branching ratio $B^g$ in $^{175}$Lu neutron capture at 5 and 25 keV. We recalculate predictions for $T_O$ as a function of $B^g$ using realistic models of the Oklo neutron flux. We find $T_O = 100 \\pm 30$ C using a new value of $B^g$, in contrast to $350 < T_O < 500 $ C using the evaluated value at thermal energy. Lutetium thermometry can be applicable to analyses of Oklo reactor data, but a better measurement of $B^g$ with thermal neutrons is needed to confirm the reliability of temperature predictions.

This report represents an analog approach to the characterization of the environmental behavior of geotoxic waste materials (toxic material emplaced in the earth's crust) as drawn from literature on the Oklo natural fission reactors and uranium ore deposits relative to radioactive wastes, and hydrothermal metal ore deposits relative to stable toxic wastes. The natural analog data were examined in terms of mobility and immobility of selected radioactive or stable waste elements and are presented in matrix relationship with their prime geochemical variables. A numerical system of ranking those relationships for purposes of hazard-indexing is proposed. Geochemical parameters (especially oxidation/reduction potential) are apparently more potent mobilizers/immobilizers than geological or hydrological conditions in many, if not most, geologic environments for most radioactive waste elements. Heavy metal wastes, by analogy to hydrothermal ore systems and geothermal systems, are less clear in their behavior but similar geochemical patterns do apply. Depth relationships between geochemical variables and waste element behavior show some surprises. It is significantly indicated that for waste isolation, deeper is not necessarily better geochemically. Relatively shallow isolation in host rocks such as shale could offer maximum immobility. This paper provides a geochemical outline for examining analog models as well as a departure point for improved quantification of geological and geochemical indexing of toxic waste hazards.

Rock samples from the Northeastern region of Brazil were analysed for their /sup 235/U isotopic abundance, in search for the occurrence of an ''Oklo Phenomenon'' here. The samples were collected in locations that could have been connected to the African continent, according to the continental-drift theory, in accordance to the Francevillian formations in the Gabon Republic, in which place the Oklo natural fossil reactor is situated. Two methods were used for the determination of the /sup 235/U abundance: activation analysis followed by high resolution gamma-ray spectrometry and activation analysis by delayed neutron counting. No evidence of /sup 235/U depletion was found in the rock samples analysed.

In 2005, the global inventory of spent nuclear fuel (SNF) is approximately 175,000 metric tonnes (slightly less than one third is in the USA) (Ewing, 2004). Most of this SNF is still at 236 nuclear power stations where it was originally generated in 36 different countries. In the USA, the inventory in 2010 will be 61,800 metric tonnes of heavy metal (tHM) with a total activity of 32.6 GCi. The USA presently has an open nuclear fuel cycle (without reprocessing) with ultimate disposal at the proposed geologic repository at Yucca Mountain. The SNF represents >95% of the radioactivity. Thus, a major challenge of successful geologic disposal of radioactive waste is to understand the long-term behavior of SNF. SNF is essentially UO{sub 2} with minor impurities, mainly the fission product (3%) and transuranium elements (1%). The precise radionuclide inventory and physical state of the fuel depend on its irradiation and thermal history. Three critical parameters change dramatically during the first 10,000 years in the repository: (1) the thermal output will decrease to < 0.1%; (2) the radioactivity will decrease to < 0.01%; (3) the inventory of radiotoxic nuclides will change. Beyond 10,000 years radionuclides of major importance under oxidizing conditions include: {sup 239}Pu, {sup 237}Np, {sup 129}I and {sup 99}Tc. Less problematic elements include: {sup 241}Am, {sup 79}Se and {sup 36}Cl. These elements exist in a variety of chemical forms: incorporated into the UO{sub 2} structure, as separate phases in inclusions and at grain boundaries. Corrosion under oxidizing conditions leads to the formation of a variety of U(VI)-phases. An understanding of their long-term behavior requires an improved knowledge of their structures, thermochemical parameters, solubilities, substitution mechanisms for trace radionuclides, surface properties and the kinetics of dissolution/precipitation reactions. Natural uranium deposits, such as the Oklo natural reactors, also provide important data. This paper reviews recent research on these topics, and its relation to the properties of SNF.

Pierre Curie gave two hypotheses at first; (1) It can be supposed that the radioactive substances borrow the energy, which they release, from an external radiation, and their radiation would then be a secondary radiation, (2) It can be supposed that the radioactive substances draw from themselves the energy which they release. The second hypothesis has shown the more fertile in explaining the properties of the radioactive substances. Consequently, the first hypothesis became more or less forgotten. It appears, however, the first hypothesis should play an important role in explaining the phenomena concerning the origin of the elements. The Oklo Phenomenon has demonstrated that a nuclear fire had once existed on our planet earth and formation of heavy elements was occurring in nature. The author pointed out that the difference in the isotopic compositions of xenon found in meteorites, lunar samples and in the earth's atmosphere can only be explained as due to the alterations of the isotropic compositions of xenon by combined effect of (a) mass-fractionation, (b) spallation, and (c) stellar temperature neutron-capture reactions. The strange xenon components are not isotopically pure substance. Instead, xenon-HL is a mixture of the {sup 244}Pu fission xenon and the xenon whose isotopic compositions is severely altered by a combined effect of the processes (a), (b) and (c) mentioned above. These results also indicate that C1 carbonaceous chondrites, which is generally as the most primitive sample of the solar system material, began to retain its xenon 5.1 billion years ago, when the plutonium to uranium ratio in the solar system was as high as almost 0.6 (atom/atom), while the C2 carbonaceous chondrite began to retain their xenon about 150 million years later and the ordinary chondrites and achondrite about 500 to 600 million years later. This means that the birth of the solar system began soon after the last supernova exploded about 5.1 billion years ago, and the generally accepted 4.55 billion year-age of the solar system is likely to be the time of the breakup of the meteorite parent body. (Y. Tanaka)

While it is possible to synthesize transuranium elements up to proton number of Z = 105, only a small success can be reported on naturally occurring transuranium nuclides. Only extremely small quantities of primordial plutonium- 244, the longest-llved transuranium nuclide, have been detected in the cerium- rich Californian mineral bastnaesite, which means that plutonium must be considered as a naturally occurring and not as an artificial element. Promethium- 147, plutonium-239, and neptunium-237, detected in several minerals and identified by ucts. No other transuranium nuclides and no superheavy elements could be found in terrestrial, lunar, and meteoritic probes. Investigations on the anomalous isotopic composition of uranium in ores from Oklo./Gabon led to a scientific sensation: a prehistoric natural nuclear reactor, working about 10/sup 9/ years ago, which showed much similarity in comparison with modern nuclear power reactors. (GE)

The sites at Bangombé and Okélobondo (Oklo) in Gabon provide a unique opportunity to study the behaviour of products from natural nuclear reactions in the vicinity of reactor zones which were active around two billion years ago. The Commission of the European Communities initiated the Oklo Natural Analogue Programme. One of the principal aims was to study indications of present time migration of elements from the reactor zones under ambient conditions. The hydrogeological and hydrochemical data from the Oklo sites were modelled in order to better understand the geochemical behaviour of radionuclides in the natural system, by using independent models and by comparing the modelling outcome. Two modelling approaches were used: M3 code (hydrochemical mixing and mass balance model), developed by the Swedish Nuclear Fuel and Waste Management Company (SKB) and HYTEC (reactive transport model) developed by Ecole des Mines de Paris. Two different reactor zones were studied: Bangombé, a shallow site, the reactor being at 11 m depth, and OK84 at Okélobondo, situated at about 450 m depth, more comparable with a real repository location. This allowed the validation of modelling tools in two different sedimentary environments: one shallow, with a more homogeneous layering situated in an area of meteoric alteration, and the other offering the opportunity to study radionuclide migration from the reaction zone over a distance of 450 m through very heterogeneous sedimentary layers. The modeling results indicate that the chemical reactions retarding radionuclide transport are very different at the two sites. At Bangombé, the decomposition of organic material consumes oxygen and at Okélobondo the oxygen is consumed by inorganic reactions resulting, in both cases, in uranium retardation. Both modelling approaches (statistic with M3 code and deterministic with HYTEC code) could describe this situation. The goal of this exercise is to test codes which can help to describe and understand the processes taking place at the sites, validate the models with in situ data, and thus build confidence in the tools used for future site characterization. Ultimately, this allows identifying and selecting processes and parameters that can be used as input into repository performance assessment calculations and modelling exercises. PMID:12598108

We present a review of recent works devoted to the variation of the fine structure constant alpha, strong interaction and fundamental masses. There are some hints for the variation in quasar absorption spectra, Big Bang nucleosynthesis, and Oklo natural nuclear reactor data. A very promising method to search for the variation of the fundamental constants consists in comparison of different atomic clocks. Huge enhancement of the variation effects happens in transition between accidentally degenerate atomic and molecular energy levels. A new idea is to build a ``nuclear'' clock based on the ultraviolet transition between very low excited state and ground state in Thorium nucleus. This may allow to improve sensitivity to the variation up to 10 orders of magnitude! Huge enhancement of the variation effects is also possible in cold atomic and molecular collisions near Feschbach resonance.

?Astroneutron?: Theoretical and Experimental Investigations into Neutron Cross-Sections of Neutron-Enriched Nuclei Capture and Fission, Creation of a Data Bank and Simulation of Natural Nucleosynthesis Processes.

Fission track dating is described, focalizing the problem of the decay constant for spontaneous fission of {sup 238} U and the use of neutron dosimetry in fission track analysis. Experimental procedures using thin films of natural uranium as neutron dosimeters and its results are presented. The author shows a intercomparison between different thin films and between the dosimetry with thin film and other dosimetries. (M.V.M.). 52 refs, 12 figs, 9 tabs.

This research program is a broadly based effort to understand the long-term behavior of spent nuclear fuel (SNF) and its alteration products in a geologic repository. We have established by experiments and field studies that natural uraninite, UO2+x, and its alteration products are excellent ''natural analogues'' for the study of the corrosion of UO2 in SNF. This on-going research program has addressed the following major issues: (1) What are the long-term corrosion products of natural UO2+x, uraninite, under oxidizing and reducing conditions? (2) What is the paragenesis or the reaction path for the phases that form during alteration? (3) What is the radionuclide content in the corrosion products as compared with the original UO2+x? Do the trace element contents substantiate models developed to predict radionuclide incorporation into the secondary phases? Are the corrosion products accurately predicted from geochemical codes (e.g., EQ3/6 or Geochemist's Workbench) that are used in performance assessments? Can these codes be tested by studies of natural analogue sites (e.g., Oklo, Cigar Lake or Pena Blanca).

Understanding of the environmental behavior of Ra is important from the viewpoint of the long-termed repository safety of radioactive waste, but investigation of Ra behavior in natural environment is difficult to detect. We found isotopic evidence of Ra transportation and its selective uptake into clay minerals from Pb isotopic analyses. Illite grains found in calcite veins included in sandstone near the Oklo uranium deposit, Republic of Gabon, show extremely low 207Pb/206Pb (˜ 0.0158) isotopic ratios. Although the Pb isotopic ratios of calcite and quartz coexisting with illite indicate the formation age of each component, those of illite do not. In addition, illite grains having low 207Pb/206Pb isotopic ratios contain a strongly large amount of Ba (1230 to 6010 ppm) in contrast with low contents of Ba in calcite and quartz (< 0.26 ppm). Considering the chemical similarity between Ba and Ra, the 207Pb/206Pb isotopic data suggest an excess of 206Pb due to selective adsorption of 226Ra (and also Ba) into illite grains. This is a very rare example to show evidence of the selective adsorption behavior of Ra from the isotopic excesses of 206Pb, although the adsorption ability of Ra itself in nature was largely reported.

The event-by-event fission model FREYA is extended to spontaneous fission of actinides and a variety of neutron observables are studied for spontaneous fission and fission induced by thermal neutrons with a view towards possible applications for SNM detection. We have shown that event-by-event models of fission, such as FREYA, provide a powerful tool for studying fission neutron correlations. Our results demonstrate that these correlations are significant and exhibit a dependence on the fissioning nucleus. Since our method is phenomenological in nature, good input data are especially important. Some of the measurements employed in FREYA are rather old and statistics limited. It would be useful to repeat some of these studies with modern detector techniques. In addition, most experiments made to date have not made simultaneous measurements of the fission products and the prompt observables, such as neutron and photons. Such data, while obviously more challenging to obtain, would be valuable for achieving a more complete understanding of the fission process.

The talk presents the results of an investigation of the main characteristics (mass and energy distributions of fission fragments and multiplicity of neutrons) of the fission of the nuclei of {sup 220}Ra, {sup 226}Th, {sup 256}No, {sup 270}Sg, {sup 286}112 produced in reactions with ions of {sup 18}O, {sup 22}Ne and {sup 48}Ca at energies close to and essentially below the Coulomb barrier. The data obtained show that the form of the mass and energy distributions of the fission fragments of {sup 226}Th and {sup 270}Sg is accounted for by the multimodal nature of the fission. In addition, for {sup 226}Th, a new phenomenon was established: there is a significant difference between the numbers of prescission neutrons for symmetric and asymmetric fission modes. It was found that, for the low-energy fission of the nucleus of {sup 286}112, the mass distribution of the fragments is of a clear-cut asymmetric form, contrary to what is observed for the spontaneous fission of the nuclei with Z=100-104 and for the induced fission of {sup 270}Sg.

A report on the immobilization of uranium in the earth's crust has been completed. Techniques have been developed to do a comprehensive mass inventory of the Oklo reactor zones. These techniques were applied to a compilation of data from Oklo zones 2 and 3-4. The study shows large deficiencies of neodymium, ruthenium, and mass 99 elements (/sup 99/Tc or /sup 99/Ru) in the reactor zones. The extent of these deficiencies are correlated with the intensity of the nuclear reactions. Analyses of ores from the Key Lake uranium mineralization show that 60 to 70% of the radiogenic lead is missing from the ores.

High-speed cinematography has been used to characterize the macroscopic behavior of irradiated and unirradiated fuel subjected to thermal transients prototypical of fast reactor transients. The results demonstrate that as the cladding melts, the fuel can disperse via spallation if the fuel contains in excess of approx. 16 ..mu..moles/gm of fission gas. Once the cladding has melted, the macroscopic behavior (time to failure and dispersive nature) was strongly influenced by the presence of volatile fission products and the heating rate.

Natural production rates of the 16 million-year I/sup 129/ from spontaneous fission and cosmic-ray reactions are estimated to contribute a steady- state concentration of more than 10/sup -14/ gram of I/sup 129/ per gram of 1/sup 127/ to the hydrosphere, atmosphere, and biosphere. The concentrations are expected to be detectable by neutron-activation analysis in natural materials that concentrate iodine. Substantial additional contributions were made by fission-produced I/sup 129/ in fallout, serving as an intrinsic tracer for natural iodine kinetics. I/sup 129/ activation analysis has advantages as a tracer technique. (auth)

The Laser Inertial Confinement Fusion Fission Energy (LIFE) Engine [1] combines a neutron-rich but energy-poor inertial fusion system with an energy-rich but neutron-poor subcritical fission blanket. Because approximately 80% of the LIFE Engine energy is produced from fission, the requirements for laser efficiency and fusion target performance are relaxed, compared to a pure-fusion system, and hence a LIFE Engine prototype can be based on target performance in the first few years of operation of the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL). Similarly, because of the copious fusion neutrons, the fission blanket can be run in a subcritical, driven, mode, without the need for control rods or other sophisticated reactivity control systems. Further, because the fission blanket is inherently subcritical, fission fuels that can be used in LIFE Engine designs include thorium, depleted uranium, natural uranium, spent light water reactor fuel, highly enriched uranium, and plutonium. Neither enrichment nor reprocessing is required for the LIFE Engine fuel cycle, and burnups to 99% fraction of initial metal atoms (FIMA) being fissioned are envisioned. This paper discusses initial calculations of the thermal behavior of spent LIFE fuel following completion of operation in the LIFE Engine [2]. The three time periods of interest for thermal calculations are during interim storage (probably at the LIFE Engine site), during the preclosure operational period of a geologic repository, and after closure of the repository.

Three parts of the 1991 book `Oklo: reacteurs nucleaires fossiles. Etude physique` have been translated in this report. The chapters bear the titles `Study of criticality`(45 p.), `Some problems with the overall functioning of the reactor zones`(45 p.) and `Conclusions` (15 p.), respectively.

Measurements of neutron yields and their angular distributions in coincidence with fission fragments produced in antiproton annihilation at rest in a natural uranium target have been carried out Low Energy Antiproton Ring (LEAR) at CERN. A total of 16.3±0.9 neutrons per annihilation have been found, distributed among direct knockout (27%), evaporation (21%), and fission (52%) processes. Angular distributions show that neutrons below approximately 5 MeV result entirely from moving fission fragments, and above approximately 12 MeV entirely from the excited, prescission nucleus. An estimate of the angular momentum of the excited fission fragment gives ?13?. We are able to account for all baryons produced in annihilation, including neutrons from this experiment and light charged nuclei found in another LEAR experiment, to within 4.5±2.5 of the initial 237 units in the initial state.

In this study, we present cytogenetic data regarding 66 Myzus persicae strains collected in different regions of Italy. Together with the most common 2n = 12 karyotype, the results showed different chromosomal rearrangements: 2n = 12 with A1-3 reciprocal translocation, 2n = 13 with A1-3 reciprocal translocation and A3 fission, 2n = 13 with A3 fission, 2n = 13 with A4 fission, 2n = 14 with X and A3 fissions. A 2n = 12-13 chromosomal mosaicism has also been observed. Chromosomal aberrations (and in particular all strains showing A1-3 reciprocal translocation) are especially frequent in strains collected on tobacco plants, and we suggest that a clastogenic effect of nicotine, further benefited by the holocentric nature of aphid chromosomes, could be at the basis of the observed phenomenon. PMID:22647317

The talk presents the results of an investigation of the main characteristics (mass and energy distributions of fission fragments and multiplicity of neutrons) of the fission of the nuclei of sup 2 sup 2 sup 0 Ra, sup 2 sup 2 sup 6 Th, sup 2 sup 5 sup 6 No, sup 2 sup 7 sup 0 Sg, sup 2 sup 8 sup 6 112 produced in reactions with ions of sup 1 sup 8 O, sup 2 sup 2 Ne and sup 4 sup 8 Ca at energies close to and essentially below the Coulomb barrier. The data obtained show that the form of the mass and energy distributions of the fission fragments of sup 2 sup 2 sup 6 Th and sup 2 sup 7 sup 0 Sg is accounted for by the multimodal nature of the fission. In addition, for sup 2 sup 2 sup 6 Th, a new phenomenon was established: there is a significant difference between the numbers of prescission neutrons for symmetric and asymmetric fission modes. It was found that, for the low-energy fission of the nucleus of sup 2 sup 8 sup 6 112, the mass distribution of the fragments is of a clear-cut asymmetric form, contrary to ...

The aim of this study is to use the hydrogeological and hydrochemical data from Oklo Natural Analogue to compare the outcome of two independent modelling approaches (HYTEC-2D and M3) which can be used to model natural conditions surrounding the reactor. HYTEC-2D represents a 2D, deterministic, transport and multi-solutes reactive coupled code developed at Ecole des Mines de Paris. M3 (named Multivariate Mixing and Mass balance) is a mathematical-statistical concept code developed for SKB. The M3 results are visualised using the Voxel Analyst code and the outcome of the uranium transport predictions are made from a performance assessment point of view. This exercise was in the beginning intended to represent a validation for M3, by comparing this statistic approach with the standard hydrodynamic - geochemical coupled code HYTEC-2D. It was realized that the codes complete each other and a better understanding of the geochemical studied system is obtained. Thus, M3 can relatively easily be used to calculate mixing portions and to identify sinks or sources of element concentrations that may exist in a geochemical system. This can help to address the reactions in the coupled code such as HYTEC-2D, to identify the hydrodynamic and hydrochemical system and to reduce the computation time. M3 shows the existence of the buffer around the reactor. No transport of uranium was indicated downstream the reactor. HYTEC-2D gives the same result in the case when we consider the existence of the redox buffer in the model. M3 shows an increase of the alkalinity in the reactor zone. The increase of the alkalinity was indicated by the M3 modelling to be associated with microbial decomposition of organic material which added reducing capacity to the system. The modelling result was supported by new results from the last field campaign, which included in-situ Eh measurements and microbial sampling and identification. The effects from the same process was indicated also by the HYTEC-2D predictions which show an increase of the pH in the reactor zone, due to the existence of the buffer. The two modelling approaches can be used to complete each other and to better understand the processes that can take place in nature. Thus, we can build confident tools which can be used to support performance assessment 16 refs, 16 figs

We have used the Oak Ridge Electron Linear Accelerator (ORELA) to measure neutron total and the fission cross sections of ²³³U in the energy range from 0.36 eV to ~700 keV. We report average fission and total cross sections. Also, we measured the neutron total cross sections of ²⁷Al and Natural chlorine as well as the capture cross section of Al over an energy range from 100 eV up to about 400 keV.

Four experiments are being conducted in the TREAT facility to investigate the behavior of fission products released from typical LWR fuel overheated to the point of catastrophic cladding degradation. Heatup and steam flow transients are used that simulate the conditions expected in operating power reactors undergoing various types of hypothetical severe accidents. The experiments are integral in nature and are aimed at the physicochemical characterization, near the point of origin, of the biologically important volatile fission products released early in such accidents. Detailed program objectives are discussed, a test matrix is presented, and the test apparatus is described. Pretest analysis and preliminary results are reported for the first test.

A series of laboratory annealing experiments are designed to evaluate the thermal annealing characteristics of fission tracks for a large number of glasses of varying chemical compositions. A comparative study of thermal stability has been made by comparing the calculated value of activation energy (Ea) for each glass. The present investigations suggest that fission tracks in glass samples with a high silica content are more resistant to annealing than those with a low silica content. In addition, the influence of chemical composition on etching characteristics for both synthetic and natural glasses has also been investigated.

Spontaneous fission provides a very natural production mechanism for neutron-rich nuclei. The large excesses of neutrons over protons in heavy systems near the valley of stability, coupled with the tendency for the charge-to-mass ratio to be preserved in the fission process, results in fission fragments that lie well to the neutron-rich side of stability. The excitation energy of the primary fragments produced at scission results in the evaporation of neutrons (Fig 1) to produce secondary fragments whose excitation energy is dissipated through {gamma}-ray emission. In the last few years the study of such {gamma} rays has produced a dramatic increase in the available information on the structure of neutron-rich nuclei. This renaissance in fission-fragment spectroscopy has been due primarily to the much increased resolving power of large arrays of germanium {gamma}-ray detectors such as Euroball and Gammasphere. Through the improved detection efficiency and high granularity, these arrays have made it possible to obtain reasonable rates for the detection of three coincident {gamma} rays out of the cascades of multiplicity ten that are typically produced in spontaneous fission. In addition to the determination of the energies of excited levels, developments in spectroscopic techniques have allowed for information to be deduced regarding the spins and parities [1,2] and lifetimes [3,4] of such states of extreme isospin. In this paper we present research that has been carried out using the Euroball and Eurogam arrays to detect {gamma} rays emitted from spontaneously fissioning {sup 248}Cm and {sup 252}Cf. The paper focuses on three sub-areas of current activity, namely, the measurement of yields of secondary fragment pairs, the measurement of state lifetimes at around spin 10, and recent measurements of g-factors of excited states in fission fragments.

Relatively neutron-rich fission-fragment secondary beams have been produced using a 20 MeV/u beam of {sup 197}Au{sup 29+} incident on a natural carbon target of areal density 47 mg/cm{sup 2} at the National Superconducting Cyclotron Laboratory at Michigan State University (PAC17-NSCL Expt. No. 94027). The reaction products are kinematically focused, resulting in good collection efficiency as a large superconducting solenoid at {theta} = 0{degrees} is used to collect and separate the fragments. The latter is configured for a 7 meter flight path with its focus on a stack of silicon detectors. {Delta}E, E{sub Total}, TOF, and B{rho} allow for the identification of the fragments. Analysis shows predominantly symmetric fission of Au nuclei with fragments concentrated about Z = 34 to 40 and A = 80 to 96. Well focused secondary, neutron-rich fission fragment beams of 10{sup 2}/s to 10{sup 3}/s appear feasible.

The principles for structuring the channel, intended for measuring the thermal neutrons flux in the reactor control system are set forth. The channel contains a fission chambers a set of boron-containing gas-filled ionization chambers and an electron apparatus unit, providing for the pulse fission counting rate up to approx 1 x 10 sup 6 s sup - sup 1. It is shown that the wide linear range is achieved in the channels, provided with the fission chamber and gas-filled ionization chamber, the electrodes whereof are coated with a natural boron. The channel makes it possible to measure the thermal neutrons flux within the density interval from 0.5 up to 5 x 10 sup 1 sup 1 cm sup - sup 2 s sup - sup 1

Volume 4 of this series contains information regarding heavy ions and low-energy nuclear fission; fission of highly excited and rotating nuclei; nuclear fission and nucleus-nucleus collisions with salient examples of large-scale collective nuclear motion; the nuclear properties of the transuranium elements (TUEs); chemical and physical properties of the TUEs; predictions for the properties of superheavy elements; and attempts to synthesize superheavy elements with heavy-ion beams. Other topics include the possibility of producing superheavy elements by nucleosynthesis; the search for superheavy elements in galactic cosmic rays; the physical and chemical aspects of the search for superheavy elements in nature; and nuclear physics in relativistic heavy-ion collisions. Theoretical models are presented for relativistic heavy-ion reactions.

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

A conservative modeling and analysis were attempted to explain the presence of nonradioactive fission-like products with nonnatural isotopic ratios observed in some D2O/Pd electrolysis experiments. The collective deformation of a Pd nucleus by multiphoton E1 resonance absorption in a dynamic PdDx lattice was assumed to induce low-energy photofissions via the selective scission channels within the lowest band (11-20 MeV) of channel-dependent fission barriers. Values of channel dependent fission barriers were calculated by using liquid drop model potentials for Pd isotopes. Fission products were analyzed in detail. Major fission products (FPs) are stable isotopes and the isotopic ratios of FP elements are very different from those of natural abundances. The present theoretical results have shown good agreement with the experimental data of Mizuno et al. [Denki Kagaku 64 (1996) 1660] and others in terms of Z-distribution, mass distribution and isotopic ratios. Selective channel photofissions with positive Q-values are possible for A > 90 nuclei, which may provide us with a clean method for the incineration for the radio isotope (RI) waste of nuclear plants.   

In spent nuclear fuel (SNF), the micron- to submicron-sized epsilon phase (Mo-Ru-Pd-Tc-Rh) is an important host of {sup 99}Tc which has a long half life (2.13 x 10{sup 5} years) and can be an important contributor to dose in safety assessments of nuclear waste repositories. In addition, Tc is predominantly present as TcO{sub 4}{sup -} under oxidizing conditions at wide range of pH, weakly adsorbed onto mineral surfaces, and unlikely to be incorporated into alteration uranyl minerals. In the Oklo natural reactor (2.0 Ga), essentially all of the {sup 99}Tc has decayed to {sup 99}Ru. Thus, this study focuses on Ru and the other metals of the epsilon phase in order to investigate the occurrence and the fate of the epsilon phase during the corrosion of this natural SNF. Samples from reactor zone (RZ)-10 (836, 819, 687); from RZ-13 (864, 910); were investigated using TEM (transmission electron microscopy). Within the UO{sub 2} matrix, a Bi-Pd particle (40-60 nm), fioodite, PdBi{sub 2}, was observed with trace amounts of As, Fe, and Te surrounded by an amorphous Pb-rich area. (Pd,Rh){sub 2}As, palladodymite or rhodarsenide, was observed (400-500 nm in size). Ruthenarsinite, (Ru,Ni)As, was identified in most samples: with a representative composition of As, 59.9: Co, 2.5: Ni, 5.2; Ru, 18.6; Rh, 8.4; Pd, 3.1; Sb, 2.4 in atomic percent. The particles diameters are a few hundred nanometers and, in most cases, surrounded by a Pb-rich phase (400-500 nm). Typically, the ruthenarsenite does not occur as single particle but an aggregate of {approx}200 nm-sized particles. Some Ru-particles revealed a complex phase separation within the grain such as a Ru-particle (600-700 nm) with Pb at the core of the particle and enrichment of Ni, Co, and As at the rim. Some ruthenarsenite crystals were embedded in chlorite immediately adjacent to uraninite. A few particles were still coated by Pb. These results suggest a history for the epsilon phases: (1) The original epsilon phase was transformed to, in most cases, ruthenarsenite. (2) All Mo and most of the Tc were released from the epsilon phase. Some portion of the other metals was also leached and provided a space for a precipitation of PbS between the ruthenarsenite and uraninite. (3) Once the uraninite matrix dissolved, the epsilon particles were released and sometimes captured within adjacent alteration minerals.

Transmutation of fission products /sup 137/Cs and /sup 90/Sr using the neutron produced by high energy proton collision with heavy nuclei were investigated. Because of the small thermal neutron cross section for (n,..gamma..) reaction of /sup 137/Cs (0.1 barn), a high neutron flux of 10/sup 17/ n/cm/sup 2/ sec is required to transmute /sup 137/Cs at a rate ten times faster than the natural decay. This range of high flux is attainable in the spallation reaction of high energy proton beam interact with liquid Pb target. The neutronic calculation by using NMTC, HIST3D, EPR, TAPEMAKER and ANISN codes indicates that the spallation neutron can transmute 222 kg /sup 137/Cs and 155 kg /sup 90/Sr fission products per year (at a rate of 10 and 30 times faster than their natural decay rate) by running a 300 mA, 1.5 GeV proton beam. Thus, if we transmute these fission products, just after a burning cycle, this accelerator can transmute these fission products produced in five or six 1000 MWe power plants.

A concept for detecting the presence of special nuclear material ({sup 235}U or {sup 239}Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their {beta}-delayed neutron emission or {beta}-delayed high-energy {gamma}-radiation between beam pulses provide the detection signature. Fission product {beta}-delayed {gamma}-rays above 3 MeV are nearly ten times more abundant than {beta}-delayed neutrons and are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified. An important goal in the US is the detection of nuclear weapons or special nuclear material (SNM) concealed in intermodal cargo containers. This must be done with high detection probability, low false alarm rates, and without impeding commerce, i.e. about one minute for an inspection. The concept for inspection has been described before and its components are now being evaluated. While normal radiations emitted from plutonium may allow its detection, the majority of {sup 235}U {gamma} ray emission is at 186 keV, is readily attenuated by cargo, and thus not a reliable detection signature for passive detection. Delayed neutron detection following a neutron or photon beam pulse has been used successfully to detect lightly or unshielded SNM targets. While delayed neutrons can be easily distinguished from beam neutrons they have relatively low yield in fission, approximately 0.008 per fission in {sup 239}Pu and 0.017 per fission in {sup 235}U, and are rapidly attenuated in hydrogenous materials making that technique unreliable when challenged by thick hydrogenous cargo overburden. They propose detection of {beta}-delayed high-energy {gamma} radiation as a more robust signature characteristic of SNM.

A natural apatite crystal from Durango, Mexico was studied with electron spin resonance (ESR), thermoluminescence (TL) and fission track (FT) methods. Isochronal annealing experiments revealed that the thermal stability of the ESR signal related to a hole centre, O{sup -} substituting F{sup -}, is comparable to that of fission tracks. The decay of the hole centres probably corresponds to the 380{sup o}C peak of the TL glow curve. The stability of the ESR signal enhanced by gamma irradiation is lower than that of the natural ESR signal. Accordingly, the ESR ages estimated by the additive dose method gave much younger ages than the FT age. The spatial distribution of the signal intensity was obtained by microwave scanning ESR microscope. (author).

A spallation neutron source was modeled using a high energy proton accelerator for transmutation of 239Pu, minor actinides 237Np, 241Am and long-lived fission products 99Tc, 129I, which are created from the operation of nuclear power reactors for the production of electricity.The acceleration driven system (ADS) is composed of a natural lead target, beam window, subcritical core, reflector, and structural material. The neutrons are produced by the spallation reaction of protons from a high intensity linear accelerator in the spallation target, and the fission reaction in the core. It is used a hexagonal lattice for the waste and fuel assemblies. The system is driven by a 1GeV, 10mA proton beam incident on a natural lead cylindrical target. The protons were uniformly distributed across the ...

Historical review of the dating methods developed at the CEA in France since the 50`s: the lead-lead method developed for uranium deposits, {sup 14}C dating since 1956, potassium-argon methods, fission trace measurement, thermoluminescence measurement, uranium-thorium balance dating, methods for evaluating and dating natural reactors, deuterium and tritium methods for dating ice caps, which led to the development of detritiation techniques for heavy water

All of the source term calculations performed by the USNRC in conjunction with NUREG-1150 do not take into consideration the existence of natural circulation in the reactor system during the transport of the fission products from the core to the containment. In addition, long-term (post vessel failure) fission product behavior is not analyzed. The SOURCE TERM CODE package as presently configured does not account for these phenomena. The importance of the inclusion of this phenomenology in the source term calculation was examined by the authors and is reported. The authors performed code modifications to include the natural circulation and long-term fission product behavior phenomena in the source term calculation. Comparison of calculations without and with the inclusion of the natural circulation is provided and the effect of the inclusion discussed. The need for the inclusion in the long term analyses is also discussed. The authors show that these phenomena are important aspects of the source term calculation and their lack in the SOURCE TERM CODE PACKAGE is an important shortcoming of the methodology.

The mass spectrometric determination of minor abundant isotopes, {sup 234}U and {sup 236}U in naturally occurring uranium materials requires instruments of high abundance sensitivity and the use of highly sensitive detection systems. In this study the thermal ionisation mass spectrometer Finnigan MAT 262RPQ was used. It was equipped with 6 Faraday cups and a Secondary Electron Multiplier (SEM), which was operated in pulse counting mode for the detection of extremely low ion currents. The dynamic measurement range was increased considerably combining these two different detectors. The instrument calibration was performed carefully. The linearity of each detector, the deadtime of the ion counting detector, the detector normalisation factor, the baseline of each detector and the mass discrimination in the ion source were checked and optimised. A measurement technique based on the combination of a Gas Source Mass Spectrometry (GSMS) and a Thermal Ionisation Mass Spectrometry (TIMS) was developed for the accurate determination of isotopic composition in naturally occurring uranium materials. Because the expected ratio of n({sup 234}U)/n({sup 238}U) exceeded the dynamic measurement range of the Faraday detectors of the TIMS instrument, an experimental design using a combination of two detectors was developed. The n({sup 234}U)/n({sup 235}U) and n({sup 236}U)/n({sup 235}U) ratios were determined using ion counting in combination with the decelerating device. The n({sup 235}U)/n({sup 238}U) ratio was determined by the Faraday detector. This experimental design allowed the detector cross calibration to be circumvented. Precisions of less than 1 percent for the n({sup 234}U)/n({sup 235}U) ratios and 5-25 percent for the n({sup 236}U)/n({sup 235}U) ratios were achieved. The purpose of the study was to establish a register of isotopic signatures for natural uranium materials. The amount ratio, and isotopic composition of 18 ore concentrates, collected by the International Atomic Energy Agency (IAEA) from uranium milling and mining facilities (Australia, Canada, Gabon, Namibia, Czech Republic, France), were determined. These signatures form the basic register. The isotopic signatures are feasible in identifying the sample origin and in separating naturally occurring or background contributions from local anthropogenic sources. With the comparison of fingerprints of unknown samples to the isotopic fingerprints of samples of known origin, it is possible to trace back unknown samples to their origin or at least to exclude suspected origins in the case of non-identity of fingerprints. This was successfully demonstrated with a number of samples of unknown origin, which were measured during the study. Generally, no significant variability was observed in the n({sup 235}U)/n({sup 238}U) ratios except in the well known case of samples originating from Oklo (Gabon). Small variations in the n({sup 234}U)/n({sup 238}U) amount ratios were understood from the radiochemical mother-daughter relationship of the two isotopes involved. The detection limit for the n({sup 236}U)/n({sup 235}U) amount ratio (DL = 0.000001) was derived from blank measurements. The limit of quantitation 0.000003 was calculated as LQ=3DL. When the measured ratio exceeded the quantitation limit, the presence of {sup 236}U is explained. (orig.) 99 refs.

Samples from Oklo Reactor zone-9 (ORZ-9) have been analyzed for the isotopic abundances of Nd, Ce, Ru, and Mo. Interpretation of the Nd data has begun as part of the effort to reconstruct the operating parameters of the reactor. The study of ORZ-9 and the peripheral rocks is being enhanced by additional analytical capabilities. A procedure was developed to measure uranium isotopic ratios with high precision. This new method was used for the analysis of rocks peripheral to ORZ-9. Two rocks containing relatively small quantities of uranium were depleted in /sup 235/U. The result demonstrates that small quantities of uranium were removed from the reactor zone and redistributed over distances of several tens of meters. Procedures are being designed to make high precision measurements of the relative abundances of barium isotopes. They will be used as part of a study of the transport of alkali and alkaline earth elements at Oklo. Samples from distances up to 300 meters from the known mineralized area at Oklo have been selected and prepared in an effort to identify element transport paths over longer distances. A sample from the Athabasca sandstone, overlying the uranium ores at Key Lake, and another from the transition zone at the unconfromity between the sandstone and the basement were subjected to sequential leaches designed to preferentially dissolve specific minerals. Lead isotopic analyses on the leachs yielded two sets of data, which indicate the loss of uranium or the addition of lead from a radiogenic source early in the geologic history of the rocks.

With a total production of more than 6000 t of uranium in 1998, Cogema is one of the first mining uranium producer (20% of the worldwide production). This short paper summarizes Cogema's mining activities worldwide. Cogema's reserves are located in six different countries: France (Forez, Morvan, Limousin, Vendee, Herault), Niger (Air, Akouta), Gabon (Mounana, Oklo, Mikouloungou), Canada (Cluff Lake, McLean Lake, Midwest, Cigar Lake, Mc Arthur river), USA (Wyoming), Australia (Koongarra, Ranger, Jabiluka) and deposits are exploited by Cogema in five of these countries (not in Australia). Prospecting work is carried on in Canada (Kiggavik, Sissons), Australia (Arnhem Land) and central Asia (Kazakhstan, Mongolia, Uzbekistan). (J.S.)

The S-Area Defense Waste Processing Facility (DWPF) can process all of the high level radioactive wastes currently stored at the Savannah River Site with negligible risk of nuclear criticality. The characteristics which make the DWPF critically safe are: (1) abundance of neutron absorbers in the waste feeds; (2) and low concentration of fissionable material. This report documents the criticality safety arguments for the S-Area DWPF process as required by DOE orders to characterize and to justify the low potential for criticality. It documents that the nature of the waste feeds and the nature of the DWPF process chemistry preclude criticality.

Confocal scanning laser microscopy (SLM) is a technique that offers geologists a new way of studying structures in minerals at the submicrometre level. As an example we show how the non-destructive nature of confocal SLM can be used to measure and count fission tracks (line defects formed by the spontaneous fission of [sup 238]U) in the uranium-bearing mineral apatite, and to provide information about the geometry and crystallographic orientation of fluid inclusions trapped inside apatite grains during crystallization. The technique also provides a means of studying the internal geometry of chemical zonation in minerals. The digitized nature of the SLM images makes them amenable to a variety of image analysis techniques, and we show how image analysis can be used to measure fission tracks in mica sheets and provide crude estimates of track dip. Finally, using a chemically etched mica sheet we show how confocal SLM can be used to provide a detailed near-surface (1-5 [mu]m) analysis of geological materials. (author).

This paper provides description of a nuclear reactor concept aimed towards a radical safety enhancement, an increased proliferation resistance, as well as a realisation of a "long-lived waste free" NP development. It emphasizes the achievement of considerable reduction ("by design") of residual actinides in the waste streams and of the most hazardous long-lived fission products. It allows to implement only small volume of repositories for the radioactive waste (mostly fission products) and to postpone the technically arduous problems of a large scale disposal of the long-lived wastes until the next millennium, i.e. up to the exhaustion of the fertile natural resources and/or the emergence of more effective technologies of nuclide separation/transmutation.A thorough incineration/transmutati...

The thermodynamical equilibrium hypothesis has been applied to the tests ORNL (Oak Ridge National Laboratory) VI-3 and VI-5 (Vertical Induction heated), to study fission release products. Irradiated fuel samples were heated under oxidizing (VI-3) and reducing (VI-5) controlled atmosphere. They can be compared to the Vercors 4 and 5 tests. The thermodynamical calculus results of barium and ruthenium fission products release, agree in the lump with measures, whatever the gas phase nature. The strontium release calculus under reducing atmosphere, is largely upper than those measured and correctly evaluated under oxidizing atmosphere. These results confirm those obtained for Vercors 4 and 5 and show the interest of the select model. (A.L.B.)

We study primary and secondary reactions induced by 600 MeV proton beams in monolithic cylindrical targets made of natural tungsten and uranium by using Monte Carlo simulations with the Geant4 toolkit [1-3]. Bertini intranuclear cascade model, Binary cascade model and IntraNuclear Cascade Liège (INCL) with ABLA model [4] were used as calculational options to describe nuclear reactions. Fission cross sections, neutron multiplicity and mass distributions of fragments for 238U fission induced by 25.6 and 62.9 MeV protons are calculated and compared to recent experimental data [5]. Time distributions of neutron leakage from the targets and heat depositions are calculated. This project is supported by Siemens Corporate Technology.

This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10{sup {minus}5 }and 10{sup {minus}7} per reactor year for LCP15 and LPP9, respectively.

Largely in anticipation of a possible nuclear renaissance, there has been an enthusiastic renewal of interest in the fusion-fission hybrid concept, driven primarily by some members of the fusion community. A fusion-fission hybrid consists of a neutron-producing fusion core surrounded by a fission blanket. Hybrids are of interest because of their potential to address the main long-term sustainability issues related to nuclear power: fuel supply, energy production, and waste management. As a result of this renewed interest, the U.S. Department of Energy (DOE), with the participation of the Office of Fusion Energy Sciences (OFES), Office of Nuclear Energy (NE), and National Nuclear Security Administration (NNSA), organized a three-day workshop in Gaithersburg, Maryland, from September 30 through October 2, 2009. Participants identified several goals. At the highest level, it was recognized that DOE does not currently support any R&D in the area of fusion-fission hybrids. The question to be addressed was whether or not hybrids offer sufficient promise to motivate DOE to initiate an R&D program in this area. At the next level, the workshop participants were asked to define the research needs and resources required to move the fusion-fission concept forward. The answer to the high-level question was given in two ways. On the one hand, when viewed as a standalone concept, the fusion-fission hybrid does indeed offer the promise of being able to address the sustainability issues associated with conventional nuclear power. On the other hand, when participants were asked whether these hybrid solutions are potentially more attractive than contemplated pure fission solutions (that is, fast burners and fast breeders), there was general consensus that this question could not be quantitatively answered based on the known technical information. Pure fission solutions are based largely on existing both fusion and nuclear technology, thereby prohibiting a fair side-by-side comparison. Another important issue addressed at the conference was the time scale on which long-term sustainability issues must be solved. There was a wide diversity of opinion and no consensus was possible. One group, primarily composed of members of the fission community, argued that the present strategies with respect to waste management (on-site storage) and fuel supply (from natural uranium) would suffice for at least 50 years, with the main short-term problem being the economics of light water reactors (LWRs). Many from the fusion community believed that the problems, particularly waste management, were of a more urgent nature and that we needed to address them sooner rather than later. There was rigorous debate on all the issues before, during, and after the workshop. Based on this debate, the workshop participants developed a set of high-level Findings and Research Needs and a companion set of Technical Findings and Research Needs. In the context of the Executive Summary it is sufficient to focus on the high-level findings which are summarized.

Characteristics of fission and NTP indicate that useful ... ing of the energy deposition and heat transfer processes ..... Fusion. Fission. Nuclear Fission Process. 180 MeV prompt useful energy (plus 10 MeV ... Last significant criticality accident in ...

Independent scientific work has unambiguously demonstrated life to be present in most deep geological formations investigated, down to depths of several kilometres. Microbial processes have consequently become an integral part of the performance safety assessment of high-level radioactive waste (HLW) repositories. This report presents the research record from the last decade of the microbiology research programme of the Swedish Nuclear Fuel and Waste Management Company (SKB) and gives current perspectives of microbial processes in HLW disposal. The goal of the microbiology programme is to understand how microbes may interact with the performance of a future HLW repository. First, for those who are not so familiar with microbes and their ways of living, the concept of 'microbe' is briefly defined. Then, the main characteristics of recognised microbial assemblage and microbial growth, activity and survival are given. The main part of the report summarises data collected during the research period of 1987-1999 and interpretations of these data. Short summaries introduce the research tasks, followed by reviews of the results and insight gained. Sulphate-reducing bacteria (SRB) produce sulphide and have commonly been observed in groundwater environments typical of Swedish HLW repositories. Consequently, the potential for sulphide corrosion of the copper canisters surrounding the HLW must be considered. The interface between the copper canister and the buffer is of special concern. Despite the fact that nowhere are the environmental constraints for life as strong as here, it has been suggested that SRB could survive and locally produce sulphide in concentrations large enough to cause damage to the canister. Experiments conducted thus far have indicated the opposite. Early studies in the research programme revealed previously unknown microbial ecosystems in igneous rock aquifers at depths exceeding 1000 m. This discovery triggered a thorough exploration of the subterranean biosphere in the aquifers of the Fennoscandian Shield. Microbial processes at repository depths will have several important influences on repository performance. Some identified processes are the production of sulphide, carbon dioxide, organic carbon and methane, and the reduction of oxygen. The repository performance must be predicted for very long times. Natural analogues therefore become invaluable. Time-related processes of radionuclide migration have been studied at analogue sites that have been evolving over very long time periods. High pH conditions occur in Maqarin, Jordan; fuel-related processes have evolved at the natural reactors in Oklo, Gabon, and uranium migration processes have developed around the uranium body of Palmottu, in Finland. The majority of the radionuclides are metals. The transport, chemical speciation, and ultimate fate of dissolved metals in aqueous systems are controlled to a large extent by reactions that occur at solid surfaces. Bacteria are at least as widely distributed and probably as reactive as many inorganic solids in aqueous systems. The behaviour of bacteria as geochemically reactive solids can be inferred from extensive research documenting their performance as sorbents of dissolved metals, and as nucleation templates for a wide range of authigenic minerals. They may consequently play a significant role in radionuclide retention and transport processes.

The following report provides a rich resource of information for exploring fuel cycle characteristics. The most noteworthy trends can be traced back to the utilization efficiency of natural uranium resources. By definition, complete uranium utilization occurs only when all of the natural uranium resource can be introduced into the nuclear reactor long enough for all of it to undergo fission. Achieving near complete uranium utilization requires technologies that can achieve full recycle or at least nearly full recycle of the initial natural uranium consumed from the Earth. Greater than 99% of all natural uranium is fertile, and thus is not conducive to fission. This fact requires the fuel cycle to convert large quantities of non-fissile material into fissile transuranics. Step increases in waste benefits are closely related to the step increase in uranium utilization going from non-breeding fuel cycles to breeding fuel cycles. The amount of mass requiring a disposal path is tightly coupled to the quantity of actinides in the waste stream. Complete uranium utilization by definition means that zero (practically, near zero) actinide mass is present in the waste stream. Therefore, fuel cycles with complete (uranium and transuranic) recycle discharge predominately fission products with some actinide process losses. Fuel cycles without complete recycle discharge a much more massive waste stream because only a fraction of the initial actinide mass is burned prior to disposal. In a nuclear growth scenario, the relevant acceptable frequency for core damage events in nuclear reactors is inversely proportional to the number of reactors deployed in a fuel cycle. For ten times the reactors in a fleet, it should be expected that the fleet-average core damage frequency be decreased by a factor of ten. The relevant proliferation resistance of a fuel cycle system is enhanced with: decreasing reliance on domestic fuel cycle services, decreasing adaptability for technology misuse, enablement of material accountability, and decreasing material attractiveness.

Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m/sup 3/ and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater than 0.6 b.a. old and if they contained sufficient water but lacked neutron-absorbing impurities. While uranium deposits of northern Canada and northern Australia have received most attention, ore samples have been obtained from the following worldwide locations: the Shinkolobwe and Katanga regions of Zaire; Southwest Africa; Rio Grande do Norte, Brazil; the Jabiluka, Nabarlek, Koongarra, Ranger, and El Sharana ore bodies of the Northern Territory, Australia; the Beaverlodge, Maurice Bay, Key Lake, Cluff Lake, and Rabbit Lake ore bodies and the Great Bear Lake region, Canada. The ore samples were tested for isotopic variations in uranium, neodymium, samarium, and ruthenium which would indicate natural fission. Isotopic anomalies were not detected. Criticality was not achieved in these deposits because they did not have sufficient /sup 235/U content (a function of age and total uranium content) and/or because they had significant impurities and insufficient moderation. A uranium mill monitoring technique has been considered where the ''yellowcake'' output from appropriate mills would be monitored for isotopic alterations indicative of the exhumation and processing of a natural reactor.

The first detectable activities of radioiodine (131)I, and radiocaesium (134)Cs and (137)Cs in the air over Central Poland were measured in dust samples collected by the ASS-500 station in the period of 21(st) to 24(th) of March, 2011. However, the highest activity of both fission products, (131)I and (137)Cs: 8.3 mBq m(-3) and 0.75 mBq m(-3), respectively, were obtained in the samples collected on 30(th) March, i.e.?18 days after the beginning of the fission products' discharge from the damaged units of the Fukushima Daiichi Nuclear Power Plant. The simultaneously determined corrected aerosol residence time for the same samples by (210)Pb/(210)Bi and (210)Pb/(210)Po methods was equal to 10 days. Additionally, on the basis of the activity ratio of two other natural cosmogenic radionuclides, (7)Be and (22)Na in these aerosol samples, it was possible to estimate the aerosol residence time at ?150 days for the solid particles coming from the stratospheric fallout. These data, as well as the differences in the activity size distribution of (7)Be and (131)I in the air particulate matter, show, in contrast to the Chernobyl discharge, a negligible input of stratospheric transport of Fukushima-released fission products. PMID:22481111

Mobile detection of kilogram quantities of special nuclear materials (SNM) during maritime transportation is a challenging problem for the U.S. Department of Homeland Security. Counting neutrons emitted by the SNM and partitioning them from background neutrons of multiple origins is the most effective passive means of detecting the SNM. Unfortunately, neutron detection, counting, and partitioning in a maritime environment is complex due to the presence of spallation neutrons (commonly known as "ship effect") and to the complicated nature of the neutron scattering in that environment. This work studied the possibilities of building a prototype neutron detector using boron- 10 (10B) as the converter in a novel form factor called "straws" that would address the above problem by examining multiplicity distributions of neutrons originating from a fissioning source. Currently, commercially manufactured fission meters (FM) are available that separate cosmic neutrons from non-cosmic neutrons and quantitatively determine the strength of a fissioning source; however, these FMs use 3He, which is becoming increasingly difficult to procure; also the size and weight of a commercial FM is not conducive to manual neutron detection operations in a maritime environment. The current project may provide a near-term solution to the crisis that has arisen from the global scarcity of 3He by offering a viable alternative to the FM. The prototype detector provides a large-area, efficient, lightweight, more granular neutron responsive detection surface (to facilitate imaging) to ease the application of the new FMs.

Analysis of the AFIP-6 experiment is summarized in this report in order to determine the cause of gaseous fission product release observed during irradiation. During the irradiation, a series of small fission product releases were observed. In order to limit the potential for primary coolant contamination, the operating cycle was terminated and the AFIP-6 experiment was removed for examination. Both in-canal and post-irradiation examination revealed the presence of an unusually thick oxide layer and discrete surface blisters on the fuel plates. These blisters were the likely cause of fission product release. Subsequent detailed thermal hydraulic analysis of the experiment indicated that the combination of the high operating power and test vehicle configuration led to high nominal operating temperatures for the fuel plates. This elevated temperature led to accelerated surface corrosion and eventually spallation of the fuel plate cladding. The thermal insulating nature of this corrosion layer led to significantly elevated fuel meat temperatures that induced blistering. Analysis was performed to validate a corrosion rate model and criteria for onset of spallation type surface corrosion were determined. The corrosion rate model will be used to estimate the oxide thickness anticipated for experiments in the future. The margin to the spallation threshold will then be used to project the experiment performance.

The origins and near-surface distributions of the approximately 250 known uranium and/or thorium minerals elucidate principles of mineral evolution. This history can be divided into four phases. The first, from ~4.5 to 3.5 Ga, involved successive concentrations of uranium and thorium from their initial uniform trace distribution into magmatic-related fluids from which the first U4+ and Th4+ minerals, uraninite (UO2), thorianite (ThO2) and coffinite (USiO4), precipitated in the crust. The second period, from ~3.5 to 2.2 Ga, saw the formation of large low-grade concentrations of detrital uraninite (containing several weight percent Th) in the Witwatersrand-type quartz-pebble conglomerates deposited in a highly anoxic fluvial environment. Abiotic alteration of uraninite and coffinite, including radiolysis and auto-oxidation caused by radioactive decay and the formation of helium from alpha particles, may have resulted in the formation of a limited suite of uranyl oxide-hydroxides. Earth’s third phase of uranium mineral evolution, during which most known U minerals first precipitated from reactions of soluble uranyl (U6+O2)2+ complexes, followed the Great Oxidation Event (GOE) at ~2.2 Ga and thus was mediated indirectly by biologic activity. Most uraninite deposited during this phase was low in Th and precipitated from saline and oxidizing hydrothermal solutions (100 to 300°C) transporting (UO2)2+-chloride complexes. Examples include the unconformity- and vein-type U deposits (Australia and Canada) and the unique Oklo natural nuclear reactors in Gabon. The onset of hydrothermal transport of (UO2)2+ complexes in the upper crust may reflect the availability of CaSO4-bearing evaporites after the GOE. During this phase, most uranyl minerals would have been able to form in the O2-bearing near-surface environment for the first time through weathering processes. The fourth phase of uranium mineralization began approximately 400 million years ago, as the rise of land plants led to non-marine organic-rich sediments that promoted new sandstone-type ore deposits. The modes of accumulation and even the compositions of uraninite, as well as the multiple oxidation states of U (4+, 5+, and 6+), are a sensitive indicator of global redox conditions. In contrast, the behavior of thorium, which has only a single oxidation state (4+) that has a very low solubility in the absence of aqueous F-complexes, cannot reflect changing redox conditions. Geochemical concentration of Th relative to U at high temperatures is therefore limited to special magmatic-related environments, where U4+ is preferentially removed by chloride or carbonate complexes, and at low temperatures by mineral surface reactions. The near-surface mineralogy of uranium and thorium provide a measure of a planet’s geotectonic and geobiological history. In the absence of extensive magmatic-related fluid reworking of the crust and upper mantle, uranium and thorium will not become sufficiently concentrated to form their own minerals or ore deposits. Furthermore, in the absence of surface oxidation, all but a handful of the known uranium minerals are unlikely to form.

This paper presents an overview of radiolysis concerns with regard to water shields for fission surface power. A review of the radiolysis process is presented and key parameters and trends are identified. From this understanding of the radiolytic decomposition of water, shield pressurization and corrosion are identified as the primary concerns. Existing experimental and modeling data addressing concerns are summarized. It was found that radiolysis of pure water in a closed volume results in minimal, if any net decomposition, and therefore reduces the potential for shield pressurization and corrosion. With the space program focus m emphasize more on permanent return to the Moon and eventually manned exploration of Mars, there has been a renewed look at fission power to meet the difficult technical & design challenges associated with this effort. This is due to the ability of fission power to provide a power rich environment that is insensitive to solar intensity and related aspects such as duration of night, dusty environments, and distance from the sun, etc. One critical aspect in the utilization of fission power for these applications of manned exploration is shielding. Although not typically considered for space applications, water shields have been identified as one potential option due to benefits in mass savings and reduced development cost and technical risk (Poston, 2006). However, the water shield option requires demonstration of its ability to meet key technical challenges including such things as adequate natural circulation for thermal management and capability for operational periods up to 8 years. Thermal management concerns have begun to be addressed and are not expected to be a problem (Pearson, 2007). One significant concern remaining is the ability to maintain the shield integrity through its operational lifetime. Shield integrity could be compromised through shield pressurization and corrosion resulting from the radiolytic decomposition of water.

An expression for the fission width of actinide nuclei is obtained, using a model in which vibrations in the fission degree of freedom are considered to be doorway states for fission. The coupling of these states to compound states in both potential wells is considered explicitly. Comparisons to other models for the calculations of the fission widths are made.

Coulomb fission of {sup 238}U has been studied in the interaction of a 24.3 MeV/nucleon U beam with an Au target. A novel experimental approach is followed, allowing to isolate the Coulomb fission from the nuclear fission on an event by event basis. The Z distribution of the fragments is studied for both fission processes. (author). 31 refs.

This patent describes a solid tag material for identifying a nuclear reactor component when placed within the component comprising: solid materials which contain at least one non-gaseous element that undergoes nuclear transmutation into one or more detectable, identifiable, and measurable tag gases on irradiation in a nuclear reactor. The tag gases generated are stable, are not fission products, and are generated in predetermined proportions different from their natural occurrence. Upon the failure of a component the tag gases can be detected in the reactor cover gas and the failed component identified according to the tag gas composition.

The long-term (> 1000 years) hazards of high-level wastes can be reduced substantially by practising waste-actinide partitioning-transmutation. The waste-actinide transmutation performance of a uranium hexafluoride actinide transmutation reactor (UHATR) is investigated. Using mostly present-day and near-term technology, a preliminary UHATR design is established. Because of the gaseous nature of the fuel, very high neutron fluxes are obtained. Compared with an LWR, the average blanket thermal flux of this UHATR is about 10-30 times higher, leading to a 15-fold improvement in the percentage of actinides fissioned per year of irradiation.

We report results of air monitoring started due to the recent natural catastrophe on March 11, 2011 in Japan and the severe ensuing damage to the Fukushima nuclear reactor complex. On March 17-18, 2011 we detected the first arrival of the airborne fission products 131-I, 132-I, 132-Te, 134-Cs, and 137-Cs in Seattle, WA, USA, by identifying their characteristic gamma rays using a germanium detector. The highest detected activity to date is <~32 mBq/m^3 of 131-I.

A procedure was developed for separating rubidium from irradiated aluminum encapsulated uranium. The separations procedure produces a final ultra-high purity RbCl product for subsequent high performance mass spectrometric analysis. The procedure involves first removing most of the macro-components and fission products by strong base anion exchange using, first, concentrated HCl, then oxalic acid media and second, selectively separating rubidium from alkaline-earth ions and other alkali-metal ions, including cesium, using Bio-Rex-40 cation-exchange resin. The resultant RbCl is then put through a final vacuum sublimation step. Ultra-pure reagents and specially clean glassware are used throughout the procedure to minimize contamination by naturally-occurring rubidium.

This paper presents science potential of a deep ocean antineutrino observatory being developed at Hawaii. The observatory design allows for relocation from one site to another. Positioning the observatory some 60 km distant from a nuclear reactor complex enables precision measurement of neutrino mixing parameters, leading to a determination of neutrino mass hierarchy and Formula Not Shown . At a mid-Pacific location the observatory measures the flux and ratio of uranium and thorium decay neutrinos from earths mantle and performs a sensitive search for a hypothetical natural fission reactor in earths core. A subsequent deployment at another mid-ocean location would test lateral heterogeneity of uranium and thorium in earths mantle.

This paper presents science potential of a deep ocean antineutrino observatory being developed at Hawaii. The observatory design allows for relocation from one site to another. Positioning the observatory some 60 km distant from a nuclear reactor complex enables precision measurement of neutrino mixing parameters, leading to a determination of neutrino mass hierarchy and ?. At a mid-Pacific location the observatory measures the flux and ratio of uranium and thorium decay neutrinos from earth's mantle and performs a sensitive search for a hypothetical natural fission reactor in earth's core. A subsequent deployment at another mid-ocean location would test lateral heterogeneity of uranium and thorium in earth's mantle.

{sup 197}Au(800 A MeV)-on-proton collisions are used to investigate the fission dynamics at high excitation energy. The kinematic properties together with the isotopic identification of the fission fragments allow to determine the mass, charge and excitation energy of the fissioning nucleus at saddle. The comparison of these observables and the measured total fission cross section with model calculations evidences a clear hindrance of fission at high excitation energy that can be explained in terms of nuclear dissipation. Assuming a statistical evaporation for other de-excitation channels than fission, an estimated value of the transient time of fission of (3 {+-} 1) . 10{sup -21} s is obtained. (orig.)

The degrees of the deformation of nuclei at the scission points in the symmetric and asymmetric fission are experimentally determined. They are found to be constant, respectively, in a wide range of Af among the asymmetric fission and among the asymmetric fission of highly excited nuclei although in the latter for the spontaneous fission and low energy induced fission, the degree of deformation decreases gradually for Af>245 until Af~260. The constancy of the degree of the scission deformation in each fission mode leads to a TKE systematic formula for each fission mode which is comparable to Viola's TKE formula. Based on the new TKE formulas derived from the deformation parameter of the fissioning nucleus, the extension of the present knowledge to predict the fission properties of the superheavy elements is addressed. It is pointed out that the symmetric and asymmetric fission valleys may merge into one for the superheavy elements around the A=280-290 region. .

Many advanced reactor designs incorporate passive systems mainly to enhance the operational safety and possible elimination of severe accident condition. Some reactors are even designed to remove the nominal fission heat passively by natural circulation without using mechanical pumps e.g. ESBWR, AHWR, CHTR, CAREM, etc. while in most other new reactor concepts, the decay heat is removed passively by natural circulation following the pump trip conditions. The design and safety analysis of these reactors are carried out using the best estimate codes such as RELAP5, TRAC and CATHARE, etc. These best estimate codes have been developed for pumped circulation systems and it is not proven about their adequacy or applicability for natural circulation systems wherein the driving mechanism is complet...

The quasi-resonance structure appearing above the fission threshold in neutron-induced fission cross section of {sup 244}Cm(n,f) is interpreted. It is shown to be due to excitation of few-quasiparticle states in fissioning {sup 245}Cm and residual {sup 244}Cm nuclides. The estimate of quasiparticle excitation thresholds in fissioning nuclide {sup 245}Cm is consistent with pairing gap and fission barrier parameters. (author)

The development of fission reaction theory in relation to its predictive power in the calculation of neutron cross-sections is reviewed. The topics covered include the transition state spectrum and the channel theory; the discovery of complex topography in the fission barrier and the consequences of intermediate structure in fission cross-sections; the evidence of experimental data in parameterizing the fission barrier; and the role of other aspects of collective nuclear motion in controlling fission reaction rates. 51 refs., 6 figs.

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including unenriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed in millions of fuel pebbles, which can be tagged as individual accountable items, and thus made difficult to divert in large quantities. This report discusses the application of the LIFE concept to nonproliferation issues, initially looking at the LIFE (Laser Inertial Fusion-Fission Energy) engine as a means of completely burning WG Pu and HEU. By combining a neutron-rich inertial fusion point source with energy-rich fission, the once-through closed fuel-cycle LIFE concept has the following characteristics: it is capable of efficiently burning excess weapons or separated civilian plutonium and highly enriched uranium; the fission blanket is sub-critical at all times (keff < 0.95); because LIFE can operate well beyond the point at which light water reactors (LWRs) need to be refueled due to burn-up of fissile material and the resulting drop in system reactivity, fuel burn-up of 99% or more appears feasible. The objective of this work is to develop LIFE technology for burning of WG-Pu and HEU.

Studies of zircon grains using optical microscopy, micro-Raman spectroscopy, and scanning electron microscopy (SEM) have been carried out to characterize the surface of natural zircon as a function of etching time. According to the surface characteristics observed using an optical microscope after etching, the zircon grains were classified as: (i) homogeneous; (ii) anomalous, and (iii) hybrid. Micro-Raman results showed that, as etching time increases, the crystal lattice is slightly altered for homogeneous grains, it is completely damaged for anomalous grains, and it is altered in some areas for hybrid grains. The SEM (energy dispersive X-ray spectroscopy, EDS) results indicated that, independent of the grain types, where the crystallinity remains after etching, the chemical composition of zircon is approximately 33% SiO(2):65% ZrO(2) (standard natural zircon), and for areas where the grain does not have a crystalline structure, there are variations of ZrO(2) and, mainly, SiO(2). In addition, it is possible to observe a uniform surface density of fission tracks in grain areas where the determined crystal lattice and chemical composition are those of zircon. Regarding hybrid grains, we discuss whether the areas slightly altered by the chemical etching can be analyzed by the fission track method (FTM) or not. Results of zircon fission track and U-Pb dating show that hybrid and homogeneous grains can be used for dating, and not only homogeneous grains. More than 50 sedimentary samples from the Bauru Basin (southeast Brazil) were analyzed and show that only a small amount of grains are homogeneous (10%), questioning the validity of the rest of the grains for thermo-chronological evolution studies using zircon FTM dating. PMID:22524960

Antiproton-induced fission has been investigated with a novel double-arm fission fragment spectrometer. The correlations in mass, energy, and velocity between two fragments were measured. The dependence of total kinetic energy, velocity, momentum of the fissioning nucleus, and momentum of the fission fragments on the mass loss was deduced and analyzed in the framework of the dynamical statistical model. This model takes into account all stages before, during and after {ital {bar p}}-induced fission (atomic cascade, intranuclear cascade, evaporation cascade, fission of the compound nucleus, evaporation from the fission fragments). The mass loss was used as a measure of the excitation energy to classify the fission events according to the corresponding excitation energies. Some discrepancies between the model calculation and the experiment show the important role of dissipative effects in the {ital {bar p}}-induced fission process. {copyright} {ital 1996 The American Physical Society.}

Metal Matrix Microencapsulated (M3) fuel consists of TRISO or BISO coated fuel particles directly dispersed in a matrix of zirconium metal to form a solid rod (Fig. 1). In this integral fuel concept the cladding tube and the failure mechanisms associated with it have been eliminated. In this manner pellet-clad-interactions (PCI), thin tube failure due to oxidation and hydriding, and tube pressurization and burst will be absent. M3 fuel, given the high stiffness of the integral rod design, could as well improve grid-to-rod wear behavior. Overall M3 fuel, compared to existing fuel designs, is expected to provide greatly improved operational performance. Multiple barriers to fission product release (ceramic coating layers in the coated fuel particle and te metal matrix) and the high thermal conductivity zirconium alloy metal matrix contribute to the enhancement in fuel behavior. The discontinuous nature of fissile material encapsulated in coated particles provides additional assistance; for instance if the M3 fuel rod is snapped into multiple pieces, only the limited number of fuel particles at the failure cross section are susceptible to release fission products. This is in contrast to the conventional oxide fuel where the presence of a small opening in the cladding provides the pathway for release of the entire inventory of fission products from the fuel rod. While conventional metal fuels (e.g. U-Zr and U-Mo) are typically expected to experience large swelling under irradiation due to the high degree of damage from fission fragments and introduction of fission gas into the lattice, this is not the case for M3 fuels. The fissile portion of the fuel is contained within the coated particle where enough room is available to accommodate fission gases and kernel swelling. The zirconium metal matrix will not be exposed to fission products and its swelling is known to be very limited when exposed solely to neutrons. Under design basis RIA and LOCA, fuel performance will be superior to the conventional oxide fuel since PCMI and rod burst, respectively, have been mitigated. Under beyond design basis accident scenarios where the fuel is exposed to high temperature steam for prolonged periods, larger inventory of zirconium metal in the core could negatively affect the accident progression. A thin steam resistant layer (e.g. alumina forming alloy steel), integrated into the solid rod during fabrication by co-extrusion or hot-isostatic-pressing, offers the potential to provide additional fuel protection from steam interaction: blanketing under a range of boiling regimes and under severe accident conditions up to high temperatures well beyond what is currently possible in the conventional fuel. A crucial aspect to the viability of M3 fuel in light water reactors is the reduced heavy metal load compared to standard pellet fuel. This study evaluated the design requirements to operate a Pressurized Water Reactor (PWR) with M3 fuel in order to obtain fuel cycle length, reactivity coefficients, and power peaking factors comparable to that of standard fuel.

The fusion-fission process for the synthesis of superheavy elements is discussed based on fluctuation-dissipation dynamics. Recent data from Dubna on a fusion-fission cross section derived from fission experiments in the reactions 48Ca+208Pb, 48Ca+238U, 48Ca+244Pu and 48Ca+248Cm are analyzed using a three-dimensional Langevin calculation. We found that the quasi-fission process contributes to the yield of mass symmetric fission fragments. In superheavy mass region, the dynamical deformation of fragments is very important. The one- or two-dimensional calculation is not enough to describe the dynamics of fusion-fission process.   

Proton even-odd staggering in fission fragment yields and its relation with energy dissipated in the fission process are investigated with respect to symmetric and asymmetric charge splits. Thermal-neutron and electromagnetic induced fission provide different fissioning systems in a range of incident energies, allowing a systematic study of this even-odd effect in the fragment yields. These data show evidence for a smaller or absent dependence of the even-odd effect at symmetry on the fissility of the fissioning system. This behaviour is still to be explained in the present picture of the fission process.

There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., 235U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either depleted uranium or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia...

During on-site inspections to verify the comprehensive nuclear-test-ban treaty (CTBT), soil gas samples may be taken and analysed for their content of the xenon isotopes 131mXe, 133Xe, 133mXe and 135Xe in order to identify a suspected underground nuclear test. These samples might contain natural radioxenon which is present as a trace gas in the ground. This work analyses the different production mechanisms of natural lithospheric radioxenon to assess theoretically the background concentration under different sampling conditions. The results imply that the equilibrium concentrations of the examined xenon isotopes can be measured in certain rock types using actual CTBTO on-site inspection equipment. Radioxenon production is dominated by spontaneous fission of 238U, resulting in a reactor-like xenon isotopic signature rather than an explosion-like signature.

Monitoring natural waters for the inadvertent release of radioactive fission products produced as a result of nuclear power generation downstream from these facilities is essential for maintaining water quality. To this end, we evaluated sorbents for simultaneous in-situ large volume extraction of radionuclides with both soft (e.g., Ag) and hard metal (e.g., Co, Zr, Nb, Ba, and Cs) or anionic (e.g., Ru, Te, Sb) character. In this study, we evaluated a number of conventional and novel nanoporous sorbents in both fresh and salt waters. In most cases, the nanoporous sorbents demonstrated enhanced retention of analytes. Salinity had significant effects upon sorbent performance and was most significant for hard cations, specifically Cs and Ba. The presence of natural organic matter had little effect on the ability of chemisorbents to extract target elements.

This study presents the measurement results of environmental radioactivity levels for Gaziantep, an industrial and trade centre in the southeastern part of Turkey. The outdoor gamma absorbed dose was measured as 50.1 nGy h-1, corresponding to a total gamma radiation level (of terrestrial and cosmic origin) of 61.5 Sv y-1. The activity concentrations in the surface soil samples collected from the study area were determined as 25.2, 23.7 and 289.2 Bq kg-1 for the natural radionuclides 238U, 232Th and 40K, respectively, and 8.02 Bq kg-1 for the fission product 137Cs. These natural radioactivity sources result in a terrestrial gamma level of 46.9 Sv y-1. The drinking water samples collected from the region carry an average of 0.0493 Bq l-1 of gross alpha and 0.1284 Bq l-1 of gross beta activit...

The fission yield, the transfer factors in the food chain and the dose coefficient are large for the nuclear fission product Sr-90. The surveillance of Sr-90 in the food chain is therefore important in precautionary radiation protection and in assessing the radiation dose to the public especially after a nuclear incident. Prior to analysis, as it is a pure {beta}-emitter, Sr must be separated from the sample by procedures which, for complex organic samples, are lengthy, laborious and dependent on operator skill. Ubiquitous natural radionuclides and short-lived fission products in samples contaminated with fresh fallout may interfere. Here we describe a fast, reproducible and efficient method for extracting Sr from grass, clover, maize, whole meal rye, baby food, and total diet. The method depends on obtaining an ash free of traces of organic interferences. Sr may be separated from a dilute nitric acid leachate of such ash with a solution of dicyclohexyl-18-crown-6 in chloroform. Interfering radionuclides are removed with a special manganese (IV) oxide (active, precipitated from Merck). Sr is precipitated as carbonate then dispersed in a cocktail for liquid scintillation spectrometry. This allows simultaneous counting of Sr-89 (a short-lived {beta}-emitter in fresh fallout) and Sr-90. The chemical yields of Sr determined with the gamma-emitting Sr-85 tracer are reproducible and greater than 75% in all cases. The sample ashing requires 18 h and the extraction 4.0 to 4.5 h. Thus, a duplicate analysis may be completed within 2 days of receipt of the sample. (orig.)

The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now available to provide improved understanding of the VHTR during accidents.

Delayed fission is one of the fission modes of low-lying excited states of nuclei along with spontaneous fission and spontaneously fission shape isomer. The first observation of this phenomena was made in 1966 at JINR. Fission products with half-life on the order of minutes were observed. The nuclei responsible for fission products were identified and it was concluded that they are the precursors of fissioning nuclides: their daughter nuclei are likely to undergo fission from an excited state after electron capture of the parent nucleus. Detailed interpretation of the decay of the nuclear predecessors 228Np, 232Am and 234Am entering into the delayed-fission process was synthesizes the neutron-deficient nuclei as emitters of delayed fission in reaction with beam of heavy ions in the region from neptunium to mendelevium isotopes. The full set experiments showed that delayed fission is common decay channel of heavy nuclei with sufficiently large Qß. Now we have possibilities to make a few new delayed fissioning isotopes of berkelium, einsteinium and other daughter products can undergo fission from excited state.

A large number of fissioning systems have been studied in the frame of the Los Alamos model which showed interesting regular behaviours of the average model parameters (the energy release in fission Er, the total kinetic energy of the fission fragments TKE, the average neutron separation energy Sn from the fission fragments, the prompt gamma-ray energy Eg and the level density parameter a parameterized as C=A/a, where A is the mass number of the fissioning nucleus)) as well as of other quantities in connection with the prompt fission neutron emission (such as the total average prompt neutron multiplicity at thermal incident energy, the total average fission fragment excitation energy leading to prompt neutron emission, the total average prompt fission energy deposition, the average excitat...

We discuss fission properties of the heaviest elements. In particular we focus on stability with respect to spontaneous fission and on the prospects of extending the region of known nuclei beyond the peninsula of currently known nuclides. (author).

Problems of reactor neutron dosimetry in fission track dating are discussed in connection with neutron energy spectra at irradiation facilities commonly used for fission track dating. Problems of spontaneous decay constant of U-238, B value for dosimetry glass, zeta calibration for fission track ages and fission track dating studies in Japan are also briefly reviewed from the viewpoint of reactor neutron dosimetry. Concluding remarks are as follows ; Absolute measurements of reactor neutron fluence and spectra by activation monitors is very difficult for usual irradiation of fission track dating. Fission reaction rate of U-235 of irradiated samples, therefore, must be evaluated directly from uranium standard itself without any parameters of neutron fluence and fission cross section estimated independently. Further confirmation for calibration of dosimeter glasses and age standards is desirable to be done for the standardization of fission track ages. (author).

drop model of Bohr and Wheeler used in explaining nuclear fission. These modes of ... did not consider the problem of what modal patterns ..... Drop Model Vibrations and Simulated Fission," ... Drops," NYU-AA-68-45, New York liniv, , K.Y. - ...

Fission fragment mass-energy correlations have been measured for a large range of fissioning systems. The ''cold fragmentations'' mass distributions show the strong influence of the potential energy surface around scission (shell and charge effects).

remove the energy deposited by fission, hydrogen gas is pumped through small ..... manipulator systems which penetrate high density concrete walls up to six feet in .... is expected to remove virtually all fission products except tritium. During ...

Fission mass yields in different structural elements and mineral separates were studied for the element X. The fission component for Pu-244, and the element X are discussed along with radiogenic Xe-129 and neutron activitation.

The fission decay of giant resonances in the actinide region is reviewed. Results from various experiments which are invariably conflicting are discussed. These include inclusive electron and positron-induced fission, as well as experiments in which fission fragments were detected in coincidence with inelastically scattered electrons or hadrons. Attention is focussed on a recent (..cap alpha..,..cap alpha..'f) experiment in which the fission decay of the giant monopole inelastically scattered ..cap alpha..-particles at and around 0/sup 0/. 49 references.

The SPES project at Laboratori di Legnaro of INFN (Italy) is concentrating on the production of neutron-rich radioactive nuclei for nuclear physics experiment, by the Uranium fission at a rate of 1013 fission/s. The emphasis to neutron-rich isotopes is justified by the fact that this vast territory has been little explored. The Radioactive Ion Beam (RIB) will be produced by ISOL technique using the proton induced fission on a Direct Target of UCx.

According to recent data on equatorial and polar {alpha}-emission in the ternary fission of {sup 252}Cf, the necessary energy to be supplied to the fissioning nucleus is of the order of the energy of the double giant dipole resonance. The authors show that a cluster model of fission can explain the appearance of various very energy-rich processes in an early stage of fission.

Subject matter includes structure of matter (what is matter, forces holding atoms together, visualizing the atom, the chemical elements, atomic symbols, isotopes, radiation from the atom), radioactivity (what holds the nucleus together, can one element change into another element, radiation from the nucleus, half-life, chart of the nuclides), and nuclear fission (nuclear energy release, the fission process, where does fission energy go, radiation and radioactivity resulting from fission).

The dependence of the fusion-fission process on Skyrme forces is studied by using the dynamical cluster-decay model (DCM) and the ?-summed extended-Wong model in the 132Sn+64Ni?196Pt* reaction, where the nuclear proximity potential is obtained by using the semiclassical extended Thomas-Fermi (ETF) approach in the Skyrme energy density formalism (SEDF) under the frozen density approximation. The DCM gives an excellent fit to the measured fusion evaporation residue (ER) and the fission cross sections below and above barrier energies, with ER data needing “barrier lowering” at below-barrier energies for each Skyrme force (an in-built property of the DCM). The fission cross sections show a contribution of quasifission (qf) at the above-barrier two or three highest energies, depending on the Skyrme force. Calculations are illustrated for three Skyrme forces, GSkI, SSk, and SIII. Another interesting result is that there is a change of fission mass distribution from a predominantly asymmetric one to a symmetric one with a decrease in the N/Z ratio of the compound nucleus, independent of the choice of nuclear interaction potential, which gives an opportunity to address the isospin effects in the Pt* nucleus. Within the ?-summed extended-Wong model we find that the GSkI and SSk forces fit the total fusion cross-section data exactly, whereas the SIII force needs “barrier modification” in order to fit the data at below-barrier energies. This happens because the isospin and neutron-proton asymmetry nature of GSkI and SSk forces is different from that of the SIII force, and because the center-of-mass energy Ec.m. dependence of the barrier height for the SIII force and that of Blocki [Ann. Phys. (NY)10.1016/0003-4916(77)90249-4 105, 427 (1977)] differs strongly (by a constant amount of ˜7 MeV) from those for GSKI and SSk forces. Note that, because of the associated preformation factor with each fragment, the DCM has the advantage of treating various decay processes separately, whereas the Wong model describes only the total fusion cross section, a sum of cross sections due to all contributing processes.

In this paper, a series of neutronics analysis of hybrid power reactor is proposed. The ideas of loading different fuels in a modular-type fission blanket is analyzed, fitting different level of fusion developments, i.e., the current experimental power output, the level can be obtained in the coming future and the high-power fusion reactor like ITER. The energy multiplication of fission blankets and tritium breeding ratio are evaluated as the criterion of design. The analysis is implemented based on the D-type simplified model, aiming to find a feasible 1000MWe hybrid power reactor for 5 years' lifetime. Three patterns are analyzed: 1) for the low fusion power, the reprocessed fuel is chosen. The fuel with high plutonium content is loaded to achieve large energy multiplication. 2) For the middle fusion power, the spent fuel from PWRs can be used to realize about 30 times energy multiplication. 3) For the high fusion power, the natural uranium can be directly used and about 10 times energy multiplication can be achieved.

The surrogate materials system was constructed considering PWR spent fuel and the molten salts of the ER: 1) U and TRU {yields} Natural U and Mn (surrogate for Am), 2) Fission products {yields} CsCl and SrCl{sub 2}, 3) Residual salts {yields} LiCl and Li{sub 2}O, 4) Rare earth elements {yields} Y{sub 2}O{sub 3}. Residual salts including fission products were completely vaporized during smelting in the LiCl excess condition (>99%). On the contrary rare earth elements remained in the metal ingot (>99%). The oxide layer on the uranium metal powder inhibited its melting. However the residual salts and Li can prevent the fine uranium powder from surface oxidation. The smelting operation procedure was established according to the experimental results as follows: 1) removal of residual salts by vaporization, 2) melting of metal powder, 3) solidification to ingot form, and 4) separation of dross from the ingot. The smelter based on the gas cooling and resistive heating was designed meeting the design requirements of ACP hot cell demonstration. The key considerations were 1) smelting performance of it for the preparation of spent fuel metal ingot (heating temperature > 1,200 .deg. C, control of inert gas or vacuum atmosphere), and 2) suitability to the KAERI IMEF hot cell facility of the Smelting (especially considering the remote operation/maintenance/repair)

The possibility of obtaining high yields of hydrogen through the exposure of calcium hydroxide to natural uranium fission fragments is confirmed experimentally. The amounts of hydrogen obtained in some experiments were determined not only from the mass-spectrometry data, but also with the use of standard chemical analysis methods. The radiolytic hydrogen yield averaged over six independent experiments comprises 20.41 hydrogen molecules per 100 eV of absorbed fission fragment energy. The corresponding energy efficiency makes up to 60.62. Since on interaction with water or water vapor calcium hydroxide enters into the exothermal reaction to liberate 15.6 kcal/mole, it can easily be regenerated; this was attested to by one of irradiation experiments. Therefore, in the long run, we are dealing with a radiolytic decomposition of water at low temperatures or at temperatures readily available with modern reactor engineering techniques. Comparison of the present data with the characteristics of the facilities now in service or being under design, devised to produce hydrogen with the use of nuclear fuel, shows good prospects for the proposed method. This motivates the performance of more extensive studies in order to develop technical projects on the use of the radiolysis of crystal hydrates or hydroxides in nuclear energetics. 5 refs.

The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even neutron number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both non-flattened and flattened core geometries. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of comparable size. A mass balance analysis revealed that the heterogeneous design may reduce the number of fast reactors needed to close the current once-through light water reactor fuel cycle.

In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

The new-built ThAI test facility is designed to fulfill those validation requirements in the areas of thermal hydraulics and fission product behaviour, in particular iodine. More precise data from thermal-hydraulic experiments are needed for validation of lumped-parameter codes simulating severe accident sequences, e.g. for the containment code system COCOSYS presently under development at GRS. Furthermore, advanced measurement techniques are applied in ThAI to comply with the requirements of CFD codes for detailed data of, e.g., convection flow fields. Concerning fission products, ThAI aims at investigating mass transport phenomena of volatile iodine at a technical scale. This is necessary because iodine mass transport modelling is so far based on small-scale experiments, which cannot reproduce effects of real thermal-hydraulic conditions in severe accidents such as free natural convection flows or stratification of sump and atmosphere. For ThAI, radioactive iodine {sup 123}I is used as a tracer to allow accurate measurements of iodine at low, accident-relevant concentrations. The ThAI test programme consists of a thermal-hydraulic part starting end 2000, and an iodine part to be performed in 2001/2002. Code calculations for the first block of thermal-hydraulic experiments have been made well in advance to use the unique opportunity of predicting the thermal hydraulics of a still unknown facility and thus demonstrate the state of the art of the codes and their application. (author)

The medical isotope Molybdenum-99 is presently used for 80-85% of all nuclear medicine procedures and is produced by irradiating highly enriched uranium U-235 targets in NRU reactors. It was recently proposed that an electron linac be used for the production of 99Mo via photo-fission of a natural uranium target coming from the excitation of the giant dipole resonance around 15 MeV. The photons can be produced using the braking radiation (“bremsstrahlung”) spectrum of an electron beam impinged on a high Z material. In this paper we present an alternate concept for the production of 99Mo which is also based on photo-fission of U-238, but where the ~15 MeV gamma-rays are produced by Compton backscattering of laser photons from relativistic electrons. We assume a laser wavelength of 330 nm, resulting in 485 MeV electron beam energy, and 10 mA of average current. Because the induced energy spread on the electron beam is a few percent, one may recover most of the electron beam energy, which substantially increases the efficiency of the system. The accelerator concept, based on a three-pass recirculation system with energy recovery, is described and efficiency estimates are presented.

A review is presented of the technical basis for predicting radioactivity release resulting from depressurization of an MHTGR primary system. Consideration is restricted to so called dry events with no involvement of the steam system. The various types of deposition mechanisms effective for iodine, cesium, strontium, and silver are discussed in terms of their chemical characteristics and the nature of the materials in the primary system. Emphasis is given to iodine behavior, including means for estimating the quantity available for release, the types of plateout locations in the primary system, and the effect of dust on distribution and release. The behavior of fission products cesium, strontium, and silver in such accidents is presented qualitatively. A major part of the review deals with expected dust levels, types, and transport. Available information on the level and nature of dust in the HTGR primary system is reviewed. A summary is presented of dust deposition and liftoff mechanisms. It was concluded that recent approaches to dust liftoff modeling, based on turbulent burst concepts for removal from surfaces, probably offer advantages over the current shear ratio approach. This study concludes that iodine releases from dry depressurization events are likely to be extremely low, on the order of millicuries, due to a predictably low degree of chemical desorption, a low degree of dust liftoff, and a low involvement of iodine with dust. It was also concluded that deposition mechanisms controlling the distribution of fission product material in the primary system, and hence also controlling the degree of liftoff, depend strongly on the chemical nature of the individual elements. Therefore contrary to the current practice, both plateout and liftoff models should reflect those unique chemical and physical properties. 56 refs., 16 figs., 23 tabs.

Fission foils are commonly used as dosimetry sensors. They play a very important role in neutron spectrum determinations. This paper provides a combination of experimental measurements and calculations to quantify the importance and synergy of several factors that affect the fission response of a dosimeter. Only when these effects are properly treated can fission dosimeters be used with sufficient fidelity.

Spontaneous fission of spherical superheavy nuclei is studied in a multidimensional deformation space. Potential energy, fission barrier and fission half-life are analyzed. The half-life is studied in a dynamical approach. Even-even isotopes of the element 114, with neutron number N=174-186, are considered. Alpha-decay half lives are also given, for completeness. (author) 8 refs, 2 figs

Antiproton-induced fission has been investigated with a novel double-arm fission fragment spectrometer. Inclusive distributions of total mass, individual mass, and total kinetic energy of the fragments, and of the momentum of the fissioning nucleus, were measured. A dynamical model which takes into account all stages of [ital [bar p

The noble metals Ru, Rh, Pd, and Ag were produced in the Savannah River Site (SRS) reactors as products of the fission of U-235. Consequently they are in the High Level Waste (HLW) sludges that are currently being immobilized into a borosilicate glass in the Defense Waste Processing Facility (DWPF). The noble metals are a concern in the DWPF because they catalyze the decomposition of formic acid used in the process to produce the flammable gas hydrogen. As the concentration of these noble metals in the sludge increases, more hydrogen will be produced when this sludge is processed. In the SRS Tank Farm it takes approximately two years to prepare a sludge batch for processing in the DWPF. This length of time is necessary to mix the appropriate sludges, blend them to form a sludge batch and then wash it to enable processing in the DWPF. This means that the exact composition of a sludge batch is not known for {approx}two years. During this time, studies with simulated nonradioactive sludges must be performed to determine the desired DWPF processing parameters for the new sludge batch. Consequently, prediction of the noble metal concentrations is desirable to prepare appropriate simulated sludges for studies of the DWPF process for that sludge batch. These studies give a measure of the amount of hydrogen that will be produced when that sludge batch is processed. This report describes in detail the measurement of these noble metal concentrations in sludges and a way to predict their concentrations from an estimate of the lanthanum concentration in the sludge. Results for two sludges are presented in this report. These are Sludge Batch 3 (SB3) currently being processed by the DWPF and a sample of unwashed sludge from Tank 11 that will be part of Sludge Batch 4. The concentrations of the noble metals in HLW sludges are measured by using mass spectroscopy to determine concentrations of the isotopes that comprise each noble metal. For example, the noble metal Ru is comprised of isotopes with masses 101, 102, and 104. The element Rh has a single isotope with mass 103. The element Pd is comprised of five isotopes. These are at masses 105-108 and mass 110. As does Rh, Ag has only one isotope. This is at mass 109. However, results in this report show that the Ag concentration in the two samples was due to natural Ag being in the samples. Natural Ag has masses at 107 and 109. The Ag-107 interferes with the measurement of Pd-107. This Ag was used in one of the processes at SRS. The results also show that natural Cd is in the two samples. Cadmium has isotopes at masses 106, 108 and 110, thus it interferes with the analysis of the Pd isotopes at these masses. Cadmium was also used in one of the processes at SRS. However, the concentrations of the Pd isotopes at masses 106, 107, 108 and 110 could be calculated using the fission yields for the Pd isotopes, and the measured concentration of Pd at mass 105 where there is no Ag or Cd interference. Based on the measurements of the concentrations of the isotopes of each noble metal, the total concentration of that noble metal can be determined by summing the concentrations of the individual isotopes. The results in this report show that the relative concentrations of the isotopes of Ru and Rh are in proportion to their yields from the fission of U-235 in the reactors. These results were expected since these elements are very insoluble in caustic and thus are primarily in the sludge tanks rather then the salt tanks of the SRS Tank Farm. The relative concentration of Pd is somewhat lower than that based on the relative fission yields of its five isotopes. This indicates that some of the Pd is in the salt tanks rather than the sludge tanks of the Tank Farm. The concentrations of the noble metals were predicted using the High Level Waste Characterization System (WCS) at SRS. This system keeps record of the inventory of the major compounds and select radionuclides that are in each of the SRS HLW tanks. Using this system, the Closure Business Unit (CBU) can predict the major composition of a sludge batch by knowing the tanks involved in that batch and the estimates of the volume of sludge from each tank that will be blended to make the final sludge batch. The system does not track the inventories of Rh, Pd, and Ag. It does track Ru, but Ru is not included in the projections by CBU. However, another U-235 fission product is tracked by WCS. This fission product is La. The element La was not used in any of the chemical processes at SRS. Results in this study show that it is in the HLW primarily as a U-235 fission product. This fission product is comprised solely of the isotope La-139 which can be measured along with the isotopes of the noble metals that do not have interferences from Ag or Cd. The concentrations of La in SB3 and in the Tank 11 sample have been estimated by the CBU using WCS projections.

The fission decay of giant resonances in the actinide region is reviewed. Results from various experiments which are invariably conflicting are discussed. These include inclusive electron and positron induced fission, as well as experiments in which fission fragments were detected in coincidence with inelastically scattered electrons or hadrons. Attention is focussed on a recent (..cap alpha..,..cap alpha..'f) experiment in which the fission decay of the giant monopole resonance was investigated by measuring fission fragments in coincidence with inelastically scattered ..cap alpha..-particles at and around 0/sup 0/.

With the improved scission-point model the mass distributions are calculated for induced fission of different Hg isotopes with the masses 180-196. The drastic change in the shape of the mass distribution from asymmetric to symmetric is revealed with increasing mass number of the fissioning Hg isotope, and the reactions are proposed to verify this prediction experimentally. The asymmetric mass distribution of fission fragments observed in the recent experiment on the fission of 180Hg is explained. The calculated mass distribution and mean total kinetic energy of fission fragments are in a good agreement with the available experimental data.

As a part of the Fast-Mixed Spectrum Reactor (FMSR) Project, a study was made on the adequacy of the conventional fission product lump models for the analysis of the different FMSR core concepts. A two-lump fission product model consisting of an odd-A fission product lump and an even-A fission product lump with transmutation between the odd- and even-A lumps was developed. This two-lump model is capable of predicting the exact burnup-dependent behavior of the fission products within a few percent over a wide range of spectra and is therefore also applicable to the conventional fast breeder reactor.

Measurements were conducted for the fission neutron yields with fission fragments in the (d,pf) reactions at some excitation energies, where threshold neutrons were discovered. These data on the neutron yields in 233U(d,pfn) and 239Pu(d,pfn) reactions have been compared with the dependence of the average of fission neutrons vp(En) in the 233U(n,f) reaction as well as fission probability in the 239Pu(d,pf) reaction on excitation energy, which provides a better understanding of the nuclear fission process in a (d,pf) reaction and the vp(En) dependence on neutron energy.

A new method is used to compute the fission cross sections in which a change of the ratio of the level density parameter in fission to neutron emission channels is taken into account with the change of the incident energy of the projectile. It is shown that fission cross sections induced by nucleons and pions depend on the ratio of the level density parameter in the fission and evaporation modes, i.e. af/an, respectively. We are unable to describe well the cross sections for fission without using this new method. The computed values exhibit reasonable agreement with the experimental data found in the literature across a wide range of beam energies.

In order to estimate the fission width for the case in which the Bohr-Wheeler model is inapplicable due to a fission barrier Bƒ ? 0, we introduce the mean first passage time method and discuss the usefulness of this method. By using this method, the fission decay time can be obtained without solving the Smoluchowski equation, which involves a time consuming calculation. This method is useful when we need to calculate the fission width with a vanishing fission barrier, as when we estimate the fusion evaporation residue cross section for superheavy nuclei with Z > 114.   

The double-velocity and double-energy measurements for thermal-neutron induced fissions of 235U and 233U and spontaneous fission of 252Cf were carried out to extract correlation data among primary and secondary fragment masses, kinetic energy and neutron multiplicity event by event. The resulting data were discussed from a point of view of the multi-mode fission mechanism. Then, a new approach by means of a deformation parameter of fragments was devised to elucidate the correlation of fission configuration and fission mode.

Investigations of mass and kinetic-energy distributions from spontaneous fission have been extended in recent years to an isotope of element 104 and, for half-lives, to an isotope of element 108. The results have been surprising in that spontaneous fission half-lives have turned out to be much longer than expected and mass and kinetic- energy distributions were found to abruptly shift away from those of the lighter actinides, showing two modes of fission. These new developments have caused a re-evaluation of our understanding of the fission process, bringing an even deeper appreciation of the role played by nuclear shell effects upon spontaneous fission properties. 16 refs., 10 figs.

The ART code calculates transport behavior of aerosols and radionuclides during core meltdown accidents in the light water reactors. Since aerosols play an important role in carrying fission products from the core region to the environment, the ART code includes detailed models of aerosol behavior. Aerosols including several radionuclides are classified into many groups according to the aerosol mass. The models of aerosol behavior include agglomeration processes caused by Brownian motion, aerosol settling velocity difference and turbulent flow, and natural deposition processes due to diffusion, thermophoresis, diffusiophoresis, gravitational settling and forced convection. In order to examine validity of the ART models, the NSPP aerosol experiment was analyzed. The ART calculated results showed good agreement with the experimental data. It was ascertained that aerosol growth due to agglomeration, gravitational settling, thermophoresis in an air atmosphere, and diffusiophoresis in an air-steam atmosphere were important physical phenomena in the aerosol behavior. (author).

Interaction of heavy particles (alpha-recoil nuclei, fission fragments, implanted ions) with ceramics is complex because they have a wide range of structure types, complex compositions and chemical bonding is variable. Radiation damage can produce diverse results, but most commonly, crystalline periodic materials become either polycrystalline or aperiodic (metamict state). We studied the transition from crystalline to aperiodic state in natural materials that have been damaged by alpha recoil nuclei in the U and Th decay series and in synthetic, analogous structure types which have been amorphized by ion implantation. Transition from crystalline to aperiodic was followed by analysis of XRD, high resolution TEM, and EXAFS/XANE spectroscopy. Use of these techniques with increasing dose provided data on an increasing finer scale as the damage process progressed.

The production of the heaviest elements in nature occurs via the r-process, i.e. a combination of rapid neutron captures, the inverse photodisintegrations, and slower beta sup - -decays, beta-delayed processes as well as fission and possibly interactions with intense neutrino fluxes. A correct understanding and modeling requires the knowledge of nuclear properties far from stability and a detailed prescription of the astrophysical environment. Experiments at radioactive ion beam facilities have played a pioneering role in exploring the characteristics of nuclear structure in terms of masses and beta-decay properties. Initial examinations paid attention to highly unstable nuclei with magic neutron numbers and their beta-decay properties, related to the location and height of r-process peaks, while recent activities focus on the evolution of shell effects at large distances from the valley of stability. We show in site-independent applications the effect of both types of nuclear properties on r-process abundanc...

Recently, chromium isotopes have been enriched using gas centrifuges that were developed by the United States Department of Energy for the purpose of enriching uranium in the fissionable isotope {sup 235}U. The process gas employed was chromyl-fluoride. The enrichment was carried out using a single gas centrifuge that was set up to emphasize the enrichment of the light isotope {sup 50}Cr, which has a natural abundance of 4.35%. The highest enrichment tested to date is 21.4%. Higher values are possible with additional repasses. The heavy isotopes {sup 54}Cr can be enriched instead of the {sup 50}Cr with an adjustment to the machine flows. Currently, work is under way to refine the cost projections associated with feed production and conversion back to an oxide and to optimize the centrifuge separative performance. (orig.).

During colony founding in honey bees, a portion of a colony?s workforce (the ?swarm fraction?) departs with the old mother queen in a swarm while the remaining workforce stays with a new daughter queen in the parental nest. There is little quantitative information about swarm fraction size and about how swarm fraction size affects the growth and survival of mother-queen and daughter-queen colonies. We measured (a) the swarm fraction in naturally fissioning honey bee colonies, (b) the growth and survival of mother-queen colonies as a function of swarm size, and (c) the growth and survival of mother-queen and daughter-queen colonies as a function of the swarm fraction. We found an average swarm fraction of 0.75. We also found a significant positive effect of swarm size and swarm fraction on ...

In the TVO-92 safety assessment, five reference waters are considered, of which three are natural groundwaters and two are simulated. The reference waters represent saline, brackish fresh and bentonite-equilibrated groundwaters. Solubility and speciation data in the reference waters for some actinide elements (Am, Np, Pu, Th), as well as for some fission and corrosion products (Cu, Fe, Ni, Ra, Se, Sn, Tc) considered in TVO-92 were obtained by geochemical modelling. The code used for modelling was EQ3/G, of which the EQ3 part of the code was used for obtaining the results in the report. The thermodynamic database used with the code is a modified version of the original data base supporting the code. Modification of the data base has been undertaken by Swedish Nuclear Waste Management Company (SKB). Solubility and speciation of the considered elements are given both in reducing and oxidizing groundwater environments.

The Energy plus Transmutation (EPT) set-up of the Joint Institute of Nuclear Research (JINR), Dubna, Russia is composed of a lead spallation target surrounded by a blanket of natural uranium. The target was irradiated with 4 GeV deuterons. Rates of reactions leading to the production of ^2^0^6Bi from ^2^0^9Bi were determined through gamma spectrometry. These were compared with calculated reaction rates determined using the MCNPX code, including the INCL4 intranuclear cascade and ABLA fission/evaporation models along with available cross section data. Close agreement was found for the production of ^2^0^6Bi. The high threshold energy of neutron induced production indicates that neutrons causing these reactions are most likely to originate in the intranuclear cascade phase of the interaction...

The passage of Pioneer 11 through Saturn's magnetosphere revealed an especially intense region of high-energy particle fluxes that places unique constraints on models for sources of high-energy protons in the innermost radiation zones. Of special interest is the flux of protons with energies above 35 MeV which was measured with a fission cell in the innermost magnetosphere between the A ring and the orbit of Mimas. The negative phase space density gradients derived from the proton and electron observations in this region imply that steady-state inward diffusion from the outer magnetosphere is not an adequate source for these high-energy protons. In the present paper, the nature of the Crand source at Saturn is examined, and its significance for injection of high-energy protons into the region inside L = 4 is estimated.

The MAAP 3.0B code is a coupled thermal-hydraulics and fission product release and transport code for prediction of boiling water reactor and pressurized water reactor (PWR) responses to severe reactor accidents. MAAP 3.0B is maintained by the Electric Power Research Institute to assist in utility performance of IPEs (individual plant evaluations) and other source term calculations. Significant improvements have been made to MAAP 3.0B during 1989, including methods of interest for other code developers. Three improvements are selected for this discussion: (1) a model for flammability and combustion of possible gas mixtures, (2) a model for natural circulation intercompartmental flow including countercurrent flows, and (3) integration diagnostics and control.

Although several computer codes exist which offer the ability to simulate the response of containments to hypothesized events, one appears to offer more in the area of best estimate performance during normal operations and following an accident prior to the release of fission products. This computer code is COBRA-NC due to its ability to analyze the containment multi-dimensionally, to track several species of noncondensible gases, entrained liquid and liquid films throughout the containment, and have the ability to analyze the containment in a passive mode of operation when the only driving forces present are due to natural circulation. Objectives specific to this task were to: enhance the ability of COBRA-NC to analyze both current and advanced reactor containments; and assess the ability, stability and biases of the condensation heat transfer correlations which may be used in containment analyses performed by the utilities and their regulators. 23 refs., 21 figs., 2 tabs.

The existence of dynamical thresholds to fusion in heavy nuclei (A greater than or equal to 200) due to the nature of the potential-energy surface is shown. These thresholds exist even in the absence of dissipative forces, due to the coupling between the various collective deformation degrees of freedom. Using a macroscopic model of nuclear shape dynamics, It is shown how three different suggested dissipation mechanisms increase by varying amounts the excitation energy over the one-dimensional barrier required to cause compound-nucleus formation. The recently introduced surface-plus-window dissipation may give a reasonable representation of experimental data on fusion thresholds, in addition to properly describing fission-fragment kinetic energies and isoscalar giant multipole widths. Scaling of threshold results to asymmetric systems is discussed. 48 refs., 10 figs.

In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

/sup 238/Pu is produced by irradiating /sup 237/Np. The /sup 237/Np is produced as a byproduct when natural or enriched uranium is irradiated with neutrons. The /sup 237/Np is separated by solvent extraction and ion exchange. It is converted to NpO/sub 2/ and fabricated into targets for irradiation. The irradiated targets are cooled and dissolved in strong nitric acid. The /sup 238/Pu and /sup 237/Np are separated from fission products and other cationic impurities and from each other by three cycles of anion exchange. The /sup 237/Np is recycled to produce more targets for irradiation. The pure /sup 238/Pu solution is precipitated as Pu oxalate and calcined to PuO/sub 2/. After several powder-conditioning steps, the PuO/sub 2/ is hot pressed into fuel forms. Each form is encased in iridium for loading into a specially designed power unit for space application. 8 references, 9 figures, 1 table.

We describe the null geometry of a multiple black hole event horizon in terms of a conformal rescaling of a flat space null hypersurface. For the prolate spheroidal case, we show that the method reproduces the pair-of-pants shaped horizon found in the numerical simulation of the head-on-collision of black holes. For the oblate case, it reproduces the initially toroidal event horizon found in the numerical simulation of collapse of a rotating cluster. The analytic nature of the approach makes further conclusions possible, such as an important bearing on the hoop conjecture. From a time reversed point of view, the approach yields a description of the past event horizon of a fissioning white hole, which can be used as null data for the characteristic evolution of the exterior space-time.

One of the basic prerequisites for the use of bentonite as engineering barrier in deep geological repositories for radioactive waste and spent nuclear fuel is their stability against ionizing radiation stemming from radionuclides present in radioactive waste and spent nuclear fuel. The aim of this study was to compare the changes in the adsorption properties of selected Slovak bentonites in relation to uranium fission products (137Cs and 90Sr), prior to and after irradiation of bentonites with a 60Co ?-source and specifying the changes in the structure of Slovak bentonites induced by ?-radiation. The changes in irradiated natural forms of Slovak bentonites and the changes in their natrified analogues and fractions with different grain sizes were studied from five Slovak deposits: Jelšov...

This paper is a course of transparencies for a general presentation of nuclear and physico-chemical properties of plutonium. It comprises: an historical review of discoveries concerning plutonium isotopes fissile properties and production; fuel production and use in French reactors; nuclear properties and natural decay of plutonium nuclides; instabilities induced by neutrons (fission and capture); abundance in nuclear fuels; crystal structure and physico-chemical properties (phase diagrams) of solid plutonium compounds; chemical properties, oxidation states and absorption spectra of plutonium in aqueous solution; hydrolytic, redox properties, and complexation of plutonium ions; solvent and resins extractions for purification; and finally, production and use of plutonium (fuel cycle and fuel reprocessing). (J.S.). 10 figs., 4 tabs.

Shortly before the Second World War time, Nishina reported on a series of prominent nuclear physical and radiochemical studies in collaboration with Kimura. They artificially produced 231Th, a member of the natural actinium series of nuclides, by bombarding thorium with fast neutrons. This resulted in the discovery of 237U, a new isotope of uranium, by bombarding uranium with fast neutrons, and confirmed that 237U disintegrates into element 93 with a mass number of 237. They also identified the isotopes of several middle-weighted elements produced by the symmetric fission of uranium. In this review article, the highlights of their work are briefly summarized along with some explanatory commentaries.(Communicated by Toshimitsu YAMAZAKI, M.J.A.)   

Design studies are underway for the deep ocean antineutrino observatory Hanohano. The 10 kton monolitic underwater detector will be able to make precision measurement of neutrino mixing parameters (including ?13 and neutrino mass hierarchy) if stationed around 60 km offshore, from the nuclear reactor. Hanohano will be a mobile detector and placing it in a mid-Pacific location will provide the first ever flux measurement of geoneutrinos (antineutrinos emitted in the radioactive decay series of uranium and thorium), coming from the Earth's mantle and perform a sensitivity search for a hypothetical natural fission reactor in the Earth's core. Additional deployment at a different mid-ocean location will lead to tests of lateral heterogeneity of uranium and thorium in the Earth's mantle. These measurements would provide an important insight into deep-Earth geophysics, mantle composition and understanding of the Earth's heat flow and sources of energy inside the Earth.

This paper presents science potential of a deep ocean antineutrino observatory being developed at Hawaii. The observatory design allows for relocation from one site to another. Positioning the observatory some 60 km distant from a nuclear reactor complex enables precision measurement of neutrino mixing parameters, leading to a determination of neutrino mass hierarchy and {theta}{sub 13}. At a mid-Pacific location the observatory measures the flux and ratio of uranium and thorium decay neutrinos from earth's mantle and performs a sensitive search for a hypothetical natural fission reactor in earth's core. A subsequent deployment at another mid-ocean location would test lateral heterogeneity of uranium and thorium in earth's mantle.

Recent experiments on beta-delayed fission in the mercury-lead region and the discovery of asymmetric fission in $^{180}$Hg [1] have stimulated renewed interest in the mechanism of fission in heavy nuclei. Here we study fission modes and fusion valleys in $^{180}$Hg and $^{198}$Hg using the self-consistent nuclear density functional theory employing Skyrme and Gogny energy density functionals. We show that the observed transition from asymmetric fission in $^{180}$Hg towards more symmetric distribution of fission fragments in $^{198}$Hg can be explained in terms of competing fission modes of different geometries that are governed by shell effects in pre-scission configurations. The density distributions at scission configurations are studied and related to the experimentally observed mass splits.

New TKE formulas that will replace the previous existing ones are obtained. Recently, three types of the final deformation of fissioning nuclei were found for actinides out off which only one was observed. The final deformations of the fissioning nuclei was found to be constant and independent of the mass and temperature of the fissioning system. These hence allow to deduce a new formula for the TKE release in nuclear fission process based on the invariance of scission deformations of fissioning nuclei. They yield, TKE(sym) = 0.1173 x (Z{sub f}{sup 2}/A{sub f}{sup 1/3})+7.5 MeV for the symmetric fission and TKE(asym) = 0.1217 x (Z{sub f}{sup 2}/A{sub f}{sup 1/3})+3.5 MeV for the asymmetric fission. Details for the new formulas and their comparison with the experimental data are given. (author)

Nuclear fission process involves large scale shape changes of the nucleus, while it evolves from a nearly spherical configuration to two separated fission fragments. The dynamics of these shape changes in the nuclear many body system is governed by a strong interplay of the collective and single particle degrees of freedom. With the availability of heavy ion accelerators, there has been an impetus to study the nuclear dynamics through the investigations of nucleus--nucleus collisions involving fusion and fission process. From the various investigations carried out in the past years, it is now well recognized that there is large scale damping of collective modes in heavy ion induced fission reactions, which in other words implies that nuclear motion is highly viscous. In recent years, there have been many experimental observations in heavy ion induced fission reactions at medium bombarding energies, which suggest possible occurrence of various non-equilibrium modes of fission such as quasi-fission, fast fission and pre-equilibrium fission, where some of the internal degrees of freedom of the nucleus is not fully equilibrated. We have carried out extensive investigations on the fission fragment angular distributions at near barrier bombarding energies using heavy fissile targets. The measured fragment anisotropies when compared with the standard saddle point model (SSPM) calculations show that for projectile-target systems having zero or low ground state spins, the angular anisotropy exhibits a peak-like behaviour at the sub barrier energies, which cannot be explained by the SSPM calculations. For projectiles or targets with large ground state spins, the anomalous peaking gets washed out due to smearing of the K-distribution by the intrinsic entrance channel spins. Recently studies have been carried out on the spin distributions of fission fragments through the gamma ray multiplicity measurements. The fission fragments acquire spin mainly from two sources: (i) due to rigid rotation of the nascent fragments at scission and (ii) due to statistical excitation of the spin bearing collective modes in the fissioning nucleus. One of the collective modes -- the tilting mode depends on the K quantum number and is responsible for the emission angle dependence of fragment spin. In our studies, we have shown conclusively that the collective statistical spin modes get strongly suppressed for high K values corresponding to large rotational frequencies along the fission axis. These results bring out the importance of the dynamical effects in the heavy ion induced fusion-fission reactions. The present article will review the work carried out on the above aspects in heavy ion fission reactions as well as on the fission time scales, and some of the recent studies on the mass-energy correlations of fission fragments at near-barrier bombarding energies.

Spontaneous fission is a phenomenon exhibited by heavy nuclei, which can be a major mode of decay of nuclei of elements heavier than thorium and can be a determining factor in their stability. For purposes of this paper, spontaneous fission will be considered a process in which a nucleus breaks up into two approximately equal parts. The emission of light nuclei or heavy ions such as {sup 12}C, {sup 16}O, or {sup 32}S will not be considered. This radioactive decay mode is often much smaller than the spontaneous fission decay mode, although this is not true in all cases. Barwick noted that this might indicate that the assumed half-life for spontaneous fission of some older experiments might be partially due to heavy fragment radioactivity. Other than taking note of this potential correction to spontaneous fission half-lives, this decay mode of heavy fragment radioactivity will be ignored. Excited states of some heavy nuclei may decay via spontaneous fission. These so-called fission isomers will not be discussed here. Electron capture (EC) or beta-delayed fission is a process in which prompt fission of a sufficiently excited daughter state occurs following population by EC or beta decay. The fission activity will appear to decay with the half-life of the parent and was earlier confused in some cases with SF. This process has been discussed in detail in a review and will not be considered in this paper.

The Monte Carlo N-Particle Transport Code (MCNP) is employed to simulate the radioisotope production process that leads to the creation of Technetium-99m (Tc-99m). Tc-99m is a common metastable nuclear isomer used in nuclear medicine tests and is produced from the gamma decay of Molybdenum-99 (Mo-99). Mo-99 is commonly produced from the fission of Uranium-235, a complicated process which is only performed at a limited number of facilities. Due to the age of these facilities, coupled with the critical importance of a steady flow of Mo-99, new methods of generating Mo-99 are being investigated. Current experiments demonstrate promising alternatives, one of which consists of the neutron activation of Molybdenum-98 (Mo-98), a naturally occurring element found in nature. Mo-98 has a small cross section (.13 barns), so investigations are also aimed at overcoming this natural obstacle for producing Tc-99m. The neutron activated Mo-98 becomes Mo-99 and subsequently decays into radioactive Tc-99m. The MCNP code is being used to examine the interactions between the particles in each of these situations, thus determining a theoretical threshold to maximize the reaction's efficiency. The simulation results will be applied to ongoing experiments at the PPPL, where the empirical data will be compared to predictions from the MCNP code.

The recoil effects accompanying the {alpha} decay of radionuclides led to the discovery of several radioisotopes, and have been utilized for the separation of radioelements in carrier-free state. On a much larger scale, {alpha}-recoil effects are operative in nature, and they are the origin of the disruption of the radioactive equilibrium between the members of natural radioactive series. In {alpha} decay, the recoil atoms cause considerable recoil damage, and the ultimate fate of the recoil atoms depends on the thermal chemical reactions which in particular determine the final oxidation state. Hot atom chemistry can help to elucidate the oxidation state of actinide daughters, and to understand the role of media and of the induced radiation damage. These factors determine the leachability and subsequently the migration behavior of the actinide and fission products. The disposal of radwaste may be more appropriate by reducing media than the case of usual matrices of ceramics or glassy material. Useful guidelines were provided by natural events and hot atom chemistry experiments conducted in the laboratory. (K.I.).

A new method has been developed for the measurement of neutron flux distribution and {gamma}-ray flux distribution in extremely narrow space by using an optical fibre with which a mixture of ZnS(Ag) scintillator and neutron converter was painted on the tip of the fibre. The outer diameter of the detector can be down to 1 mm, usually 2.6 mm, and the thickness of the scintillator mixture painted is about 0.3 mm. The length of the fibre can be extended to 20 {approx} 30 m depending on the necessity. Since the diameter is very thin, it can be inserted into very narrow spaces such as those between rector fuel elements, or between shielding materials. The fibre can be driven with an appropriate constant speed or intermittently by an automatic driving unit. The geometrical distribution of neutron flux can be measured in a very short time of about 10 minutes or so and with fine position resolution of less than 1 mm by using a neutron converter such as {sup 6}Li, {sup 235}U, or {sup 232}Th in the mixture.The nuclide {sup 235}U or {sup 232}Th is applicable to high neutron flux up to about 10{sup 10}n,cm{sup -2}s. Because the cross sections of nuclear fission are not so large, 4.2 barn for thermal neutrons in the case of natural uranium, and the fission energies of them are very large compared with those of ambient {gamma}-rays, and therefore the signal pulse heights are exceedingly large, which makes it possible to easily distinguish between {gamma}-ray signals and neutron signals. The threshold energy of nuclear fission for {sup 232}Th is about 1.1 MeV, so that the nuclide can be used for the measurement of fast neutrons. Since the reaction cross section of {sup 6}Li(n,{alpha})T is rather large, 945 barns for thermal neutrons, the neutron flux can be measured down to about 10{sup 3}n,cm{sup -2}s depending on the measurement time. This method was applied to obtain thermal neutron flux distribution and fast neutron one in research reactors and around an accelerator for fast neutron generation. The distribution of {gamma}rays can also be obtained by using only ZnS(Ag) scintillator or CsI(TI) scintillator without converter, which was applied to obtain {gamma}-ray intensity distributions in a critical assembly and Co-60 irradiation facility 3 refs., 6 figs.

Closing the nuclear fuel cycle requires reprocessing spent fuel to recover the long-lived components that still have useful energy content while immobilizing the remnant waste fission products in stable forms. At the genesis of this project, next generation spent fuel reprocessing methods were being developed as part of the U.S. Department of Energy's Advanced Fuel Cycle Initiative. One of these processes was focused on solvent extraction schemes to isolate cesium (Cs) and strontium (Sr) from spent nuclear fuel. Isolating these isotopes for short-term decay storage eases the design requirements for long-term repository disposal; a significant amount of the radiation and decay heat in fission product waste comes from Cs-137 and Sr-90. For the purposes of this project, the Fission Product Extraction (FPEX) process is being considered to be the baseline extraction method. The objective of this project was to evaluate the nature and behavior of candidate materials for cesium and strontium immobilization; this will include assessments with minor additions of yttrium, barium, and rubidium in these materials. More specifically, the proposed research achieved the following objectives (as stated in the original proposal): (1) Synthesize simulated storage ceramics for Cs and Sr using an existing labscale steam reformer at Purdue University. The simulated storage materials will include aluminosilicates, zirconates and other stable ceramics with the potential for high Cs and Sr loading. (2) Characterize the immobilization performance, phase structure, thermal properties and stability of the simulated storage ceramics. The ceramic products will be stable oxide powders and will be characterized to quantify their leach resistance, phase structure, and thermophysical properties. The research progressed in two stages. First, a steam reforming process was used to generate candidate Cs/Sr storage materials for characterization. This portion of the research was carried out at Purdue University and is detailed in Appendix A. Steam reforming proved to be too rigorous for efficient The second stage of this project was carried out at Texas A&M University and is Detailed in Appendix B. In this stage, a gentler ceramic synthesis process using Cs and Sr loaded kaolinite and bentonite clays was developed in collaboration with Dr. M. Kaminski at Argonne National Laboratory.

Statistical model calculations of prompt fission neutron spectra (PFNS) from {sup 235}U(n,F) reaction were performed for incident neutron energies up to E{sub n}{approx}20 MeV. Exclusive spectra of the pre-fission (pre-saddle) (n,xnf) reaction neutrons, were calculated with Hauser-Feshbach statistical model, fission and (n,xn) reaction cross section data being described consistently. Spectra of neutrons, evaporated from the fission fragments, were approximated as a sum of two Watt distributions. The reduced neutron velocity in the center-of-mass system due to the neutron emission during fragment acceleration was assumed. PFNS component due to pre-saddle neutrons is evidenced in the shape of the measured PFNS data. We show it to be strongly correlated with the emissive fission contributions to the observed fission cross sections.

SummaryThe GTPase dynamin polymerizes into a helical coat that constricts membrane necks of endocytic pits to promote their fission. However, the dynamin mechanism is still debated because constriction is necessary but not sufficient for fission. Here, we show that fission occurs at the interface between the dynamin coat and the uncoated membrane. At this location, the considerable change in membrane curvature increases the local membrane elastic energy, reducing the energy barrier for fission. Fission kinetics depends on tension, bending rigidity, and the dynamin constriction torque. Indeed, we experimentally find that the fission rate depends on membrane tension in vitro and during endocytosis in vivo. By estimating the energy barrier from the increased elastic energy at the edge of dyna...

35 chain yields are determined for the fission of sup 2 sup 3 sup 5 U induced by 19.1 MeV neutrons by HPGe gamma-ray spectrometry. Absolute fission rate is monitored with a double-fission chamber. The efficiency of the fission chamber is checked with absolute determination of sup 1 sup 9 sup 8 Au activity from sup 1 sup 9 sup 7 Au (n, gamma) sup 1 sup 9 sup 8 Au reaction for the first time. Fission product activities of irradiated sup 2 sup 3 sup 5 U foils are measured by HPGe gamma-ray spectrometry without chemical separation. Threshold detector method is used to estimate the fission events induced by neutrons of other energies

We discuss the sensitivity of fission barrier for heavy neutron-rich nuclei to fission paths in the two dimensional neutron-proton quadrupole plane. To this end, we use the constrained Skyrme-Hartree-Fock + BCS method, and examine the difference of fission barriers obtained with three constraining operators, that is, the neutron, proton, and mass quadrupole operators. We investigate $^{220}$U, $^{236}$U, and $^{266}$U, %from proton-rich to neutron-rich uranium isotopes, that is relevant to r-process nucleosynthesis. We find that the fission barrier heights are almost the same among the three constraining operators even for neutron-rich nuclei, indicating that the usual way to calculate fission barriers with the mass quadrupole operator is well justified. We also discuss the difference between proton and neutron deformation parameters along the fission paths.

An improved calculation is presented for the prompt fission neutron spectrum N(E) from the spontaneous fission of /sup 252/Cf. In this calculation the fission-spectrum model of Madland and Nix is used, but with several improvements leading to a physically more accurate representation of the spectrum. Specifically, the contributions to N(E) from the entire fission-fragment mass and charge distributions will be calculated instead of calculating on the basis of a seven- point approximation to the peaks of these distributions as has been done in the past. Therefore, values of the energy release in fission, fission-fragment kinetic energy, and compound nucleus cross section for the inverse process will be considered on a point-by-point basis over the fragment yield distributions instead of considering averages of these quantities over the peaks of the distributions. Preliminary results will be presented and compared with a measurement, an earlier calculation, and a recent evaluation of the spectrum. 14 refs., 4 figs.

Electron-capture delayed fission was observed in {sup 244}Es produced via the {sup 237}Np({sup 12}C,5n){sup 244}Es reaction at 81 MeV (on target) with a production cross section of 0.31{+-}0.12 {micro}b. The mass-yield distribution of the fission fragments is highly asymmetric. The average preneutron-emission total kinetic energy of the fragments was measured to be 186{+-}19 MeV. Based on the ratio of the number of fission events to the measured number of {alpha} decays from the electron-capture daughter {sup 244}Cf (100% {alpha} branch), the probability of delayed fission was determined to be (1.2{+-}0.4) x 10{sup -4}. This value for the delayed fission probability fits the experimentally observed trend of increasing delayed fission probability with increasing Q value for electron-capture.

Most of the radioactivity produced by the operation of a nuclear reactor results from the fission process, during which the nucleus of a fissionable atom (such as 235U) splits into two or more nuclei, which typically are radioactive. The Radionuclide Assessment Program (RAP) has reported on fission products cesium, strontium, iodine, and technetium. Many other radionuclides are produced by the fission process. Releases of several additional fission products that result in dose to the offsite population are discussed in this publication. They are 95Zr, 95Nb, 103Ru, 106Ru, 141Ce, and 144Ce. This document will discuss the production, release, migration, and dose to humans for each of these selected fission products.

SOFIA, Studies On FIssion with Aladin, as well as FELISe, Fission@ELISe are both part of the forthcoming GSI fission experimental program. They will benefit from relativistic actinide beams available at GSI to induce electromagnetic fission in reverse kinematics. SOFIA will take place in Cave C in the curent GSI facility, while FELISe will be located at FAIR, the GSI part to come. Both will enable to determine nuclear charge and mass as well as kinetic energy for each fission fragment. The neutron multiplicity will be measured too, and the determination of the excitation energy of the fissioning system will be possible at the FAIR facility in the FELISe campaign. To be able to separate the masses, high resolution detection is required. In particular the time-of-flight should be measured with 35 ps FWHM, which presents a real technical challenge. SOFIA and FELISe experiments will be presented in this work.

The results of the Laser Fusion Breeder Design Study are given. This information primarily relates to the conceptual design of an inertial confinement fusion (ICF) breeder reactor (or fusion-fission hybrid) based upon the HYLIFE liquid metal wall protection concept developed at Lawrence Livermore National Laboratory. The blanket design for this breeder is optimized to both reduce fissions and maximize the production of fissile fuel for subsequent use in conventional light water reactors (LWRs). When the suppressed fission blanket is compared with its fast fission counterparts, a minimal fission rate in the blanket results in a unique reactor safety advantage for this concept with respect to reduced radioactive inventory and reduced fission product decay afterheat in the event of a loss-of-coolant-accident.

Most of the experimental work on the interaction of neutrons with matter has focused on materials important to reactor physics and reactor structures. By comparison, the corresponding data for minor actinides or long-lived fission products are poor. A significant demand has developed for improved neutron cross-section data of these little-studied nuclides due to the surge of interest in the transmutation of nuclear waste. With 400 kg of {sup 129}I produced yearly in the reactors of the EU countries and a very long {beta}{sup -} half-life of 1.57 x 10{sup 7} years, iodine requires disposal strategies that will isolate this isotope from the environment for long periods of time. Therefore, {sup 129}I is potentially a key long-lived fission product for transmutation applications, since {sup 129}I transmutes in {sup 130}I after a single neutron capture and decays to {sup 130}Xe with a 12.36 h half-life. Accurate capture cross sections would help to reduce uncertainties in waste management concepts. For that purpose, Time-Of-Flight measurements covering the [0.5 eV-100 keV] energy range have been carried out at the 150 MeV pulsed neutron source GELINA of the Institute for Reference Materials and Measurements (IRMM). Two types of experiments have been performed at the IRMM, namely capture and transmission experiments. They are respectively related to the neutron capture and total cross sections. Since the PbI{sub 2} samples used in this work contain natural and radioactive iodine, extensive measurements of {sup 129}I have been carried out under the same experimental conditions as for the {sup 129}I. The data reduction process was performed with the AGS system, and the resonance parameters were extracted with the SAMMY and REFIT shape analysis codes. In a last step, the parameters have been converted into ENDF-6 format and processed with the NJOY code to produce point-wise and multigroup cross sections, as well as MCNP and ERANOS libraries. (author)

The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling methods used in this study.

Dissociation and fission of small neutral, singly and doubly charged strontium clusters are studied by means of ab initio density functional theory methods and high-resolution time-of-flight mass spectrometry. Magic numbers for small strontium clusters possessing enhanced stability towards monomer evaporation and fission are determined. It is shown that ionization of small strontium clusters results in the alteration of the magic numbers. Thermal promotion of the Coulombic fission for the Sr_7^{2+} cluster is predicted.

A device was developed for detecting the failure of a fuel element in a reactor by monitoring for the presence of gaseous fission products. Small amounts of gaseous fission products were detected in the presence of radioactive argon by separating the fission product gases from the argon by chromatography. An automatic sequencing device was provided for taking samples at short intervals to ensure rapid detection of failures. (auth)

The main task of this work is the measurement of fast neutron induced fission cross section for minor actinides of {sup 238}Pu, {sup 242m}Am, {sup 243,244,245,246,247,248}Cm. The task of the work is to increase the accuracy of data in MeV energy region. Basic experimental method, fissile samples, fission detectors and electronics, track detectors, alpha counting, neutron generation, fission rate measurement, corrections to the data and error analysis are presented in this paper. (author)

There has been considerable interest in improving the fission models in the LAHET Monte Carlo code for the transport and interaction of nucleons, pions, muons, light ions, and antinucleons. Although subactinide fission contributes little to neutron production in lead or tungsten targets, it can be significant for simulation of target activation and fission product contamination. The availability of new data permits new comparisons to be made between experiment and calculation.

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.

We measured the mass and kinetic-energy partitioning in the spontaneous fission of /sup 258/Fm, /sup 259/Md, /sup 260/Md, /sup 258/No, and /sup 260/(104). Surprisingly, these energy distributions were skewed upward or downward from the peak in each case, except for /sup 260/(104), indicating a composite of two energy distributions. We interpret this as a new mode of fission in which there is mixture of liquid-drop-like and fragment-shell-directed symmetric fission.

Investigation of the 235U(n th, f) reaction using the miniFOBOS double-arm time-of-flight spectrometer of fission fragments confirmed manifestations of the earlier unknown many-body, at least ternary, decay involving almost collinear decay-product escape, which were first observed in the spontaneous fission of 252Cf(sf). The use of variables sensitive to the nuclear charge of fission fragments allowed the reliability of identification of decay events to be increased and new decay modes to be revealed.

Measurements of the half-lives for spontaneous fission of the nuclidic ground states of elements from thorium to fermium have been compiled and evaluated. Recommended values are presented. An attempt has been made to distinguish between spontaneous fission and heavy ion emission. Spontaneously fissioning isomers have not been considered here. The difference between even-even nuclides and odd-even, even-odd and odd-odd nuclides are discussed. 3 tabs.

An improvement in the bismuth phosphate carrier precipitation process is presented for the recovery and purification of plutonium. When plutonium, in the tetravalent state, is carried on a bismuth phosphate precipitate, amounts of centain of the fission products are carried along with the plutonium. The improvement consists in washing such fission product contaminated preeipitates with an aqueous solution of ammonium hydrogen fluoride. since this solution has been found to be uniquely effective in washing fission production contamination from the bismuth phosphate precipitate.

The work in this thesis is concerned with some specific aspects of fission product chemistry encountered in a severe reactor accident. The fission products selected for this study were iodine, caesium and tellurium. Considerable emphasis is also placed on the involvement of structural materials such as boric acid in determining the fate of the fission products. Matrix isolation infrared spectroscopy and mass spectrometry were the techniques chosen to investigate the chemistry. (author).

Abstract in english This report describes a method for calculating fusion and decay probabilities in reactions leading to the production of transfermium elements. The competition between quasi-fission and fusion is described on the basis of the Dinuclear System Concept (DNSC). Both the competition between fusion and quasi-fission and statistical decay of heavy highly fissionable excited compound nuclei is described in an approach based on the Monte-Carlo method.