WorldWideScience

Sample records for basic fission products

  1. Fission product yields

    International Nuclear Information System (INIS)

    Data are summed up necessary for determining the yields of individual fission products from different fissionable nuclides. Fractional independent yields, cumulative and isobaric yields are presented here for the thermal fission of 235U, 239Pu, 241Pu and for fast fission (approximately 1 MeV) of 235U, 238U, 239Pu, 241Pu; these values are included into the 5th version of the YIELDS library, supplementing the BIBFP library. A comparison is made of experimental data and possible improvements of calculational methods are suggested. (author)

  2. Fission product data library

    International Nuclear Information System (INIS)

    A library is described of data for 584 isotopes of fission products, including decay constants, branching ratios (both burn-up and decay), the type of emitted radiation, relative and absolute yields, capture cross sections for thermal neutrons, and resonance integrals. When a detailed decay scheme is not known, the mean energies of beta particles and neutrino and gamma radiations are given. In the ZVJE SKODA system the library is named BIBFP and is stored on film No 49 of the NE 803 B computer. It is used in calculating the inventory of fission products in fuel elements (and also determining absorption cross sections for burn-up calculations, gamma ray sources, heat generation) and in solving radioactivity transport problems in the primary circuit. It may also be used in the spectrometric method for burn-up determination of fuel elements. The library comprises the latest literary data available. It serves as the basis for library BIBGRFP storing group constants of fission products with independent yields of isotopes from fission. This, in turn, forms the basis for the BIBDN library collecting data on the precursors of delayed neutron emitters. (author)

  3. Current position on fission product behavior

    International Nuclear Information System (INIS)

    The following phenomena are treated and modeled: fission product release from fuel, both in-vessel and ex-vessel; fission product deposition in the primary system, fission product deposition in the containment, and fission product revolatization

  4. Chemical Production using Fission Fragments

    International Nuclear Information System (INIS)

    Some reactor design considerations of the use of fission recoil fragment energy for the production of chemicals of industrial importance have been discussed previously in a paper given at the Second United Nations International Conference on the Peaceful Uses of Atomic Energy [A/Conf. 15/P.76]. The present paper summarizes more recent progress made on this topic at AERE, Harwell. The range-energy relationship for fission fragments is discussed in the context of the choice of fuel system for a chemical production reactor, and the experimental observation of a variation of chemical effect along the length of a fission fragment track is described for the irradiation of nitrogen-oxygen mixtures. Recent results are given on the effect of fission fragments on carbon monoxide-hydrogen gas mixtures and on water vapour. No system investigated to date shows any outstanding promise for large-scale chemical production. (author)

  5. The SPIDER fission fragment spectrometer for fission product yield measurements

    Energy Technology Data Exchange (ETDEWEB)

    Meierbachtol, K.; Tovesson, F. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Shields, D. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Colorado School of Mines, Golden, CO 80401 (United States); Arnold, C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Blakeley, R. [University of New Mexico, Albuquerque, NM 87131 (United States); Bredeweg, T.; Devlin, M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Hecht, A.A.; Heffern, L.E. [University of New Mexico, Albuquerque, NM 87131 (United States); Jorgenson, J.; Laptev, A. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Mader, D. [University of New Mexico, Albuquerque, NM 87131 (United States); O' Donnell, J.M.; Sierk, A.; White, M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2015-07-11

    The SPectrometer for Ion DEtermination in fission Research (SPIDER) has been developed for measuring mass yield distributions of fission products from spontaneous and neutron-induced fission. The 2E–2v method of measuring the kinetic energy (E) and velocity (v) of both outgoing fission products has been utilized, with the goal of measuring the mass of the fission products with an average resolution of 1 atomic mass unit (amu). The SPIDER instrument, consisting of detector components for time-of-flight, trajectory, and energy measurements, has been assembled and tested using {sup 229}Th and {sup 252}Cf radioactive decay sources. For commissioning, the fully assembled system measured fission products from spontaneous fission of {sup 252}Cf. Individual measurement resolutions were met for time-of-flight (250 ps FWHM), spacial resolution (2 mm FHWM), and energy (92 keV FWHM for 8.376 MeV). Mass yield results measured from {sup 252}Cf spontaneous fission products are reported from an E–v measurement.

  6. Measurement of Fission Product Yields from Fast-Neutron Fission

    Science.gov (United States)

    Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

    2014-09-01

    One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

  7. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    This is the eleventh issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS

  8. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

    The understanding of fission product (FP) behaviour in severe accidents is important for source term assessment and accident mitigation measures. For example in accident management the operator needs to know the effect of different actions on the behaviour and release of fission products. At VTT fission product behaviour have been studied in different national and international projects. In this presentation the results of projects in EU funded 4th framework programme Nuclear Fission Safety 1994-1998 are reported. The projects are: fission product vapour/aerosol chemistry in the primary circuit (FI4SCT960020), aerosol physics in containment (FI4SCT950016), revaporisation of test samples from Phebus fission products (FI4SCT960019) and assessment of models for fission product revaporisation (FI4SCT960044). Also results from the national project 'aerosol experiments in the Victoria facility' funded by IVO PE and VTT Energy are reported

  9. Development of fission Mo production technology

    International Nuclear Information System (INIS)

    The feasibility study is accomplished in this project for the development of fission moly production. The KAERI process proposed for development in KAERI is discussed together with those of the American Cintichem and Russian IPPE, each of which would be plausible for introduction whenever the indigenous development is not much feasible. For the conceptual design of the KAERI irradiation target, analysis method is set up and some preliminary analysis is performed accordingly for the candidate design. To establish chemical process concepts for the afore-mentioned three processes, characteristics, operation conditions, and the management of the generated wastes are investigated. Basic requirements of hotcell facilities for chemical processing and a possible way of utilizing the existing hotcells are discussed in parallel with the counter-measures for the construction of new hotcell facilities. Various conditions of target irradiation for fission moly production in Hanaro are analyzed. Plan for introduction of the relevant technology introduction and for procurement of highly enriched uranium are considered. On the basis of assuming some conditions, the economic feasibility study for fission moly production is also overviewed. (author). 22 refs., 28 tabs., 24 figs

  10. Decay Chain Deduction of Uranium Fission Products.

    Science.gov (United States)

    Guo, Huiping; Tian, Chenyang; Wang, Xiaotian; Lv, Ning; Ma, Meng; Wei, Yingguang

    2016-07-01

    Delayed gamma spectrum is the fingerprint of uranium materials in arms control verification technology. The decay chain is simplified into basic state linear chain and excitation state linear chain to calculate and analyze the delayed gamma spectra of fission products. Formulas of the changing rule for nuclide number before and after zero-time are deduced. The C program for calculating the delayed gamma ray spectra data is constructed, and related experiments are conducted to verify this theory. Through analysis of the delayed gamma counts of several nuclides, the calculated results are found to be consistent with experimental values. PMID:27218290

  11. Rapid Separation of Fission Product 141La

    Institute of Scientific and Technical Information of China (English)

    XIA; Wen; YE; Hong-sheng; LIN; Min; CHEN; Ke-sheng; XU; Li-jun; ZHANG; Wei-dong; CHEN; Yi-zhen

    2013-01-01

    141La was separated and purified from fission products in this work for physical measurements aimed at improving the accuracy of its decay parameters.As the impact of 142La and other fission products,cesium(141Cs,142Cs included)was rapid separated from the fission products,141Cs and 142Ba separation was prepared after a cooling time about 25 s when 142Cs decays to daughter 142Ba,141La purification then

  12. Production of fission 131I

    International Nuclear Information System (INIS)

    A method of iodine separation from other radionuclides generated by 235U fission has been developed in order to explore the possibilities to obtain 131I as by-product of the 99Mo routine production in the Ezeiza Atomic Centre. The experiments were designed to remove this element to gas phase, and the recoveries were investigated both with and without carrier addition. High volatilization percentages were achieved in the presence of iodine carrier. Some other alternatives to increase the iodine displacement to the gaseous phase, namely vacuum distillation, addition of hydrogen peroxide and use of a carrier gas, were also studied. The method developed, which employs a carrier gas stream, without carrier addition, allows the recovery of about 97% of the 131I, with high specific activity, in a simple and clean way. (author)

  13. Fission-product retention in HTGR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed.

  14. Fission product retention in HTGR fuels

    International Nuclear Information System (INIS)

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed

  15. Aerosols and fission product transport

    International Nuclear Information System (INIS)

    A survey is presented of current knowledge of the possible role of aerosols in the consequences of in- and out-of-core LOCAs and of end fitting failures in CANDU reactors. An extensive literature search has been made of research on the behaviour of aerosols in possible accidents in water moderated and cooled reactors and the results of various studies compared. It is recommended that further work should be undertaken on the formation of aerosols during these possible accidents and to study their subsequent behaviour. It is also recommended that the fission products behaviour computer code FISSCON II should be re-examined to determine whether it reflects the advances incorporated in other codes developed for light water reactors which have been extensively compared. 47 refs

  16. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    This is the ninth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The eighth issue of this series has been published in July 1982 as INDC(NDS)-130. The present issue includes contributions which were received by NDS between 1 August 1982 and 25 June 1983

  17. Phebus FP: fission product behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Lewi, J.; Schwarz, M. [Inst. de Protection et de Surete Nucleaire (IPSN), Dept. de Recherche en Securite, Saint Paul les Durance (France); Hardt, P. von der [European Commission, Joint Research Center, Inst. for Systems, Informatics and Safety (Isis), Saint Paul les Durance (France)

    1998-02-01

    The ongoing Phebus FP programme is the centrepiece of a wide international co-operation investigating, through a series of six integral in-pile experiments, key-phenomena involved in the progression of a postulated severe accident in a Light Wate Reactor (LWR). The Phebus facility offers the capability to study the degradation of real core material, from the early phase of cladding oxidation and hydrogen production up to the late phase of melt progression and molten pool formation. The subsequent release of fission products and structural materials is also experimentally studied, including their transport in the cooling system, and their deposition in the containment, under representative physicochemical conditions. The volatility of iodine in the containment is in particular receiving a special interest in the first experiments, as large uncertainties related to its modelling subsist. FPT-0 and FPT-1, performed respectively in December 1993 and July 1996, have reached very advanced states of degradation, comparable to what was observed in TMI-2, and generated a wealth of results on core degradation and fission product behaviour in particular, pool formation was obtained for a temperature well below the melting point of (U, Zr) O{sub 2} and volatile forms of iodine were detected in the containment much earlier than expected. The resulting database is used to develop and validate the computer codes used to assess the safety of the currently operating plants, to check the efficiency of accident management procedures and also support the design of future plants as EPR. (orig.) [Deutsch] Das laufende Phebus-FP-Programm ist das Herzstueck einer weiten internationalen Zusammenarbeit, durch eine Serie von sechs realitaetsnahen Experimenten die Schluesselphaenomene zu erforschen, die fuer die Ausbreitung eines unterstellten schweren Unfalls in einem Leichtwasserreaktor (LWR) verantwortlich sind. Die Phebus-Anlage in Cadarache ermoeglicht die Untersuchung der Veraenderung

  18. Actinide and fission product separation and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-07-01

    The first international information exchange meeting on actinide and fission product separation and transmutation, took place in Mito in Japan, on 6-8 November 1990. It starts with a number of general overview papers to give us some broad perspectives. Following that it takes a look at some basic facts about physics and about the quantities of materials it is talking about. Then it proceeds to some specific aspects of partitioning, starting with evolution from today commercially applied processes and going on to other possibilities. At the end of the third session it takes a look at the significance of partitioning and transmutation of actinides before it embarks on two sessions on transmutation, first in reactors and second in accelerators. The last session is designed to throw back into the discussion the main points which need to be looked at when considering future work in this area. (A.L.B.)

  19. Development of fission Mo-99 production technology

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

    2000-05-01

    Fission Mo-99 is the only parent nuclide of Tc-99m, an extremely useful tool for mdeical diagnosis, with an estimated usage of greater than 80% of nuclear medicine applicatons. HEU and LEU targets to optimize in HANARO irradiation condition suggested and designed for domestic production of fission Mo-99. The optimum process conditions are established in each unit process to meet quality requirements of fission Mo-99 products, and the results of performance test in combined process show Mo separation and purification yield of the above 97%. The concept of Tc generator production process is established, and the result of performance test show Tc production yield of 98.4% in Tc generator procuction process. The drafts is prepared for cooperation of technical cooperation and business investment with foreign country. Evaluation on economic feasibility is accompanied for fission Mo-99 and Tc-99m generator production.

  20. Development of fission Mo-99 production technology

    International Nuclear Information System (INIS)

    Fission Mo-99 is the only parent nuclide of Tc-99m, an extremely useful tool for mdeical diagnosis, with an estimated usage of greater than 80% of nuclear medicine applicatons. HEU and LEU targets to optimize in HANARO irradiation condition suggested and designed for domestic production of fission Mo-99. The optimum process conditions are established in each unit process to meet quality requirements of fission Mo-99 products, and the results of performance test in combined process show Mo separation and purification yield of the above 97%. The concept of Tc generator production process is established, and the result of performance test show Tc production yield of 98.4% in Tc generator procuction process. The drafts is prepared for cooperation of technical cooperation and business investment with foreign country. Evaluation on economic feasibility is accompanied for fission Mo-99 and Tc-99m generator production

  1. The chemistry of the fission products

    International Nuclear Information System (INIS)

    This is a review of chemistry of some chemical elements in fission products. The elements mentioned are krypton, xenon, rubidium, caesium, silver, strontium, barium, cadmium, rare earth elements, zirconium, niobium, antimony, molybdenum, tellurium, technetium, bromine, iodine, ruthenium, rhodium and palladium. The chemistry of elements and their oxides is briefly given together with the chemical species in aqueous solution. The report also contains tables of the physical properties of the elements and their oxides, of fission products nuclides with their half-life and fission yields and of the permissible concentrations. (author)

  2. Electron spectra from decay of fission products

    Energy Technology Data Exchange (ETDEWEB)

    Dickens, J K

    1982-09-01

    Electron spectra following decay of individual fission products (72 less than or equal to A less than or equal to 162) are obtained from the nuclear data given in the compilation using a listed and documented computer subroutine. Data are given for more than 500 radionuclides created during or after fission. The data include transition energies, absolute intensities, and shape parameters when known. An average beta-ray energy is given for fission products lacking experimental information on transition energies and intensities. For fission products having partial or incomplete decay information, the available data are utilized to provide best estimates of otherwise unknown decay schemes. This compilation is completely referenced and includes data available in the reviewed literature up to January 1982.

  3. Separation of fission molybdenum for the production of technetium generators

    International Nuclear Information System (INIS)

    There are two basically different methods for Mo-99 production: Activation of Mo-98 contained at about 24% in natural isotopic mixtures. Mo-98 enriched targets are irradiated in high-flux reactors in order to achieve the highest possible specific acitivity of the product. Isolation of fission molybdenum from irradiated nuclear fuel targets which have undergone short-term cooling. Maximum fission yields can be attained by irradiation of uranium-235 with the highest possible enrichment. On account of its approximately 1000 times higher specific activity, fission molybdenum has almost replaced worldwide the product fabricated by activation. However, fission molybdenum-99 production has as its prerequisite a suitably advanced technology by which the production process taking place under high activity conditions can be controlled. An integral part of the process consists in the retention of the fission gases the recycling of non-consumed nuclear fuel, and the treatment of the waste streams arising. Ths publication will deal with the individual steps in the process. (orig.)

  4. Fission product release and thermal behaviour

    International Nuclear Information System (INIS)

    Release of fission products from the fuel matrix is an important aspect in relation to performance and safety evaluations. Of particular importance amongst fission products are the isotopes of iodine for radiological considerations and the isotopes of xenon and krypton for fuel thermal behaviour. It is believed that the main mechanism for fission gas release is diffusion but the magnitudes of the relevant diffusion coefficients, which exhibit strong temperature dependences, are not well established. The conductivity of the main gaseous fission product, xenon, is much lower than that of the fill gas helium and hence fission gas release may lead to a deterioration of the fill gas conductivity resulting in higher fuel temperatures and consequently higher fission product release. The two effects, thermal response of fuel to fill gas composition and fission gas/product release are thus intimately connected and have been investigated in a number of instrumented fuel assemblies in the Halden reactor. In such an assembly, the instrumentation includes fuel centre thermocouples, pressure sensors and neutron detectors. In addition pins in the assembly may be swept, whilst at power, with various gases, for example Xe, He or Ar or mixtures thereof. A gamma spectrometer is incorporated into the gas circuit to facilitate the performance of on-line fission product release measurements. At various stages in the lifetime of the assembly thermal tests and fission product release measurements have been made. At low operating temperatures and up to moderate burn-ups, no major fuel restructuring phenomena have been observed and consequently the fission product release has remained at low level dictated by the exposed surfaces of the fuel. Axial gas flow measurements indicate that fuel cracking and irreversible relocation occurred as early as the first ramps to power. The processes have continued throughout life and an absence of any change in response pressurization tests indicates that

  5. Fission Product Sorptivity in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Tompson, Jr., Robert V. [Univ. of Missouri, Columbia, MO (United States); Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States); Ghosh, Tushar [Univ. of Missouri, Columbia, MO (United States); Viswanath, Dabir [Univ. of Missouri, Columbia, MO (United States); Walton, Kyle [Univ. of Missouri, Columbia, MO (United States); Haffner, Robert [Univ. of Missouri, Columbia, MO (United States)

    2015-04-01

    Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one

  6. Actinide and fission product separation and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-07-01

    The second international information exchange meeting on actinide and fission product separation and transmutation, took place in Argonne National Laboratory in Illinois United States, on 11-13 November 1992. The proceedings are presented in four sessions: Current strategic system of actinide and fission product separation and transmutation, progress in R and D on partitioning processes wet and dry, progress in R and D on transmutation and refinements of neutronic and other data, development of the fuel cycle processes fuel types and targets. (A.L.B.)

  7. Chemistry of actinides and fission products

    International Nuclear Information System (INIS)

    This task is concerned primarily with the fundamental chemistry of the actinide and fission product elements. Special efforts are made to develop research programs in collaboration with researchers at universities and in industry who have need of national laboratory facilities. Specific areas currently under investigation include: (1) spectroscopy and photochemistry of actinides in low-temperature matrices; (2) small-angle scattering studies of hydrous actinide and fission product polymers in aqueous and nonaqueous solvents; (3) kinetic and thermodynamic studies of complexation reactions in aqueous and nonaqueous solutions; and (4) the development of inorganic ion exchange materials for actinide and lanthanide separations. Recent results from work in these areas are summarized here

  8. Actinide and fission product separation and transmutation

    International Nuclear Information System (INIS)

    The second international information exchange meeting on actinide and fission product separation and transmutation, took place in Argonne National Laboratory in Illinois United States, on 11-13 November 1992. The proceedings are presented in four sessions: Current strategic system of actinide and fission product separation and transmutation, progress in R and D on partitioning processes wet and dry, progress in R and D on transmutation and refinements of neutronic and other data, development of the fuel cycle processes fuel types and targets. (A.L.B.)

  9. Fission product decay heat for thermal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dickens, J. K.

    1979-01-01

    In the past five years there have been new experimental programs to measure decay heat (i.e., time dependent beta- plus gamma-ray energy release rates from the decay of fission products) following thermal-neutron fission of /sup 235/U, /sup 239/Pu, and /sup 241/Pu for times after fission between 1 and approx. 10/sup 5/ sec. Experimental results from the ORNL program stress the very short times following fission, particularly in the first few hundred sec. Complementing the experimental effort, computer codes have been developed for the computation of decay heat by summation of calculated individual energies released by each one of the fission products. By suitably combining the results of the summation calculations with the recent experimental results, a new Decay Heat Standard has been developed for application to safety analysis of operations of light water reactors. The new standard indicates somewhat smaller energy release rates than those being used at present, and the overall uncertainties assigned to the new standard are much smaller than those being used at present.

  10. Fission Product Decay Heat Calculations for Neutron Fission of 232Th

    Science.gov (United States)

    Son, P. N.; Hai, N. X.

    2016-06-01

    Precise information on the decay heat from fission products following times after a fission reaction is necessary for safety designs and operations of nuclear-power reactors, fuel storage, transport flasks, and for spent fuel management and processing. In this study, the timing distributions of fission products' concentrations and their integrated decay heat as function of time following a fast neutron fission reaction of 232Th were exactly calculated by the numerical method with using the DHP code.

  11. Actinide and fission product partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The third international information exchange meeting on actinide and fission product partitioning and transmutation, took place in Cadarache France, on 12-14 December 1994. The proceedings are presented in six sessions : an introduction session, the major programmes and international cooperation, the systems studies, the reactors fuels and targets, the chemistry and a last discussions session. (A.L.B.)

  12. Actinide and fission product partitioning and transmutation

    International Nuclear Information System (INIS)

    The third international information exchange meeting on actinide and fission product partitioning and transmutation, took place in Cadarache France, on 12-14 December 1994. The proceedings are presented in six sessions : an introduction session, the major programmes and international cooperation, the systems studies, the reactors fuels and targets, the chemistry and a last discussions session. (A.L.B.)

  13. Model for fission-product calculations

    International Nuclear Information System (INIS)

    Many fission-product cross sections remain unmeasurable thus considerable reliance must be placed upon calculational interpolation and extrapolation from the few available measured cross sections. The vehicle, particularly for the lighter fission products, is the conventional optical-statistical model. The applied goals generally are: capture cross sections to 7 to 10% accuracies and inelastic-scattering cross sections to 25 to 50%. Comparisons of recent evaluations and experimental results indicate that these goals too often are far from being met, particularly in the area of inelastic scattering, and some of the evaluated fission-product cross sections are simply physically unreasonable. It is difficult to avoid the conclusion that the models employed in many of the evaluations are inappropriate and/or inappropriately used. In order to alleviate the above unfortunate situations, a regional optical-statistical (OM) model was sought with the goal of quantitative prediction of the cross sections of the lighter-mass (Z = 30-51) fission products. The first step toward that goal was the establishment of a reliable experimental data base consisting of energy-averaged neutron total and differential-scattering cross sections. The second step was the deduction of a regional model from the experimental data. It was assumed that a spherical OM is appropriate: a reasonable and practical assumption. The resulting OM then was verified against the measured data base. Finally, the physical character of the regional model is examined

  14. A Covariance Generation Methodology for Fission Product Yields

    Directory of Open Access Journals (Sweden)

    Terranova N.

    2016-01-01

    Full Text Available Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1 no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation, developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

  15. A Covariance Generation Methodology for Fission Product Yields

    Science.gov (United States)

    Terranova, N.; Serot, O.; Archier, P.; Vallet, V.; De Saint Jean, C.; Sumini, M.

    2016-03-01

    Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1) no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM) implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation), developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

  16. Fission 2009 4. International Workshop on Nuclear Fission and Fission Product Spectroscopy - Compilation of slides

    International Nuclear Information System (INIS)

    This conference is dedicated to the last achievements in experimental and theoretical aspects of the nuclear fission process. The topics include: mass, charge and energy distribution, dynamical aspect of the fission process, nuclear data evaluation, quasi-fission and fission lifetime in super heavy elements, fission fragment spectroscopy, cross-section and fission barrier, and neutron and gamma emission. This document gathers the program of the conference and the slides of the presentations

  17. Development of fission Mo-99 production technology

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

    2001-05-01

    This R and D project is planed to supply domestic demands of Mo-99 through fission route, and consequently this project will be expected to rise up utilization of HANARO and KAERI's capability for marketing extension into domestic and oversea radiopharmaceutical market. HEU and LEU target types are decided and designed for fission Mo-99 production in domestic. Experimental study of target fabrication technology was performed and developed processing equipments. And conceptual design of target loading/unloading in/from HANARO device are performed. Tracer test of Mo-99 separation and purification process was performed, test results reach to Mo-99 recovery yield above 80% and decontamination factor above 1600. Combined Mo-99 separation and purification process was decided for hot test scheduled from next year, and performance test was performed. Conceptual design for modification of existing hot cell for fission Mo-99 production facility was performed and will be used for detail design. Assumption for the comparison of LEU and HEU target in fission Mo-99 production process were suggested and compared of merits and demerits in view of fabrication technology and economy feasibility.

  18. Actinide and fission product partitioning and transmutation

    International Nuclear Information System (INIS)

    The fourth international information exchange meeting on actinide and fission product partitioning and transmutation, took place in Mito City in Japan, on 111-13 September 1996. The proceedings are presented in six sessions: the major programmes and international cooperation, the partitioning and transmutation programs, feasibility studies, particular separation processes, the accelerator driven transmutation, and the chemistry of the fuel cycle. (A.L.B.)

  19. Actinide and fission product partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The fourth international information exchange meeting on actinide and fission product partitioning and transmutation, took place in Mito City in Japan, on 111-13 September 1996. The proceedings are presented in six sessions: the major programmes and international cooperation, the partitioning and transmutation programs, feasibility studies, particular separation processes, the accelerator driven transmutation, and the chemistry of the fuel cycle. (A.L.B.)

  20. Time dependent particle emission from fission products

    Energy Technology Data Exchange (ETDEWEB)

    Holloway, Shannon T [Los Alamos National Laboratory; Kawano, Toshihiko [Los Alamos National Laboratory; Moller, Peter [Los Alamos National Laboratory

    2010-01-01

    Decay heating following nuclear fission is an important factor in the design of nuclear facilities; impacting a variety of aspects ranging from cooling requirements to shielding design. Calculations of decay heat, often assumed to be a simple product of activity and average decay product energy, are complicated by the so called 'pandemonium effect'. Elucidated in the 1970's this complication arises from beta-decays feeding high-energy nuclear levels; redistributing the available energy between betas and gammas. Increased interest in improving the theoretical predictions of decay probabilities has been, in part, motivated by the recent experimental effort utilizing the Total Absorption Gamma-ray Spectrometer (TAGS) to determine individual beta-decay transition probabilities to individual nuclear levels. Accurate predictions of decay heating require a detailed understanding of these transition probabilities, accurate representation of particle decays as well as reliable predictions of temporal inventories from fissioning systems. We will discuss a recent LANL effort to provide a time dependent study of particle emission from fission products through a combination of Quasiparticle Random Phase Approximation (QRPA) predictions of beta-decay probabilities, statistical Hauser-Feshbach techniques to obtain particle and gamma-ray emissions in statistical Hauser-Feshbach and the nuclear inventory code, CINDER.

  1. Gas-phase transport of fission products

    International Nuclear Information System (INIS)

    The paper presents the results of an experimental investigation to show the importance of nuclear aerosol formation as a mechanism for semi-volatile fission product transport under certain postulated HTGR accident conditions. Simulated fission product Sr and Ba as oxides are impregnated in H451 graphite and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperatures. Increasing carrier-gas flow rate greatly enhances the extent of particulate transport. The release and transport of simulated fission product Ag as metal are also investigated. Electron microscopic examinations of the collected Sr and Ag aerosols show large agglomerates composed of primary particles roughly 0.06 to 0.08 μm in diameter

  2. Fission products uptake into the body of farm animals and radionuclide transilocation into farm animal products

    International Nuclear Information System (INIS)

    The effects produced by fission products in the organism of dairy cattle are considered. The presence of radionuclides in the animal body leads to the danger of radioactive contamination of milk and may result in the loss of milk productivity and reproductive capacity of the irradiated animals or even in their death. Of greatest practical interest is an experimental evaluation of the intake of fission products, since, on the one hand, milk is one of the basic and valuable foods and, on the other, it is the principal source of biologically dangerous radionuclides in human diet

  3. SPIDER Progress Towards High Resolution Correlated Fission Product Data

    Science.gov (United States)

    Shields, Dan; Meierbachtol, Krista; Tovesson, Fredrik; Arnold, Charles; Blackeley, Rick; Bredeweg, Todd; Devlin, Matt; Hecht, Adam; Jandel, Marian; Jorgenson, Justin; Nelson, Ron; White, Morgan; Spider Team

    2014-09-01

    The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) is under development with the goal of obtaining high-resolution, high-efficiency, correlated fission product data needed for many applications including the modeling of next generation nuclear reactors, stockpile stewardship, and the fundamental understanding of the fission process. SPIDER simultaneously measures velocity and energy of both fission products to calculate fission product yields (FPYs), neutron multiplicity (ν), and total kinetic energy (TKE). A detailed description of the prototype SPIDER detector components will be presented. Characterization measurements with alpha and spontaneous fission sources will also be discussed. LA-UR-14-24875. The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) is under development with the goal of obtaining high-resolution, high-efficiency, correlated fission product data needed for many applications including the modeling of next generation nuclear reactors, stockpile stewardship, and the fundamental understanding of the fission process. SPIDER simultaneously measures velocity and energy of both fission products to calculate fission product yields (FPYs), neutron multiplicity (ν), and total kinetic energy (TKE). A detailed description of the prototype SPIDER detector components will be presented. Characterization measurements with alpha and spontaneous fission sources will also be discussed. LA-UR-14-24875. This work is in part supported by LANL Laboratory Directed Research and Development Projects 20110037DR and 20120077DR.

  4. Progress in fission product nuclear data. No. 14

    International Nuclear Information System (INIS)

    This is the 14th issue of a report series on Fission Product Nuclear Data published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of fission product yields, neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data from neutron induced and spontaneous fission, lumped fission product data. The first part of the report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. The third part contains requirements for further measurements

  5. Production techniques of fission 99Mo

    International Nuclear Information System (INIS)

    Generally two different techniques are available for molybdenum-99 production for use in medical technetium-99 generation. The first one is based on neutron irradiation of molybdenum targets of natural isotopic composition or enriched in molybdenum-98. In these cases the Mo-99 is generated via the nuclear reaction 98Mo (n,γ) 99Mo. Although this process can be carried out at low expenditure it gives a product of low specific activity and, hence, restricted applicability. In a second process Mo-99 is obtained as a result of the neutron induced fission of U-235 according to 235U (n,f) 99Mo. This technique provides a product with a specific activity several orders of magnitude higher than that obtained from the 98Mo (n,γ) 99Mo nuclear reaction and perhaps even more important up to several thousands curies of Mo-99 per production run. In this paper a modern production procedure of Mo-99 via the fission reaction, which was developed at the Institute of Radiochemistry of the Nuclear Research Centre Karlsruhe will be described. The targeting, irradiation of U-235, the separation and purification steps involved as well as the recycling of the non-converted U-235, which should be a major consideration in any production technique, will be discussed. (author). 24 refs, 14 figs, 1 tab

  6. Geochemistry of actinides and fission products in natural aquifer systems

    International Nuclear Information System (INIS)

    The progress in the research area of the community project MIRAGE: 'Geochemistry of actinides and fission products in natural aquatic systems' has been reviewed. This programme belongs to a specific research and technical development programme for the European Atomic Energy Community in the field of management and storage of radioactive waste. The review summarizes research progresses in subject areas: complexation with organics, colloid generation in groundwater and basic retention mechanisms in the framework of the migration of radionuclides in the geosphere. The subject areas are being investigated by 23 laboratories under interlaboratory collaborations or independent studies. (orig.)

  7. Fission product and aerosol behaviour within the containment

    International Nuclear Information System (INIS)

    Experimental studies have been undertaken to characterise the behaviour of fission products in the containment of a pressurised water reactor during a severe accident. The following aspects of fission product transport have been studied: (a) aerosol nucleation, (b) vapour transport processes, (c) chemical forms of high-temperature vapours, (d) interaction of fission product vapours with aerosols generated from within the reactor core, (e) resuspension processes, (f) chemistry in the containment. (author)

  8. Basic Physics Data: Measurement of Neutron Multiplicity from Induced Fission

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, Sara [Univ. of Michigan, Ann Arbor, MI (United States); Haight, Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-05-04

    From October 1 to October 17 a team of researchers from UM visited the LANSCE facility for an experiment during beam-time allotted from October 4 to October 17. A total of 24 detectors were used at LANSCE including liquid organic scintillation detectors (EJ-309), NaI scintillation detectors, and Li-6 enriched glass detectors. It is a double time-offlight (TOF) measurement using spallation neutrons generated by a target bombarded with pulsed high-energy protons. The neutrons travel to an LLNL-manufactured parallel plate avalanche chamber (PPAC) loaded with thin U-235 foils in which fission events are induced. The generated fission neutrons and photons are then detected in a detector array designed and built at UM and shipped to LANSCE. Preparations were made at UM, where setup and proposed detectors were tested. The UM equipment was then shipped to LANSCE for use at the 15L beam of the weapons neutron research (WNR) facility.

  9. Ceramic Hosts for Fission Products Immobilization

    International Nuclear Information System (INIS)

    Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent

  10. Ceramic Hosts for Fission Products Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Peter C Kong

    2010-07-01

    Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent

  11. Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Bhatia, C.; Fallin, B. [Department of Physics, Duke University, Durham, NC 27708 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708 (United States); Gooden, M.E., E-mail: megooden@tunl.duke.edu [Triangle Universities Nuclear Laboratory, Durham, NC 27708 (United States); Department of Physics, North Carolina State University, Raleigh, NC 27605 (United States); Howell, C.R. [Department of Physics, Duke University, Durham, NC 27708 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708 (United States); Kelley, J.H. [Triangle Universities Nuclear Laboratory, Durham, NC 27708 (United States); Department of Physics, North Carolina State University, Raleigh, NC 27605 (United States); Tornow, W. [Department of Physics, Duke University, Durham, NC 27708 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708 (United States); Arnold, C.W.; Bond, E.M.; Bredeweg, T.A.; Fowler, M.M.; Moody, W.A.; Rundberg, R.S.; Rusev, G.; Vieira, D.J.; Wilhelmy, J.B. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Becker, J.A.; Macri, R.; Ryan, C.; Sheets, S.A.; Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); and others

    2014-09-01

    A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

  12. Fuel rod internal chemistry and fission products behaviour

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the members of the International Working Group on Water Reactor Fuel Performance and Technology. Forty-six participants representing fourteen countries and two international organizations attended the meeting. Twenty-one presentations were discussed in four sessions: thermodynamics of fission products (six papers); fission products migration and release (seven papers); fission product release in transients or accident conditions (four papers); fission products to cladding interaction - stress corrosion cracking (five papers). A separate abstract was prepared for all twenty-one papers

  13. Observation of attachment ratio of fission products on solution aerosol

    International Nuclear Information System (INIS)

    Attachment behavior of fission products to solution aerosols has been observed to elucidate the role of chemical effects in the generation mechanism of fissionproduct aerosols. Primary aerosols generated from aqueous solution of sodium chloride or ammonium sulfate were passed through a fission-product chamber, and radioactive aerosols were generated by attaching fission products to the primary aerosol particles. Attachment ratios of the fission products on aerosols were estimated from activity measurements. It was found that the attachment ratio of the sodium chloride solution aerosol is larger than that of the ammonium sulfate solution aerosol. (author)

  14. Resuspension of fission products from sump water

    International Nuclear Information System (INIS)

    Resuspension of fission products from the boiling sump in the container has long been known as a source of airborne radioactivity. Since this source is very weak, however, not much attention had been paid to it as long as radiological source terms were governed by stronger sources. Recently, the continuous reduction of source terms and the introduction of accident management measures led to a situation where weak but longlasting sources of radioactivity may become important, either as a contribution to the radiological sources term or as an impact to accident filtration systems. Existing data on resuspension from boiling contaminated water all suffered from two deficiencies: they were measured under conditions unlike those in a reactor accident and they scattered over more than two orders of magnitude. In a precursor study this uncertainty was considered to be too large to use the data for source term calculations. A later experimental research programme REST (REsuspension Source Term) was carried out at the Laboratorium fuer Aerosolphysik und Filtertechnik (LAF), Kernforschungszentrum Karlsruhe (KfK). The programme was supported by the Commission of the European Communities Ispra, under Contract No 3009-86-07 ELISPD in the framework of the shared-cost action programme on reactor safety. The investigations started in 1987 and ended in 1990. The objectives of the REST programme were to measure resuspension source characteristics under simulated accident conditions such that an application of the data in fission product transport and depletion models is possible

  15. Operation of plant to produce Mo-99 from fission products

    International Nuclear Information System (INIS)

    As it is well known, the production of Mo-99/Tc-99m generators has an outstanding place in radioisotope programs of the Argentine National Atomic Energy Commission. The basic raw material is Mo-99 from fission of U-235. In 1985 the production plant of this radionuclide began to operate, according to an adaptation of the method that was developed in Kernforschungszentrum Karlsruhe. The present work describes the target irradiation conditions in the reactor RA-3 (mini plates of U/Al alloy with 90% enriched uranium), the flow diagram and the operative conditions of the production process. The containment, filtration and removal conditions of the generated fission gases and the disposal of liquid and solid wastes are also analyzed. On the basis of the experience achieved in the development of more than twenty production processes, process efficiency is analyzed, taking into account the theoretical evaluation resulting from the application of the computer program 'Origin'(ORML) to the conditions of our case. The purity characteristics of the final product are reported (Zr-95 0,1 ppm; Nb-95 1 ppm; Ru-103 20 ppm; I-131 10 ppm) as well as the chemical characteristics that make it suitable to be used in the production of Mo-99/I c-99m generators. (Author)

  16. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) is currently working toward the licensing of a methodology for using actinide-only burnup credit for the transportation of spent nuclear fuel (SNF). Important margins are built into this methodology. By using comparisons with a representative experimental database to determine bias factors, the methodology ensures that actinide concentrations and worths are estimated conservatively; furthermore, the negative net reactivity of certain actinides and all fission products (FPs) is not taken into account, thus providing additional margin. A future step of DOE's effort might aim at establishing an actinide and FP burnup credit methodology. The objective of this work is to establish the uncertainty to be applied to the total FP worth in SNF. This will serve two ends. First, it will support the current actinide-only methodology by demonstrating the margin available from FPs. Second, it will identify the major contributions to the uncertainty and help set priorities for future work

  17. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D. [Argonne National Lab., IL (United States)

    1996-10-01

    The goal of this project is to develop and demonstrate the feasibility of a novel low-temperature solidification/stabilization (S/S) technology for immobilizing waste streams containing fission products such as cesium, strontium, and technetium in a chemically bonded phosphate ceramic. This technology can immobilize partitioned tank wastes and decontaminate waste streams containing volatile fission products.

  18. ESF system fission product retention effectiveness

    International Nuclear Information System (INIS)

    The objective of this research is to develop validated analytical models for use in estimating fission product retention effectiveness of light water reactor (LWR) engineered safety feature (ESF) systems. Program planning is directed toward reducing the highest priority uncertainties in severe accident/source term phenomena. Candidate ESF systems include spray, suppression pool, containment cooler, and containment and auxiliary air cleaning systems as well as the ice compartments of ice condenser containment systems. The work involves identifying, planning, and conducting experiments needed to validate models and providing guidelines for system design and operating and maintenance requirements. It also includes developing information that will not only identify the most important systems but will permit these systems to be emphasized in future regulatory processes. During FY 1987 work focused on activities related to the validation of the ICEDF and SPARC computer codes. The codes were developed at the Pacific Northwest Laboratory (PNL) to estimate the extent of fission product retention in the ice compartments of pressurized water reactor (PWR) ice condenser containment systems and boiling water reactor (BWR) suppression pools. Scope of the efforts related to ICEDF code validation ranged from construction of an engineering-scale test facility to the subsequent use of the facility for the conduct of experiments to obtain data for comparison with code calculations. Validation efforts associated with the SPARC code involved comparison of calculations from a modified version of the code with data from tests sponsored by the Electric Power Research Institute (EPRI). The ICEDF and SPARC codes were also used in support of several major NRC activities in FY 1987

  19. Fission product and aerosol behaviour within the containment

    International Nuclear Information System (INIS)

    Experimental studies have been undertaken to characterise the behaviour of fission products in the containment of a pressurised water reactor during a severe accident. The following aspects of fission product transport have been studied: (a) aerosol nucleation, (b) vapour transport processes, (c) chemical forms of high-temperature vapours, (d) interaction of fission product vapours with aerosols generated from within the reactor core, (e) resuspension processes, (f) chemistry in the containment. Chemical effects have been shown to be important in defining and quantifying fission product source terms in a wide range of accident sequences. Both the chemical forms of the fission product vapours and their interactions with reactor materials aerosols could have a major effect on the magnitude and physicochemical forms of the radioactive emission from a severe reactor accident. Only the main conclusions are presented in this summary document; detailed technical aspects of the work are described in separate reports listed in the annex

  20. The distribution and behavior of fission products inside the containment

    International Nuclear Information System (INIS)

    Following accident scenarios resulting in core melt and failure of reactor pressure vessel, the molten core debris will be ejected from the vessel by the process of high pressure melt ejection or relocation by gravity to the reactor cavity. After the ejection of the fission products laden molten core debris, the fission products will be released and distributed to the containment atmosphere. Noble gases and other high-volatile fission products, such as Xe, I, Cs, and Te, contained in the molten core debris will be released completely to the containment, while the more refractory fission products, which include lanthanides and actinides (Sr, Ba, Ru, La) will be partially released. Fission products are distributed in the containment atmosphere in the forms of gases, aerosols, particles, and deposition on surfaces and water pools

  1. Characteristics of fission product release from a molten pool

    Energy Technology Data Exchange (ETDEWEB)

    Yun, J.I.; Suh, K.Y.; Kang, C.S. [Seoul National Univ., Dept. of Nuclear Engineering (Korea, Republic of)

    2001-07-01

    The volatile fission products are released from the debris pool, while the less volatile fission products tend to remain as condensed phases because of their low vapor pressure. The release of noble gases and the volatile fission products is dominated by bubble dynamics. The release of the less volatile fission products from the pool can be analyzed based on mass transport through a liquid with the convection flow. The physico-numerical models were orchestrated from existing submodels in various disciplines of engineering to estimate the released fraction of fission products from a molten pool. It was assumed that the pool has partially filled hemispherical geometry. For the high pool pressure, the diameter of the bubbles at detachment was calculated utilizing the Cole and Shulman correlation with the effect of system pressure. Sensitivity analyses were performed and results of the numerical calculations were compared with analysis results for the TMI-2 accident. (author)

  2. Fission Product Release from Spent Nuclear Fuel During Melting

    International Nuclear Information System (INIS)

    The Melt-Dilute process consolidates aluminum-clad spent nuclear fuel by melting the fuel assemblies and diluting the 235U content with depleted uranium to lower the enrichment. During the process, radioactive fission products whose boiling points are near the proposed 850 degrees C melting temperature can be released. This paper presents a review of fission product release data from uranium-aluminum alloy fuel developed from Severe Accident studies. In addition, scoping calculations using the ORIGEN-S computer code were made to estimate the radioactive inventories in typical research reactor fuel as a function of burnup, initial enrichment, and reactor operating history and shutdown time.Ten elements were identified from the inventory with boiling points below or near the 850 degrees C reference melting temperature. The isotopes 137Cs and 85Kr were considered most important. This review serves as basic data to the design and development of a furnace off-gas system for containment of the volatile species

  3. distribution of Release Fission Products Through the Nuclear Reactor Site

    International Nuclear Information System (INIS)

    Through the operation of nuclear reactors, radioactive fission products could be release to the environment as a result of severe accidents e.g. Chernobyl accident. Estimation of the atmospheric dispersion, distribution and transport of the radioactive fission products is essential to assessment of the risk to the public from such accidents. In this work, the polluted plume is treated as a matrix of isolated particles.These particles are the fission product isotopes, which compose the radioactive plume.The fission products were classified depending on its half live into three category, long-lived, medium lived and small half-life.The normalized concentrations of the fission product isotopes in the radioactive plume were calculated.The travel time (the time elapsed from the released instant till the deposited time) of each fission products was calculated. The area around the nuclear reactor stack was divided into different zones, started from the reactor stack position until 5 km.The deposited radioactive fission products in each zone was estimated.The calculations were done using the spherical Gaussian plume model

  4. Estimate of exposure impact to fission product transformation

    International Nuclear Information System (INIS)

    Materials on transmutation radionuclides - method of processing of radionuclides, which recently acquires a greater importance for the countries developing nuclear power engineering are presented. characterization of waste products of nuclear power engineering and a forecast of their accumulation during operation of atomic power station are made. Streams are estimated and the choice of response of action of irradiation on transformation of fission products is discussed. Some regularities are considered during utilization of fission debris in neutron and gamma-fields. Feasibility of creation of powerful neutron fields and principles of transformation of fission products and other radionuclides in neutron fields are discussed. 21 refs., 2 tab., 14 figs

  5. Fission Product Transmutation in Mixed Radiation Fields

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, Frank [Idaho State Univ., Pocatello, ID (United States); Burgett, Erick [Idaho State Univ., Pocatello, ID (United States); Starovoitova, Valeriia [Idaho State Univ., Pocatello, ID (United States); Tsveretkov, Pavel [Idaho State Univ., Pocatello, ID (United States)

    2015-01-15

    Work under this grant addressed a part of the challenge facing the closure of the nuclear fuel cycle; reducing the radiotoxicity of lived fission products (LLFP). It was based on the possibility that partitioning of isotopes and accelerator-based transmutation on particular LLFP combined with geological disposal may lead to an acceptable societal solution to the problem of management. The feasibility of using photonuclear processes based on the excitation of the giant dipole resonance (GDR) by bremsstrahlung radiation as a cost effective transmutation method was accessed. The nuclear reactions of interest: (γ,xn), (n,γ), (γ,p) can be induced by bremsstrahlung radiation produced by high power electron accelerators. The driver of these processes would be an accelerator that produces a high energy and high power electron beam of ~ 100 MeV. The major advantages of such accelerators for this purpose are that they are essentially available “off the shelf” and potentially would be of reasonable cost for this application. Methods were examined that used photo produced neutrons or the bremsstrahlung photons only, or use both photons and neutrons in combination for irradiations of selected LLFP. Extrapolating the results to plausible engineering scale transmuters it was found that the energy cost for 129I and 99Tc transmutation by these methods are about 2 and 4%, respectively, of the energy produced from 1000MWe.

  6. The Oklo natural reactor: Cumulative fission yields and retentivity of the symmetric mass region fission products

    Science.gov (United States)

    De Laeter, J. R.; Rosman, K. J. R.; Smith, C. L.

    1980-10-01

    Solid source mass spectrometry has been used to determine the relative cumulative fission yields of five elements in three samples of uranium ore from reactor zones in the Oklo mine site. Eighteen fission chains covering the mass range from 105 ≤ A ≤ 130 have been measured for Pd, Ag, Cd, Sn and Te. These measurements have enabled a number of nuclear parameters to be calculated including the relative proportions of 235U, 238U and 239Pu involved in the fission process. The concentration of the five elements in the Oklo samples have also been measured using the stable isotope dilution technique. These values have then been compared to the estimates of the amount of these elements produced by fission under the conditions that are appropriate to the three samples. This procedure enables the retentivity of the elements in the reactor zones to be evaluated. Our work confirms the fact that Pd and Te are retained almost in their entirety in the samples, whereas the other three elements have been partially lost from the reactor site. Almost all the Cd fission products have been lost, and more than 50% of the Ag and Sn fission-produced material has been removed.

  7. Transport of fission products in matrix and graphite

    International Nuclear Information System (INIS)

    In the past years new experimental methods were applied to or developed for the investigation of fission product transport in graphitic materials and to characterization of the materials. Models for fission product transport and computer codes for the calculation of core release rates were improved. Many data became available from analysis of concentration profiles in HTR-fuel elements. New work on the effect on diffusion of graphite corrosion, fast neutron flux and fluence, heat treatment, chemical interactions and helium pressure was reported on recently or was in progress in several laboratories. It seemed to be the right time to discuss the status of transport of metallic fission products in general, and in particular the relationship between structural and transport properties. Following a suggestion a Colloquium was organized at the HMI Berlin. Interdisciplinary discussions were stimulated by only inviting a limited number of participants who work in different fields of graphite and fission product transport research. (orig./RW)

  8. Fission product release from highly irradiated LWR fuel

    International Nuclear Information System (INIS)

    A series of experiments was conducted with highly irradiated light-water reactor fuel rod segments to investigate fission products released in steam in the temperature range 500 to 12000C. (Two additional release tests were conducted in dry air.) The primary objectives were to quantify and characterize fission product release under conditions postulated for a spent-fuel transportation accident and for a successfully terminated loss-of-coolant accident (LOCA). In simulated, controlled LOCA-type tests, release at the time of rupture proved to be more significant than the diffusional release that followed. Comparison of the release data for the dry-air tests with the release data of similarly conducted tests in steam indicated significant increases in the releases of iodine, ruthenium, and cesium in air. Various parameters that affect fission product release are discussed, and experimental observations and analysis of the chemical behavior of releasable fission products in inert, steam, and dry-air atmospheres are examined

  9. Conceptual design report of hot cell modification and process for fission Mo-99 production

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Choung, W. M.; Lee, K. I.; Hwang, D. S.; Kim, Y. K.; Park, K. B.; Jung, Y. J.; Kim, D. S.; Park, Y. C

    2001-05-01

    In this conceptual design report, the basic data and design guides for detail design of fission Mo-99 production process and hot cell modification are included.The basic data and design guides for detail design of fission Mo-99 production process contains following contents. -design capacity, the basic process, process flow diagram, process material balance, process data. The basic data and design guides for modification of existing hot cell contains following contents. - plot plan of hot cell facility, the plan for shield reinforcement of hot cell, the plan for management and storage of high level liquid wastes, the plan of ventilation system, the plan for modification of auxiliary facilities. And also, the results of preliminary safety analysis(normal operation and accidents) and criticality analysis are included in this conceptual design report.

  10. Studies on the Separation of Cesium From Fission Products

    Institute of Scientific and Technical Information of China (English)

    QIANLi-juan; ZHANGSheng-dong; GUOJing-ru; CUIAn-zhi; YANGLei; WUWang-suo

    2003-01-01

    135Cs is a long-life fission product. When measuring its thermal cross section, we must separate radiochemical purity cesium from fission products. Except for decontaminating radio- nuclides, others which can be activated must be avoided to come into solution. So ion exchanger is used. Inorganic ion exchangers have received increased attention because of their high resistance to radiation and their very efficient separation of alkali metal ions.

  11. Spray removal of fission products in PWR containments

    International Nuclear Information System (INIS)

    Models and parameters for assessing the rate and extent of removal of various fission product species are described. A range of droplet sizes and of spray additive options is considered and removal of vapour phase inorganic iodine species, of organic iodides and of aerosols containing fission products is discussed. Aerosol removal is assessed in terms of contributing removal mechanisms and the removal rate modelled as a function of the radius of the aerosol particulate species. (author)

  12. Interactions of fission product vapours with aerosols

    Energy Technology Data Exchange (ETDEWEB)

    Benson, C.G.; Newland, M.S. [AEA Technology, Winfrith (United Kingdom)

    1996-12-01

    Reactions between structural and reactor materials aerosols and fission product vapours released during a severe accident in a light water reactor (LWR) will influence the magnitude of the radiological source term ultimately released to the environment. The interaction of cadmium aerosol with iodine vapour at different temperatures has been examined in a programme of experiments designed to characterise the kinetics of the system. Laser induced fluorescence (LIF) is a technique that is particularly amenable to the study of systems involving elemental iodine because of the high intensity of the fluorescence lines. Therefore this technique was used in the experiments to measure the decrease in the concentration of iodine vapour as the reaction with cadmium proceeded. Experiments were conducted over the range of temperatures (20-350{sup o}C), using calibrated iodine vapour and cadmium aerosol generators that gave well-quantified sources. The LIF results provided information on the kinetics of the process, whilst examination of filter samples gave data on the composition and morphology of the aerosol particles that were formed. The results showed that the reaction of cadmium with iodine was relatively fast, giving reaction half-lives of approximately 0.3 s. This suggests that the assumption used by primary circuit codes such as VICTORIA that reaction rates are mass-transfer limited, is justified for the cadmium-iodine reaction. The reaction was first order with respect to both cadmium and iodine, and was assigned as pseudo second order overall. However, there appeared to be a dependence of aerosol surface area on the overall rate constant, making the precise order of the reaction difficult to assign. The relatively high volatility of the cadmium iodide formed in the reaction played an important role in determining the composition of the particles. (author) 23 figs., 7 tabs., 22 refs.

  13. Data summary report for fission product release test VI-4

    International Nuclear Information System (INIS)

    This was the fourth in a series of high-temperature fission product release tests in a vertical test apparatus. The test specimen, a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium, had been irradiated to a burnup of 47 MWd/kg. In simulation of a severe accident in a light-water reactor, it was heated in hydrogen in a hot cell-mounted test apparatus to a maximum test temperature of 2400 K for a period of 20 min. The released fission products were collected on components designed to facilitate sampling and analysis. On-line radioactivity measurements and posttest inspection revealed that the fuel had partially collapsed at about the time the cladding melted. Based on fission product inventories measured in the fuel or calculated by ORIGEN2, analyses of test components showed total releases from the fuel of 85% for 85Kr, 106Ru, 3.9% for 125Sb, 96% for both 134Cs and 137Cs, and 13% for 154Eu. Large fractions of the released fission products (up to 96% of the 154Eu) were retained in the furnace. Small release fractions for several other fission products -- Rb, Br, Sr, Te, I, and Ba -- were detected also. In addition, very small amounts of fuel material -- uranium and plutonium -- were released. Total mass release from the furnace to the collection system, which included fission products, fuel material, and structural materials, was 0.40g, with 40% of this material being deposited as vapor and 60% of it being collected as aerosols. The results from this test were compared with previous tests in this series and with an in-pile test at similar conditions at Sandia National Laboratories. There was no indication that the mode of heating (fission heat vs radiant heat) significantly affected fission product release. 24 refs., 25 figs., 14 tabs

  14. Long-Lived Fission Product Transmutation Studies

    International Nuclear Information System (INIS)

    A systematic study on long-lived fission products (LLFPs) transmutation has been performed with the aim of devising an optimal strategy for their transmutation in critical or subcritical reactor systems and evaluating impacts on the geologic repository. First, 99Tc and 129I were confirmed to have highest transmutation priorities in terms of transmutability and long-term radiological risk reduction. Then, the transmutation potentials of thermal and fast systems for 99Tc and 129I were evaluated by considering a typical pressurized water reactor (PWR) core and a sodium-cooled accelerator transmutation of waste system. To determine the best transmutation capabilities, various target design and loading optimization studies were performed. It was found that both 99Tc and 129I can be stabilized (i.e., zero net production) in the same PWR core under current design constraints by mixing 99Tc with fuel and by loading CaI2 target pins mixed with ZrH2 in guide tubes, but the PWR option appears to have a limited applicability as a burner of legacy LLFP. In fast systems, loading of moderated LLFP target assemblies in the core periphery (reflector region) was found to be preferable from the viewpoint of neutron economy and safety. By a simultaneous loading of 99Tc and 129I target assemblies in the reflector region, the self-generated 99Tc and 129I as well as the amount produced by several PWR cores could be consumed at a cost of ∼10% increased fuel inventory. Discharge burnups of ∼29 and ∼37% are achieved for 99Tc and 129I target assemblies with an ∼5-yr irradiation period.Based on these results, the impacts of 99Tc and 129I transmutation on the Yucca mountain repository were assessed in terms of the dose rate. The current Yucca Mountain release evaluations do not indicate a compelling need to transmute 99Tc and 129I because the resulting dose rates fall well below current regulatory limits. However, elimination of the LLFP inventory could allow significant relaxation of

  15. Analysis of fission product release behavior during the TMI-2 accident

    International Nuclear Information System (INIS)

    An analysis of fission product release during the Three Mile Island Unit 2 (TMI-2) accident has been initiated to provide an understanding of fission product behavior that is consistent with both the best estimate accident scenario and fission product results from the ongoing sample acquisition and examination efforts. First principles fission product release models are used to describe release from intact, disrupted, and molten fuel. Conclusions relating to fission product release, transport, and chemical form are drawn. 35 references

  16. Compilation of fission product yields Vallecitos Nuclear Center

    International Nuclear Information System (INIS)

    This document is the ninth in a series of compilations of fission yield data made at Vallecitos Nuclear Center in which fission yield measurements reported in the open literature and calculated charge distributions have been utilized to produce a recommended set of yields for the known fission products. The original data with reference sources, as well as the recommended yields are presented in tabular form for the fissionable nuclides U-235, Pu-239, Pu-241, and U-233 at thermal neutron energies; for U-235, U-238, Pu-239, and Th-232 at fission spectrum energies; and U-235 and U-238 at 14 MeV. In addition, U-233, U-236, Pu-240, Pu-241, Pu-242, Np-237 at fission spectrum energies; U-233, Pu-239, Th-232 at 14 MeV and Cf-252 spontaneous fission are similarly treated. For 1979 U234F, U237F, Pu249H, U234He, U236He, Pu238F, Am241F, Am243F, Np238F, and Cm242F yields were evaluated. In 1980, Th227T, Th229T, Pa231F, Am241T, Am241H, Am242Mt, Cm245T, Cf249T, Cf251T, and Es254T are also evaluated

  17. Relative fission product yield determination in the USGS TRIGA Mark I reactor

    Science.gov (United States)

    Koehl, Michael A.

    Fission product yield data sets are one of the most important and fundamental compilations of basic information in the nuclear industry. This data has a wide range of applications which include nuclear fuel burnup and nonproliferation safeguards. Relative fission yields constitute a major fraction of the reported yield data and reduce the number of required absolute measurements. Radiochemical separations of fission products reduce interferences, facilitate the measurement of low level radionuclides, and are instrumental in the analysis of low-yielding symmetrical fission products. It is especially useful in the measurement of the valley nuclides and those on the extreme wings of the mass yield curve, including lanthanides, where absolute yields have high errors. This overall project was conducted in three stages: characterization of the neutron flux in irradiation positions within the U.S. Geological Survey TRIGA Mark I Reactor (GSTR), determining the mass attenuation coefficients of precipitates used in radiochemical separations, and measuring the relative fission products in the GSTR. Using the Westcott convention, the Westcott flux, modified spectral index, neutron temperature, and gold-based cadmium ratios were determined for various sampling positions in the USGS TRIGA Mark I reactor. The differential neutron energy spectrum measurement was obtained using the computer iterative code SAND-II-SNL. The mass attenuation coefficients for molecular precipitates were determined through experiment and compared to results using the EGS5 Monte Carlo computer code. Difficulties associated with sufficient production of fission product isotopes in research reactors limits the ability to complete a direct, experimental assessment of mass attenuation coefficients for these isotopes. Experimental attenuation coefficients of radioisotopes produced through neutron activation agree well with the EGS5 calculated results. This suggests mass attenuation coefficients of molecular

  18. Evaluation and compilation of fission product yields 1993

    International Nuclear Information System (INIS)

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

  19. Evaluation and compilation of fission product yields 1993

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  20. Trapping technology for gaseous fission products from voloxidation process

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S

    2005-05-15

    The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, {sup 14}C, Kr, Xe, I and {sup 3}H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and {sup 14}C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for {sup 3}H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system.

  1. Trapping technology for gaseous fission products from voloxidation process

    International Nuclear Information System (INIS)

    The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, 14C, Kr, Xe, I and 3H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and 14C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for 3H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system

  2. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  3. Comparison of Fission Product Yields and Their Impact

    Energy Technology Data Exchange (ETDEWEB)

    S. Harrison

    2006-02-01

    This memorandum describes the Naval Reactors Prime Contractor Team (NRPCT) Space Nuclear Power Program (SNPP) interest in determining the expected fission product yields from a Prometheus-type reactor and assessing the impact of these species on materials found in the fuel element and balance of plant. Theoretical yield calculations using ORIGEN-S and RACER computer models are included in graphical and tabular form in Attachment, with focus on the desired fast neutron spectrum data. The known fission product interaction concerns are the corrosive attack of iron- and nickel-based alloys by volatile fission products, such as cesium, tellurium, and iodine, and the radiological transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.

  4. Waste treatment of fission product solutions containing aluminium nitrate

    International Nuclear Information System (INIS)

    In the Rossendorf molybdenum-99 production facility AMOR short-term irradiated aluminium clad fuel elements from the Rossendorf Research Reactor are reprocessed. Following extractive recovery of the enriched uranium the facility system has to be disposed of the fission product-Al(NO3)3 solution. Investigations on waste conditioning of such solutions are presented. (author)

  5. Status report on actinide and fission product transmutation studies

    International Nuclear Information System (INIS)

    The management of radioactive waste is one of the key issues in today's political and public discussions on nuclear energy. One of the fields that looks into the future possibilities of nuclear technology is the neutronic transmutation of actinides and of some most important fission products. Studies on transmutation of actinides are carried out in various countries and at an international level. This status report which gives an up-to-date general overview of current and planned research on transmutation of actinides and fission products in non-OECD countries, has been prepared by a Technical Committee meeting organized by the IAEA in September 1995. 168 refs, 16 figs, 34 tabs

  6. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

    A specialist's meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions)

  7. An alternate procedure in the recovery of no fissioned remainder uranium in the production of molybdenum 99 from fission

    International Nuclear Information System (INIS)

    An effective modification of the chemical processes to dissolve the U-IV in the dissolver has been obtained, using its highly alkaline pH and extracting it as Uranyl Triperoxidate soluble anionic complex, in its experimental design without fission products. Even when the extraction of uranium is usually more complete through acidic dissolution, the characteristics for the dissolver used in production of fission Mo-99 do not allow this kind of extraction and alkaline option is more adecuate for this purpose. The dissolution of the insoluble residue, through the production of the anionic Triperoxidate Uranyl complexes, arises rapidly due to the presence of and oxidizing agent. The best results in the extraction of soluble Uranium were obtained with and organic solvent and a mixture of carbonate/bicarbonate. The concentrated Uranium in the aqueous alkaline solution was separated through fixation as an anion Tricarbonate of Uranyl in columns of anionic resin, moderately basic in dynamic conditions. The superiority of the resin used, over other exchangers, was evident in the elution with nitric acid that may be done for small volumes with a quite favorable separation of Uranium. The eluate contains the Uranium as an hexahydrated Uranyl Nitrate with a high degree of purity in reduced volume, in an average concentration of 90.2 % with respect to the initial concentration of Uranium (Author)

  8. Fission product retention during faults involving steam generator tube rupture

    International Nuclear Information System (INIS)

    In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

  9. Data summary report for fission product release test VI-5

    Energy Technology Data Exchange (ETDEWEB)

    Osborne, M.F.; Lorenz, R.A.; Travis, J.R.; Webster, C.S.; Collins, J.L. (Oak Ridge National Lab., TN (United States))

    1991-10-01

    Test VI-5, the fifth in a series of high-temperature fission product release tests in a vertical test apparatus, was conducted in a flowing mixture of hydrogen and helium. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium which had been irradiated to a burnup of {approximately}42 MWd/kg. Using a hot cell-mounted test apparatus, the fuel rod was heated in an induction furnace under simulated LWR accident conditions to two test temperatures, 2000 K for 20 min and then 2700 K for an additional 20 min. The released fission products were collected in three sequentially operated collection trains on components designed to measure fission product transport characteristics and facilitate sampling and analysis. The results from this test were compared with those obtained in previous tests in this series and with the CORSOR-M and ORNL diffusion release models for fission product release. 21 refs., 19 figs., 12 tabs.

  10. Applications for fission product data to problems in stellar nucleosynthesis

    International Nuclear Information System (INIS)

    A general overview of the nucleosynthesis mechanisms for heavy (A greater than or equal to 70) nuclei is presented with particular emphasis on critical data needs. The current state of the art in nucleosynthesis models is described and areas in which fission product data may provide useful insight are proposed. 33 references, 10 figures

  11. Fission product filter for hot reactor cooling gas

    International Nuclear Information System (INIS)

    The fission product filter for He consists of a winding body composed of two corrugated metal sheets simultaneously wound on a core laterally reversed. It is inserted into an enclosing tube and held at top and bottom by a star-shaped yoke. (DG)

  12. A model for fission-product calculations, 1

    International Nuclear Information System (INIS)

    Many fission-product cross sections remain unmeasurable thus considerable reliance must be placed upon calculational interpolation and exstrapolation from the few available measured cross sections. The vehicle, particularly for the lighter fission products, is the conventional the optical-statistical model. The applied goals generally are: capture cross sections to 7 - 10 % accuracies and inelastic-scattering cross sections to 25 - 50 %. Comparisons of recent evaluations and experimental results indicate that these goals have too often are far from met, particularly in the area of inelastic scattering, and some of the evaluated fission-product cross sections are simply physically unreasonable. An example of these discrepancies is shown in a figure. The evaluated inelastic-scattering cross sections of palladium are nearly a 100 % discrepant with observation and the isotopes are prominent fission products with large inelastic-scattering cross sections at relatively low energies. It is difficult to avoid the conclusion that the models employed in many of the evaluations are inappropriate and/or inappropriately used. (author)

  13. Results of recent ORNL fission product release tests

    International Nuclear Information System (INIS)

    Four fission product release tests have been performed with Zircaloy-clad uranium dioxide (UO2) fuel rod segments in the Oak Ridge National Laboratory (ORNL) vertical induction-heated (VI) apparatus at temperatures up to 2700 K. The first three tests (VI-1, VI-2, and VI-3) were performed in a steam-helium atmosphere, and test VI-4 was performed in a hydrogen-helium atmosphere. In test VI-4, the strongly reducing atmosphere created by melted Zircaloy in hydrogen caused significant release of the fission product europium and good retention of the fission product antimony. The releases of krypton and cesium were similar in both atmospheres even though the fuel rod collapsed shortly after the melting point of the cladding was reached. The formation of volatile iodine species (I2, HI, and CH3I) remained low (<0.5%) in hydrogen atmosphere test VI-4. Good release correlations for volatile fission products have been obtained using the ORNL Diffusion Release Model. Cesium transport behavior was affected by the hydrogen atmosphere

  14. Progress in fission product nuclear data. Issue no. 6

    International Nuclear Information System (INIS)

    This is the sixth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed

  15. Early results utilizing high-energy fission product gamma rays to detect fissionable material in cargo

    International Nuclear Information System (INIS)

    Full text: A concept for detecting the presence of special nuclear material (235U or 239Pu) concealed in inter modal cargo containers is described. It is based on interrogation with a pulsed beam of 6-8 MeV neutrons and fission events are identified between beam pulses by their β-delayed neutron emission or β -delayed high-energy γ-radiation. The high-energy γ-ray signature is being employed for the first time. Fission product γ-rays above 3 MeV are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. High-energy γ-radiation is nearly 10X more abundant than the delayed neutrons and penetrates even thick cargo's readily. The concept employs two large (8x20 ft) arrays of liquid scintillation detectors that have high efficiency for the detection of both delayed neutrons and delayed γ-radiation. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified. This information, together with predicted signature strength, has been applied to the estimation of detection probability for the nuclear material and estimation of false alarm rates. This work was performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48

  16. JNDC nuclear data library of fission products, second version

    International Nuclear Information System (INIS)

    The second version of the JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library is described in this report. The library contains nuclear decay and fission yield data for 1078 unstable and 149 stable FP nuclides, and neutron cross section data for 166 nuclides. The decay data include half-life, branching ration, and total beta- and gamma-ray energies released per decay of each unstable nuclide. The theoretical and the experimental values of average beta and gamma decay energies have been thoroughly reexamined for each nuclide, and the best values or most reliable ones have been chosen for inclusion into the new version. The comparison of decay power curves between the calculations with the new version and the measurements performed at the University of Tokyo, Oak Ridge National Laboratory and Los Alamos National Laboratory for variety of fissiles from 232Th to 241Pu shows clear improvement in agreement, in particular, around 1000 s and also after 1000 s. The decay power of fission products has been calculated for twenty fission types and the results have been fitted by an analytical function with 33 exponentials. This permits the easy application of the present results of decay power calculations to a LOCA (Loss-of-Coolant Accident) analysis of a light water reactor and so on. (author)

  17. Report on simulation of fission gas and fission product diffusion in UO2

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Perriot, Romain Thibault [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation Dept.; Tonks, Michael R. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation Dept.; Cooper, Michael William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Goyal, Anuj [Univ. of Florida, Gainesville, FL (United States). Dept. of Materials Science and Engineering; Uberuaga, Blas P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division

    2016-07-22

    In UO2 nuclear fuel, the retention and release of fission gas atoms such as xenon (Xe) are important for nuclear fuel performance by, for example, reducing the fuel thermal conductivity, causing fuel swelling that leads to mechanical interaction with the clad, increasing the plenum pressure and reducing the fuel–clad gap thermal conductivity. We use multi-­scale simulations to determine fission gas diffusion mechanisms as well as the corresponding rates in UO2 under both intrinsic and irradiation conditions. In addition to Xe and Kr, the fission products Zr, Ru, Ce, Y, La, Sr and Ba have been investigated. Density functional theory (DFT) calculations are used to study formation, binding and migration energies of small clusters of Xe atoms and vacancies. Empirical potential calculations enable us to determine the corresponding entropies and attempt frequencies for migration as well as investigate the properties of large clusters or small fission gas bubbles. A continuum reaction-­diffusion model is developed for Xe and point defects based on the mechanisms and rates obtained from atomistic simulations. Effective fission gas diffusivities are then obtained by solving this set of equations for different chemical and irradiation conditions using the MARMOT phase field code. The predictions are compared to available experimental data. The importance of the large XeU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and high binding energy. We find that the XeU3O cluster gives Xe diffusion coefficients that are higher for intrinsic conditions than under irradiation over a wide range of temperatures. Under irradiation the fast-­moving XeU3O cluster recombines quickly with irradiation induced interstitial U ions, while this mechanism is less important for intrinsic conditions. The net result is higher

  18. Report on simulation of fission gas and fission product diffusion in UO2

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Perriot, Romain Thibault [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation Dept.; Tonks, Michael R. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation Dept.; Cooper, Michael William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Goyal, Anuj [Univ. of Florida, Gainesville, FL (United States). Dept. of Materials Science and Engineering; Uberuaga, Blas P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division; Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science and Technology Division

    2016-07-22

    In UO2 nuclear fuel, the retention and release of fission gas atoms such as xenon (Xe) are important for nuclear fuel performance by, for example, reducing the fuel thermal conductivity, causing fuel swelling that leads to mechanical interaction with the clad, increasing the plenum pressure and reducing the fuel–clad gap thermal conductivity. We use multi-­scale simulations to determine fission gas diffusion mechanisms as well as the corresponding rates in UO2 under both intrinsic and irradiation conditions. In addition to Xe and Kr, the fission products Zr, Ru, Ce, Y, La, Sr and Ba have been investigated. Density functional theory (DFT) calculations are used to study formation, binding and migration energies of small clusters of Xe atoms and vacancies. Empirical potential calculations enable us to determine the corresponding entropies and attempt frequencies for migration as well as investigate the properties of large clusters or small fission gas bubbles. A continuum reaction-­diffusion model is developed for Xe and point defects based on the mechanisms and rates obtained from atomistic simulations. Effective fission gas diffusivities are then obtained by solving this set of equations for different chemical and irradiation conditions using the MARMOT phase field code. The predictions are compared to available experimental data. The importance of the large XeU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and high binding energy. We find that the XeU3O cluster gives Xe diffusion coefficients that are higher for intrinsic conditions than under irradiation over a wide range of temperatures. Under irradiation the fast-­moving XeU3O cluster recombines quickly with irradiation-induced interstitial U ions, while this mechanism is less important for intrinsic conditions. The net result is higher

  19. Fission Product Yield Study of 235U, 238U and 239Pu Using Dual-Fission Ionization Chambers

    Science.gov (United States)

    Bhatia, C.; Fallin, B.; Howell, C.; Tornow, W.; Gooden, M.; Kelley, J.; Arnold, C.; Bond, E.; Bredeweg, T.; Fowler, M.; Moody, W.; Rundberg, R.; Rusev, G.; Vieira, D.; Wilhelmy, J.; Becker, J.; Macri, R.; Ryan, C.; Sheets, S.; Stoyer, M.; Tonchev, A.

    2014-05-01

    To resolve long-standing differences between LANL and LLNL regarding the correct fission basis for analysis of nuclear test data [M.B. Chadwick et al., Nucl. Data Sheets 111, 2891 (2010); H. Selby et al., Nucl. Data Sheets 111, 2891 (2010)], a collaboration between TUNL/LANL/LLNL has been established to perform high-precision measurements of neutron induced fission product yields. The main goal is to make a definitive statement about the energy dependence of the fission yields to an accuracy better than 2-3% between 1 and 15 MeV, where experimental data are very scarce. At TUNL, we have completed the design, fabrication and testing of three dual-fission chambers dedicated to 235U, 238U, and 239Pu. The dual-fission chambers were used to make measurements of the fission product activity relative to the total fission rate, as well as for high-precision absolute fission yield measurements. The activation method was employed, utilizing the mono-energetic neutron beams available at TUNL. Neutrons of 4.6, 9.0, and 14.5 MeV were produced via the 2H(d,n)3He reaction, and for neutrons at 14.8 MeV, the 3H(d,n)4He reaction was used. After activation, the induced γ-ray activity of the fission products was measured for two months using high-resolution HPGe detectors in a low-background environment. Results for the yield of seven fission fragments of 235U, 238U, and 239Pu and a comparison to available data at other energies are reported. For the first time results are available for neutron energies between 2 and 14 MeV.

  20. Evaluation of independent and cumulative fission product yields with gamma spectrometry

    International Nuclear Information System (INIS)

    Fission product yields are critical data for a variety of nuclear science and engineering applications; however, independent yields have not been extensively measured to date. We have previously documented a methodology to measure the cumulative and independent fission product yields using gamma spectrometry and nuclide buildup and decay modeling, and numerical optimization. We have produced fission products by bombarding 235U with 14.1 MeV neutrons and made measurements of fission product yields. In this paper, we summarize our approach, describe initial experiments, and present preliminary results where we have determined nine fission product yields for long-lived nuclides. (author)

  1. Superabsorbing gel for actinide, lanthanide, and fission product decontamination

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, Michael D.; Mertz, Carol J.

    2016-06-07

    The present invention provides an aqueous gel composition for removing actinide ions, lanthanide ions, fission product ions, or a combination thereof from a porous surface contaminated therewith. The composition comprises a polymer mixture comprising a gel forming cross-linked polymer and a linear polymer. The linear polymer is present at a concentration that is less than the concentration of the cross-linked polymer. The polymer mixture is at least about 95% hydrated with an aqueous solution comprising about 0.1 to about 3 percent by weight (wt %) of a multi-dentate organic acid chelating agent, and about 0.02 to about 0.6 molar (M) carbonate salt, to form a gel. When applied to a porous surface contaminated with actinide ions, lanthanide ions, and/or other fission product ions, the aqueous gel absorbs contaminating ions from the surface.

  2. Quantitative analysis of fission products by γ spectrography

    International Nuclear Information System (INIS)

    The activity of the fission products present in treated solutions of irradiated fuels is given as a function of the time of cooling and of the irradiation time. The variation of the ratio (144Ce + 144Pr activity)/ 137Cs activity) as a function of these same parameters is also given. From these results a method is deduced giving the 'age' of the solution analyzed. By γ-scintillation spectrography it was possible to estimate the following elements individually: 141Ce, 144Ce + 144Pr, 103Ru, 106Ru + 106Rh, 137Cs, 95Zr + 95Nb. Yield curves are given for the case of a single emitter. Of the various existing methods, that of the least squares was used for the quantitative analysis of the afore-mentioned fission products. The accuracy attained varies from 3 to 10%. (author)

  3. NEANDC specialists meeting on yields and decay data of fission product nuclides

    International Nuclear Information System (INIS)

    Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information

  4. NEANDC specialists meeting on yields and decay data of fission product nuclides

    Energy Technology Data Exchange (ETDEWEB)

    Chrien, R.E.; Burrows, T.W. (eds.)

    1983-01-01

    Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information. (WHK)

  5. Decontamination of radioactive waste fission products by treated natural clays

    International Nuclear Information System (INIS)

    The removal of carrier free long living fission products such as iodine-131, strontium-90 and cesium-137 by treated local clays is successfully achieved with large capacity. Iodine-131 which is difficultly adsorbed has been removed completely by silver treated phosphate clay. Strontium-90 and cesium-137 have been almost removed by adequate heat treating of the clays. The results of column experiments agree well with the authors' batch experiments. (author)

  6. Forced decontamination of fission products deposited on urban areas

    International Nuclear Information System (INIS)

    Long-lived fission products may be deposited in the environment following a serious reactor accident. Areas of special concern are cities where the collective dose might be high because of the population. An extensive literature list is presented here. Only a few of the references deal with the problem as a whole. Some references deal with non-radiaoctive materials but give us useful information about the behaviour of particles on outdoor surfaces. (author)

  7. Irradiation effects upon activities of fission product iodine

    International Nuclear Information System (INIS)

    This report describes the experimental study of the irradiation effects upon activities of fission product iodine made in the period from June, 1981 to March, 1982. Chemical transport of iron was studied under irradiation of cesium iodide by electron beam. Deposited ion was identified on the high temperature surface, which can be taken to certify the appropriateness of the model of the iodine-including chemical transport of stainless-steel cladding components to fuel in the LMFBR fuel pins. (author)

  8. DAMD code for producing nuclear data library of fission products

    International Nuclear Information System (INIS)

    Computer codes DAMD, TACA and TREE have been developed. The code DAMD produces a nuclear data library from ENDF/B format library for the computer code DCHAIN which analyzes buildup and decay of fission products. The code TACA punches out and prints out the contents of the nuclear data library for DCHAIN. The code TREE prints out the decay schemes of the nuclides contained in the library. (auth.)

  9. Neutron cross section calculations for fission-product nuclei

    International Nuclear Information System (INIS)

    To satisfy nuclear data requirements for fission-product nuclei, Hauser-Feshbach statistical calculations with preequilibrium corrections for neutron-induced reactions on isotopes of Se, Kr, Sr, Zr, Mo, Sn, Xe, and Ba between 0.001 and 20 MeV. Spherical neutron optical parameters were determined by simultaneous fits to resonance data and total cross sections. Isospin coefficients appearing in the optical potentials were determined through analysis of the behavior of s- and p-wave strengths as a function of mass for a given Z. Gamma-ray strength functions, determined through fits to stable-isotope capture data, were used in the calculation of capture cross sections and gamma-ray competition to particle emission. The resulting (n,γ), (n,n'), (n,2n), and (n,3n) cross sections, the secondary neutron emission spectra, and angular distributions calculated for 19 fission products will be averaged to provide a resulting ENDF-type fission-product neutronics file. 11 references

  10. ACRR fission product release tests: ST-1 and ST-2

    International Nuclear Information System (INIS)

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model

  11. Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

    Science.gov (United States)

    Herrmann, Steven Douglas

    2014-05-27

    Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

  12. RAPID QUANTITATION OF URANIUM FROM MIXED FISSION PRODUCT SAMPLES

    Energy Technology Data Exchange (ETDEWEB)

    Haney, Morgan M.; Seiner, Brienne N.; Finn, Erin C.; Friese, Judah I.

    2016-03-09

    Chemical similarities between U(VI) and Mo(VI) create challenges for separation and quantification of uranium from a mixed fission product sample. The purpose of this work was to demonstrate the feasibility of using Eichrom’s® UTEVA resin in addition to a tellurium spontaneous deposition to improve the quantitation of 235U using gamma spectroscopy. The optimized method demonstrated a consistent chemical yield of 74 ± 3 % for uranium. This procedure was evaluated using 1.41x1012 fissions produced from an irradiated HEU sample. The uranium was isotopically yielded by HPGe, and the minimum detectable activity (MDA) determined from the gamma spectra. The MDA for 235U, 237U, and 238U was reduced by a factor of two. The chemical isolation of uranium was successfully achieved in less than four hours, with a separation factor of 1.41x105 from molybdenum.

  13. Effect of culinary and technological treatment of the farm animal products on the fission products content in them

    International Nuclear Information System (INIS)

    The application of various procedures aimed at reducing the content of artificial radionuclides at the stage of technological and culinary processing of agricultural produce is considered. During the processing of milk and meat, which are basic farming produce, much of the contained fission products can be removed with low-value wastes. One of the more readily accessible methods of milk purification is recognized to be that of ion ixchange. A large role in reducing the radionuclide content of farming products is played by the time factor, i.e. the time spent for the manufacture and marketing of the products

  14. Fission and corrosion products behavior in primary circuits of LMFBR's

    International Nuclear Information System (INIS)

    Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

  15. Thermoradiation treatment of sewage sludge using reactor waste fission products

    Energy Technology Data Exchange (ETDEWEB)

    Reynolds, M. C.; Hagengruber, R. L.; Zuppero, A. C.

    1974-06-01

    The hazards to public health associated with the application of municipal sewage sludge to land usage are reviewed to establish the need for disinfection of sludge prior to its distribution as a fertilizer, especially in the production of food and fodder. The use of ionizing radiation in conjunction with mild heating is shown to be an effective disinfection treatment and an economical one when reactor waste fission products are utilized. A program for researching and experimental demonstration of the process on sludges is also outlined.

  16. Thermoradiation treatment of sewage sludge using reactor waste fission products

    International Nuclear Information System (INIS)

    The hazards to public health associated with the application of municipal sewage sludge to land usage are reviewed to establish the need for disinfection of sludge prior to its distribution as a fertilizer, especially in the production of food and fodder. The use of ionizing radiation in conjunction with mild heating is shown to be an effective disinfection treatment and an economical one when reactor waste fission products are utilized. A program for researching and experimental demonstration of the process on sludges is also outlined

  17. Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2005-08-12

    U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.

  18. Fission Product Separation from Pyrochemical Electrolyte by Cold Finger Melt Crystallization

    Energy Technology Data Exchange (ETDEWEB)

    Versey, Joshua R. [Univ. of Idaho, Moscow, ID (United States)

    2013-08-01

    This work contributes to the development of pyroprocessing technology as an economically viable means of separating used nuclear fuel from fission products and cladding materials. Electrolytic oxide reduction is used as a head-end step before electrorefining to reduce oxide fuel to metallic form. The electrolytic medium used in this technique is molten LiCl-Li2O. Groups I and II fission products, such as cesium (Cs) and strontium (Sr), have been shown to partition from the fuel into the molten LiCl-Li2O. Various approaches of separating these fission products from the salt have been investigated by different research groups. One promising approach is based on a layer crystallization method studied at the Korea Atomic Energy Research Institute (KAERI). Despite successful demonstration of this basic approach, there are questions that remain, especially concerning the development of economical and scalable operating parameters based on a comprehensive understanding of heat and mass transfer. This research explores these parameters through a series of experiments in which LiCl is purified, by concentrating CsCl in a liquid phase as purified LiCl is crystallized and removed via an argon-cooled cold finger.

  19. Status of the French research programme for actinides and fission products partitioning and transmutation

    International Nuclear Information System (INIS)

    The paper focus on separation and transmutation research and development programme and main results over these ten last years. The massive research programme on enhanced separation, conducted by CEA and supported by broad international cooperation, has recently achieved some vital progress. Based on real solutions derived from the La Hague process, the CEA demonstrated the lab-scale feasibility of extracting minor actinides and some fission products (I, Cs and Tc) using an hydrometallurgical process that can be extrapolated on the industrial scale. The CEA also conducted programmes proving the technical feasibility of the elimination of minor actinides and fission products by transmutation: fabrication of specific targets and fuels for transmutation tests in the HFR and Phenix reactors, neutronics and technology studies for ADS developments in order to support the MEGAPIE, TRADE and MYRRHA experiments and the future 100 MW international ADS demonstrator. Scenarios studies aimed at stabilizing the inventory with long-lived radionuclides, plutonium, minor actinides and certain long-lived fission products in different nuclear power plant parks and to verify the feasibility at the level of the cycle facilities and fuels involved in those scenarios. Three French Research Groups CEA-CNRS carry out partitioning (PRACTIS) and transmutation (NOMADE and GEDEON) more basic studies. (author)

  20. Measurement and characterization of fission products released from LWR fuel

    International Nuclear Information System (INIS)

    Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from proportional 2% at 14000C to >50% at 20000C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag. (orig./HP)

  1. Actinides and fission products partitioning from high level liquid waste

    International Nuclear Information System (INIS)

    The presence of small amount of mixed actinides and long-lived heat generators fission products as 137Cs and 90Sr are the major problems for safety handling and disposal of high level nuclear wastes. In this work, actinides and fission products partitioning process, as an alternative process for waste treatment is proposed. First of all, ammonium phosphotungstate (PWA), a selective inorganic exchanger for cesium separation was chosen and a new procedure for synthesizing PWA into the organic resin was developed. An strong anionic resin loaded with tungstate or phosphotungstate anion enables the precipitation of PWA directly in the resinous structure by adding the ammonium nitrate in acid medium (R-PWA). Parameters as W/P ratio, pH, reactants, temperature and aging were studied. The R-PWA obtained by using phosphotungstate solution prepared with W/P=9.6, 9 hours digestion time at 94-106 deg C and 4 to 5 months aging time showed the best capacity for cesium retention. On the other hand, Sr separation was performed by technique of extraction chromatography, using DH18C6 impregnated on XAD7 resin as stationary phase. Sr is selectively extracted from acid solution and >99% was recovered from loaded column using distilled water as eluent. Concerning to actinides separations, two extraction chromatographic columns were used. In the first one, TBP(XAD7) column, U and Pu were extracted and its separations were carried-out using HNO3 and hydroxylamine nitrate + HNO3 as eluent. In the second one, CMP0-TBP(XAD7) column, the actinides were retained on the column and the separations were done by using (NH4)2C2O4 , DTPA, HNO3 and HCl as eluent. The behavior of some fission products were also verified in both columns. Based on the obtained data, actinides and fission products Cs and Sr partitioning process, using TBP(XAD7) and CMP0-TBP(XAD7) columns for actinides separation, R-PWA column for cesium retention and DH18C6(XAD7) column for Sr isolation was performed. (author)

  2. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    Science.gov (United States)

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-01

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  3. Overview of experimental programs on core melt progression and fission product release behaviour

    International Nuclear Information System (INIS)

    An overview of experimental programs that have been conducted to better understand core melt progression phenomena and fission product behaviour during severe reactor accidents in water reactors is presented. This discussion principally focuses on the melting and liquefaction of core materials at different temperatures, materials oxidation and relocation, hydrogen generation behaviour, and the release and transport of fission products and aerosols. A comparison of fission product release results from annealing and in-reactor experiments is also presented. (author)

  4. Decay characteristics of fission products and summation calculation

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Tadashi [Faculty of Engineering, Musashi Institute of Technology, Tokyo (Japan)

    1999-02-01

    This paper reviews the decay characteristics of fission products on the viewpoint of summation calculation. The fission products (FPs) are accumulated in the operating power reactors. As they are neutron-rich at the time of scission, they undergo successive beta decays toward stable nuclides. To grasp the quantity of an arbitrary nuclide, fission yields, decay constants and blanching ratios of the nuclide in the same decay chain ( a mass chain of the fixed mass is sufficient) must be known. As a neutron capture increases the mass, and release of a delayed neutron decreases the mass, capture cross sections and delayed neutron emitting ratios are also required. If these values of all FP are known, the quantities such as time dependent decay heat and the delayed neutron fraction can be calculated by summation of the contribution of the nuclides. A computer code ORIGEN-2 is a typical example to compute these quantities. The more important than computer code is the data library for summation calculation which includes physical constants such as fission yields, decay constants, blanching ratio, beta and gamma energy emitted at a beta decay, delayed neutron emitting ratios, and neutron capture cross sections for more than 1000 FP nuclides. They are realized in JNDC FP Decay Data Library-Version 2 of Japan, JEF-2 by western European countries, and ENDF/B-VI of USA. The early versions (until early 80's) of these full-scale libraries showed worse agreement with experiment than the old libraries based on approximations and estimates. The application of the gross theory to beta-decay' to short-lived FPs could solve the problem. The above disagreement is explained by having dropped of high excitation levels of short lived daughter nuclides. This is called as Pandemonium Problem. The summation calculation for the gamma ray spectrum succeeded to predict the experimental value by correcting theoretical spectrum. However, there remains still an underestimate for cooling

  5. Airborne measurements of fission product fall-out

    Energy Technology Data Exchange (ETDEWEB)

    Hovgaard, J.; Korsbech, U.

    1992-12-01

    During 1993 the Danish Emergency Management Agency will install an airborne [gamma]-ray detector system for area survey of contamination with radioactive nuclides - primarily fission products that may be released during a heavy accident at a nuclear power plant or from accidents during transport of radioactive material. The equipment is based on 16 liter NaI(TI) crystals and multichannel analysers from Exploranium (Canada). A preliminary investigation of the possibilities for detection of low and high level contamination - and the problems that may be expected during use of the equipment, and during interpretation of the measured data, is described. Several days after reactor shut-down some of the nuclides can be identified directly from the measured spectrum, and contamination levels may be determined within a factor two. After several weeks, most fission products have decayed. Concentrations and exposure rates can be determined with increasing accuracy as time passes. Approximate calibration of the equipment for measurements of surface contamination and natural radioactivity can be performed in the laboratory. Further checks of equipment should include accurate measurements of the spectrum resolution. Detectors should be checked individually, and all together. Further control of dead time and pulse pile-up should be performed. Energy calibration, electronics performance and data equipment should be tested against results from the original calibration. (AB).

  6. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D.; Wagh, A. [Argonne National Lab., IL (United States)

    1997-10-01

    Chemically bonded phosphate ceramics (CBPCs) have several advantages that make them ideal candidates for containing radioactive and hazardous wastes. In general, phosphates have high solid-solution capacities for incorporating radionuclides, as evidenced by several phosphates (e.g., monazites and apatites) that are natural analogs of radioactive and rare-earth elements. The phosphates have high radiation stability, are refractory, and will not degrade in the presence of internal heating by fission products. Dense and hard CBPCs can be fabricated inexpensively and at low temperature by acid-base reactions between an inorganic oxide/hydroxide powder and either phosphoric acid or an acid-phosphate solution. The resulting phosphates are extremely insoluble in aqueous media and have excellent long-term durability. CBPCs offer the dual stabilization mechanisms of chemical fixation and physical encapsulation, resulting in superior waste forms. The goal of this task is develop and demonstrate the feasibility of CBPCs for S/S of wastes containing fission products. The focus of this work is to develop a low-temperature CBPC immobilization system for eluted {sup 99}Tc wastes from sorption processes.

  7. Integrated separation scheme for measuring a suite of fission and activation products from a fresh mixed fission and activation product sample

    International Nuclear Information System (INIS)

    Mixed fission and activation materials resulting from various nuclear processes and events contain a wide range of isotopes for analysis spanning almost the entire periodic table. This work describes the production of a complex synthetic sample containing fission products, activation products, and irradiated soil, and determines the percent chemical recovery of select isotopes through the integrated chemical separation scheme. Based on the results of this experiment, a complex synthetic sample can be prepared with low atom/fission ratios and isotopes of interest accurately and precisely measured following an integrated chemical separation method. (author)

  8. Preliminary investigation of a technique to separate fission noble metals from fission-product mixtures

    International Nuclear Information System (INIS)

    A variation of the gold-ore fire assay technique was examined as a method for recovering Pd, Rh and Ru from fission products. The mixture of fission product oxides is combined with glass-forming chemicals, a metal oxide such as PbO (scavenging agent), and a reducing agent such as charcoal. When this mixture is melted, a metal button is formed which extracts the noble metals. The remainder cools to form a glass for nuclear waste storage. Recovery depended only on reduction of the scavenger oxide to metal. When such reduction was achieved, no difference in noble metal recovery efficiency was found among the scavengers studied (PbO, SnO, CuO, Bi2O3, Sb2O3). Not all reducing agents studied, however, were able to reduce all scavenger oxides to metal. Only graphite would reduce SnO and CuO and allow noble metal recovery. The scavenger oxides Sb2O3, Bi2O3, and PbO, however, were reduced by all of the reducing agents tested. Similar noble metal recovery was found with each. Lead oxide was found to be the most promising of the potential scavengers. It was reduced by all of the reducing agents tested, and its higher density may facilitate the separation. Use of lead oxide also appeared to have no deterimental effect on the glass quality. Charcoal was identified as the preferred reducing agent. As long as a separable metal phase was formed in the melt, noble metal recovery was not dependent on the amount of reducing agent and scavenger oxide. High glass viscosities inhibited separation of the molten scavenger, while low viscosities allowed volatile loss of RuO4. A viscosity of approx. 20 poise at the processing temperature offered a good compromise between scavenger separation and Ru recovery. Glasses in which PbO was used as the scavenging agent were homogeneous in appearance. Resistance to leaching was close to that of certain waste glasses reported in the literature. 12 figures. 7 tables

  9. Fission product release in high-burn-up UO2 oxidized to U3O8

    International Nuclear Information System (INIS)

    Results of oxidation experiments on high-burn-up UO2 are presented where fission-product vaporisation and release rates have been measured by on-line mass spectrometry as a function of time/temperature during thermal annealing treatments in a Knudsen cell under controlled oxygen atmosphere. Fractional release curves of fission gas and other less volatile fission products in the temperature range 800-2000 K were obtained from BWR fuel samples of 65 G Wd t-1 burn-up and oxidized to U3O8 at low temperature. The diffusion enthalpy of gaseous fission products and helium in different structures of U3O8 was determined

  10. Analysis of Fission Products on the AGR-1 Capsule Components

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Jason M. Harp; Philip L. Winston; Scott A. Ploger

    2013-03-01

    The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

  11. Phosphonates as alternative to tributyl phosphate for the separation of actinides from fission products

    Energy Technology Data Exchange (ETDEWEB)

    Vyas, Chirag K.; Joshirao, Pranav M.; Manchanda, Vijay K. [Sungkyunkwan Univ., Suwon (Korea, Republic of). Dept. of Energy Science; Rao, C.V.S. Brahmmananda; Jayalakshmi, S. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Fuel Chemistry Div.

    2015-06-01

    The present work investigates the role of increase in the basicity of organophosphorus extractant (dialkylalkyl phosphonates) on the uptake of actinides and fission products vis-a-vis tributyl phosphate (TBP), currently employed as a universal extractant. Two dialkylalkyl phosphonates viz. dibutylpropyl phosphonate (DBPrP) and dibutylpentyl phosphonate (DBPeP) were synthesized, characterized and evaluated for their solvent extraction behavior towards U(VI), Th(IV), Eu(III) and Tc(VII) in nitric acid medium ranging from 0.01-6 M. It was observed that increasing the basicity of the phosphoryl oxygen enhanced the uptake of the actinides and the distribution coefficient values were significantly larger as compared to TBP. The limiting organic concentration (LOC) value was estimated for Th(IV) for these extractants and compared with the TBP system. The separation factors of actinides with phosphonates over Tc(VII) are distinctly better than that with TBP.

  12. Model to simulate the fission-product transport process in the Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    So, B.Y.C.

    1979-01-01

    When fission products are released from a cladding breach in EBR-II, they mix turbulently with the sodium in the core, in the upper plenum and in the intermediate heat exchanger. Eventually the fission products are discharged 12 to 13 s later into the primary tank. Fission gases migrate upward through a 9-ft layer of sodium and enter the cover gas. Loss of fission gas is due to decay, leakage of cover gas, cold trapping of iodine and bromine parents. Depending on the reactor operation requirement, it may purge with fresh argon. The assumptions made and differential equations used to develop a model for such transport are presented.

  13. UKFY2: The UK fission product yield library version 2, 1991

    International Nuclear Information System (INIS)

    The UKFY2 Fission Product Yields Library contains 7 files with fission yield information in different formats and references, as received at the IAEA Nuclear Data Section in February 1991. File 2 contains the complete set of adjusted independent and cumulative yields in ENDF-6 format as adopted for the JEF-2 fission product yield file. It contains yields for 21 different fissioning nuclides. Many more chain yield and fractional yield sets are given in tabular form in other files of this library. The data are available costfree on magnetic tape from the IAEA Nuclear Data Section. (author)

  14. A model for fission product distribution in CANDU fuel

    International Nuclear Information System (INIS)

    This paper describes a model to estimate the distribution of active fission products among the UO2 grains, grain-boundaries, and the free void spaces in CANDU fuel elements during normal operation. This distribution is required for the calculation of the potential release of activity from failed fuel sheaths during a loss-of-coolant accident. The activity residing in the free spaces (''free'' inventory) is available for release upon sheath rupture, whereas relatively high fuel temperatures and/or thermal shock are required to release the activity in the grain boundaries or grains. A preliminary comparison of the model with the data from in-reactor sweep-gas experiments performed in Canada yields generally good agreement, with overprediction rather than under prediction of radiologically important isotopes, such as I131. The model also appears to generally agree with the ''free'' inventory release calculated using ANS-5.4. (author)

  15. Approximation of the decay of fission and activation product mixtures

    International Nuclear Information System (INIS)

    The decay of the exposure rate from a mixture of fission and activation products is a complex function of time. The exact solution of the problem involves the solution of more than 150 tenth order Bateman equations. An approximation of this function is required for the practical solution of problems involving multiple integrations of this function. Historically this has been a power function, or a series of power functions, of time. The approach selected here has been to approximate the decay with a sum of exponential functions. This produces a continuous, single valued function, that can be made to approximate the given decay scheme to any desired degree of closeness. Further, the integral of the sum is easily calculated over any period. 3 refs

  16. Review of fission product plateout investigations at General Atomic

    International Nuclear Information System (INIS)

    The status of fission product plateout studies at General Atomic is reviewed and suggestions are offered for future work. The deposition, or plateout, of condensible radionuclides in the primary circuits of gas-cooled reactors affects shielding requirements, maintenance procedures, and plant availability as well as representing a significant radiological source and/or sink for certain hypothetical accidents. Physical models and computer codes used to describe these plateout phenomena for reactor analysis are presented along with their limitations and possible refinements. The review includes portions of the recent AIPA study which sought to quantify the effects of uncertainties in input parameters on plateout code predictions. Major emphasis is placed upon the design methods verification program to assess the validity of plateout predictions by comparison of calculated behavior with experimental transport data

  17. ORNL studies of fission product release under LWR accident conditions

    International Nuclear Information System (INIS)

    High burnup Zircaloy-clad UO2 fuel specimens have been heated to study the release of fission products in tests simulating LWR accident conditions. The dominant variable was found to be temperature, with atmosphere, time, and burnup also being significant variables. Comparison of data from tests in steam and hydrogen, at temperatures of 2000 to 2700 K, have shown that the releases of the most volatile species (Kr, Xe, I, and Cs) are relatively insensitive to atmosphere. The releases of the less-volatile species (Sr, Mo, Ru, Sb, Te, Ba, and Eu), however, may vary by orders of magnitude depending on atmosphere. In addition, the atmosphere may drastically affect the mode and extent of fuel destruction

  18. Small Scale Indigenous Molybdenum-99 Production Using LEU Fission at Chilean Nuclear Energy Commission [Country report: Chile

    International Nuclear Information System (INIS)

    This report contains the results of the activities carried out in the Chilean Nuclear Energy Commission (CCHEN) under CRP Nº 13358 “Small Scale Indigenous Molybdenum-99 Production Using LEU Fission” started in October 2005 to November 2011. The object of the project was to develop the basic infrastructure and to establish the conditions to obtain fission molybdenum-99 (99Mo) by neutron irradiation of uranium-235 (235U) targets in RECH-1 reactor located in Santiago, Chile

  19. Assessment of fission product yields data needs in nuclear reactor applications

    International Nuclear Information System (INIS)

    Studies on the build-up of fission products in fast reactors have been performed, with particular emphasis on the effects related to the physics of the nuclear fission process. Fission product yields, which are required for burn-up calculations, depend on the proton and neutron number of the target nucleus as well as on the incident neutron energy. Evaluated nuclear data on fission product yields are available for all relevant target nuclides in reactor applications. However, the description of their energy dependence in evaluated data is still rather rudimentary, which is due to the lack of experimental fast fission data and reliable physical models. Additionally, physics studies of evaluated JEFF-3.1.1 fission yields data have shown potential improvements, especially for various fast fission data sets of this evaluation. In recent years, important progress in the understanding of the fission process has been made, and advanced model codes are currently being developed. This paper deals with the semi-empirical approach to the description of the fission process, which is used in the GEF code being developed by K.-H. Schmidt and B. Jurado on behalf of the OECD Nuclear Energy Agency, and with results from the corresponding author's diploma thesis. An extended version of the GEF code, supporting the calculation of spectrum weighted fission product yields, has been developed. It has been applied to the calculation of fission product yields in the fission rate spectra of a MOX fuelled sodium-cooled fast reactor. Important results are compared to JEFF-3.1.1 data and discussed in this paper. (authors)

  20. Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S. [Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2003-07-01

    During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- {epsilon} turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential.

  1. First-principles study of the stability of fission products in uranium monocarbide

    Science.gov (United States)

    Bévillon, Émile; Ducher, Roland; Barrachin, Marc; Dubourg, Roland

    2012-07-01

    The incorporation and stability of fission products in uranium monocarbide are studied by means of Density Functional Theory using the generalized gradient approximation and projector-augmented waves method. The computations are performed considering incorporation sites of UC, such as the U, C and interstitial sites, and Schottky defects. The computed incorporation energies are discussed on the basis of the atomic size of the fission products, their chemical environment and the electronic structure. These energies show that all the studied fission products would preferentially occupy the U site. However, incorporation energies do not provide any further information on the fission product location in the case of unavailability of the sites which is why the concept of solution energies is also used. The solution energies obtained confirm that all the fission products are expected to be more stable on a U site of a single uranium vacancy or within a non-bound Schottky defect in equilibrium conditions.

  2. First-principles study of the stability of fission products in uranium monocarbide

    Energy Technology Data Exchange (ETDEWEB)

    Bevillon, Emile, E-mail: emile.bevillon@yahoo.fr [IRSN, SEMIC, DPAM, LETR, Centre de Cadarache, 13115 Saint Paul Lez Durance (France); Ducher, Roland; Barrachin, Marc; Dubourg, Roland [IRSN, SEMIC, DPAM, LETR, Centre de Cadarache, 13115 Saint Paul Lez Durance (France)

    2012-07-15

    The incorporation and stability of fission products in uranium monocarbide are studied by means of Density Functional Theory using the generalized gradient approximation and projector-augmented waves method. The computations are performed considering incorporation sites of UC, such as the U, C and interstitial sites, and Schottky defects. The computed incorporation energies are discussed on the basis of the atomic size of the fission products, their chemical environment and the electronic structure. These energies show that all the studied fission products would preferentially occupy the U site. However, incorporation energies do not provide any further information on the fission product location in the case of unavailability of the sites which is why the concept of solution energies is also used. The solution energies obtained confirm that all the fission products are expected to be more stable on a U site of a single uranium vacancy or within a non-bound Schottky defect in equilibrium conditions.

  3. Measurement of fission products yields in the quasi-mono-energetic neutron-induced fission of 232Th

    Science.gov (United States)

    Naik, H.; Mukherji, Sadhana; Suryanarayana, S. V.; Jagadeesan, K. C.; Thakare, S. V.; Sharma, S. C.

    2016-08-01

    The cumulative yields of various fission products in the 232Th(n, f) reaction at average neutron energies of 5.42, 7.75, 9.35 and 12.53 MeV have been determined by using an off-line γ-ray spectrometric technique. The neutron beam was produced from the 7Li(p, n) reaction by using the proton energies of 7.8, 12, 16 and 20 MeV. The mass chain yields were obtained from the cumulative fission yields by using the charge distribution correction of medium energy fission. The fine structure in the mass yield distribution was interpreted from the point of nuclear structure effect. On the other hand, the higher yield around mass number 133-134 and 143-144 as well as their complementary products were explained based on the standard I and standard II asymmetric mode of fission. From the mass yield data, the average value of light mass (), heavy mass (), the average number of neutrons () and the peak-to-valley (P / V) ratios at different neutron energies of present work and literature data were obtained in the 232Th(n, f) reaction. The different parameters of the mass yield distribution in the 232Th(n, f) reaction were compared with the similar data in the 232Th(γ, f) reaction at comparable excitation energy and a surprising difference was observed.

  4. Determining isotopic distributions of fission products with a penning trap

    Energy Technology Data Exchange (ETDEWEB)

    Penttilae, H.; Karvonen, P.; Eronen, T.; Elomaa, V.V.; Hager, U.; Hakala, J.; Jokinen, A.; Kankainen, A.; Moore, I.D.; Peraejaervi, K.; Rahaman, S.; Rinta-Antila, S.; Saastamoinen, A.; Sonoda, T.; Aeystoe, J. [University of Jyvaeskylae, Department of Physics, Jyvaeskylae (Finland); Rubchenya, V. [University of Jyvaeskylae, Department of Physics, Jyvaeskylae (Finland); V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation)

    2010-04-15

    A novel method to determine independent yields in particle-induced fission employing the ion guide technique and ion counting after a Penning trap has been developed. The method takes advantage of the fact that a Penning trap can be used as a precision mass filter, which allows an unambiguous identification of the fission fragments. The method was tested with 25MeV and 50MeV proton-induced fission of {sup 238}U. The data is internally reproducible with an accuracy of a few per cent. A satisfactory agreement was obtained with older ion guide yield measurements in 25MeV proton-induced fission. The results for Rb and Cs yields in 50MeV proton-induced fission agree with previous measurements performed at an isotope separator equipped with a chemically selective ion source. (orig.)

  5. Evaluation of Neutron Induced Reactions for 32 Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Il

    2007-02-15

    Neutron cross sections for 32 fission products were evaluated in the neutron-incident energy range from 10{sup -5} eV to 20 MeV. The list of fission products consists of the priority materials for several applications, extended to cover complete isotopic chains for three elements. The full list includes 8 individual isotopes, {sup 95}Mo, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, and 24 isotopes in complete isotopic chains for Nd (8), Sm (9) and Dy (7). Our evaluation methodology covers both the low energy region and the fast neutron region.In the low energy region, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. This resource was used to infer both the thermal values and the resolved resonance parameters that were validated against the capture resonance integrals. In the unresolved resonance region we performed the additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data.In the fast neutron region our evaluations are based on the nuclear reaction model code EMPIRE-2.19 validated against the experimental data. EMPIRE is the modular system of codes consisting of many nuclear reaction models, including the spherical and deformed Optical Model, Hauser-Feshbach theory with the width fluctuation correction and complete gamma-ray emission cascade, DWBA, Multi-step Direct and Multi-step Compound models, and several versions of the phenomenological preequilibrium models. The code is equipped with a power full GUI, allowing an easy access to support libraries such as RIPL and CSISRS, the graphical package, as well the utility codes for formatting and checking. In general, in our calculations we used the Reference Input Parameter Library, RIPL, for the initial set model parameters. These parameters were properly adjusted to reproduce the available experimental data taken from the CSISRS library. Our evaluations cover cross

  6. High-Resolution Correlated Fission Product Measurements of 235U (nth , f) with SPIDER

    Science.gov (United States)

    Shields, Dan; Spider Team

    2015-10-01

    The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) has obtained high-resolution, moderate-efficiency, correlated fission product data needed for many applications including the modeling of next generation nuclear reactors, stockpile stewardship, and the fundamental understanding of the fission process. SPIDER simultaneously measures velocity and energy of both fission products to calculate fission product yields (FPYs), neutron multiplicity (ν), and total kinetic energy (TKE). These data will be some of the first of their kind available to nuclear data evaluations. An overview of the SPIDER detector, analytical method, and preliminary results for 235U (nth , f) will be presented. LA-UR-15-20130 This work benefited from the use of the LANSCE accelerator facility and was performed under the auspices of the US Department of Energy by Los Alamos Security, LLC under Contract DE-AC52-06NA25396.

  7. Fabrication of atomized uranium dispersion targets for fission mo production

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Moonsoo; Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Lee, Jong Hyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Among radioisotopes for medical diagnosis, Tc-99m is most widely used. Mo-99 produced from nuclear fission of uranium in research reactors is the key radioisotopes for Tc-99m generators. Generally, major producers of Mo-99 still use targets containing highly enriched uranium (HEU). However, the international non-proliferation policy emphasizes the minimization of the use of HEU in medical radioisotopes production nowadays. Therefore, low enriched uranium (LEU) targets have been developed by casting and crushing of UAl{sub 2} compounds. The UAl{sub 2} particle dispersed target has a lower U-235 density when compared to HEU targets. In order to improve the low production efficiency of LEU targets, target designers try to develop high uranium density targets with LEU. KAERI has proposed that high density uranium alloys, instead of UAl{sub 2}, can be used as dispersing particles in an aluminum matrix. While it is very difficult to fabricate uranium alloys powder by grinding or crushing, spherical powder of uranium alloys can be produced easily by centrifugal atomization. Mini-size targets with 3, 6, and 9 g-U/cc were fabricated in this study to investigate the feasibility of high density targets with atomized uranium particles. The microstructural changes after thermal treatments were observed to analyze the interaction behavior of uranium particles and aluminum matrix.

  8. Study of the fission products fixation in the hydroxyapatite mineral

    International Nuclear Information System (INIS)

    In this research work, sorption properties of hydroxyapatite in aqueous solutions were studied using Na+ and K+ ion behavior. In addition, the fission products 99Tc and 107Pd uptake was studied to determine their sorption mechanisms on hydroxyapatite. This research was conducted in two stages. The first stage aimed to identify surface reactive sites of hydroxyapatite surface. This surface study was performed by the radiotracer method using 24Na and 42K radionuclides and applying the ion-exchange theory. It provides evidence in terms of the saturation curves of individual behaviour of the Na+ and K+ cations. Hydroxyapatite reactive sites were identified and quantified from the results and application of the ion-exchange model: a mono-functional site of 0.28 mmol g-1 for the sodium hydroxylate form and a dipr otic site with two saturation curves of 0.14 mmol g-1 each, for the sodium phosphate form. In a second stage, the sorption of fission products, Tc and Pd, on hydroxyapatite was studied. This sorption was expressed in terms of distribution coefficients obtained with equivalent radiotracers: 99mTc and 109Pd. Tc presented a low sorption affinity on hydroxyapatite in aqueous medium 0.02 M NaH2PO4 and the results also show that Tc is not sorbed from perchlorate medium (0.01 M Ca(ClO4)2). Sorption behaviour of Pd(II) on hydroxyapatite was studied for different experimental conditions, with parameter such as: ph, aqueous medium (0.01 M NaClO4, 0.01 M and 0.025 M Ca(ClO4)2, and 0.02 M NaH2PO4), the solid solution ratio (10, 4 and 0.020 g/L), and the palladium concentration were studied. Pd sorption was complete at solid-solution ratios 10 and 4 g/L. A strong sorption affinity of hydroxyapatite for palladium was obtained at solid-solution ratio 0.020 g/L. In the interpretation of the results it was considered the aqueous chemistry of palladium, solid dissolution, as well as the existence of reactive sites at the hydroxyapatite surface. The distribution coefficients were

  9. Radiation Damage and Fission Product Release in Zirconium Nitride

    Energy Technology Data Exchange (ETDEWEB)

    Egeland, Gerald W. [New Mexico Inst. of Mining and Technology, Socorro, NM (United States)

    2005-08-29

    Zirconium nitride is a material of interest to the AFCI program due to some of its particular properties, such as its high melting point, strength and thermal conductivity. It is to be used as an inert matrix or diluent with a nuclear fuel based on transuranics. As such, it must sustain not only high temperatures, but also continuous irradiation from fission and decay products. This study addresses the issues of irradiation damage and fission product retention in zirconium nitride through an assessment of defects that are produced, how they react, and how predictions can be made as to the overall lifespan of the complete nuclear fuel package. Ion irradiation experiments are a standard method for producing radiation damage to a surface for observation. Cryogenic irradiations are performed to produce the maximum accumulation of defects, while elevated temperature irradiations may be used to allow defects to migrate and react to form clusters and loops. Cross-sectional transmission electron microscopy and grazing-incidence x-ray diffractometry were used in evaluating the effects that irradiation has on the crystal structure and microstructure of the material. Other techniques were employed to evaluate physical effects, such as nanoindentation and helium release measurements. Results of the irradiations showed that, at cryogenic temperatures, ZrN withstood over 200 displacements per atom without amorphization. No significant change to the lattice or microstructure was observed. At elevated temperatures, the large amount of damage showed mobility, but did not anneal significantly. Defect clustering was possibly observed, yet the size was too small to evaluate, and bubble formation was not observed. Defects, specifically nitrogen vacancies, affect the mechanical behavior of ZrN dramatically. Current and previous work on dislocations shows a distinct change in slip plane, which is evidence of the bonding characteristics. The stacking-fault energy changes dramatically with

  10. Fission product behavior during the PBF [Power Burst Facility] Severe Fuel Damage Test 1-1

    International Nuclear Information System (INIS)

    In response to the accident at Three Mile Island Unit 2 (TMI-2), the United States Nuclear Regulatory Commission (USNRC) initiated a series of Severe Fuel Damage tests that were performed in the Power Burst Facility at the Idaho National Engineering Laboratory to obtain data necessary to understand (a) fission product release, transport, and deposition; (b) hydrogen generation; and (c) fuel/cladding material behavior during degraded core accidents. Data are presented about fission product behavior noted during the second experiment of this series, the Severe Fuel Damage Test 1-1, with an in-depth analysis of the fission product release, transport, and deposition phenomena that were observed. Real-time release and transport data of certain fission products were obtained from on-line gamma spectroscopy measurements. Liquid and gas effluent grab samples were collected at selected periods during the test transient. Additional information was obtained from steamline deposition analysis. From these and other data, fission product release rates and total release fractions are estimated and compared with predicted release behavior using current models. Fission product distributions and a mass balance are also summarized, and certain probable chemical forms are predicted for iodine, cesium, and tellurium. An in-depth evaluation of phenomena affecting the behavior of the high-volatility fission products - xenon, krypton, iodine, cesium, and tellurium - is presented. Analysis indicates that volatile release from fuel is strongly influenced by parameters other than fuel temperature. Fission product behavior during transport through the Power Burst Facility effluent line to the fission product monitoring system is assessed. Tellurium release behavior is also examined relatve to the extent of Zircaloy cladding oxidation. 81 fig., 53 tabs

  11. FITPULS: a code for obtaining analytic fits to aggregate fission-product decay-energy spectra

    International Nuclear Information System (INIS)

    The operation and input to the FITPULS code, recently updated to utilize interactive graphics, are described. The code is designed to retrieve data from a library containing aggregate fine-group spectra (150 energy groups) from fission products, collapse the data to few groups (up to 25), and fit the resulting spectra along the cooling time axis with a linear combination of exponential functions. Also given in this report are useful results for aggregate gamma and beta spectra from the decay of fission products released from 235U irradiated with a pulse (10-4 s irradiation time) of thermal neutrons. These fits are given in 22 energy groups that are the first 22 groups of the LASL 25-group decay-energy group structure, and the data are expressed both as MeV per fission second and particles per fission second; these pulse functions are readily folded into finite fission histories. 65 figures, 11 tables

  12. Background and Derivation of ANS-5.4 Standard Fission Product Release Model

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E.; Turnbull, Andrew J.

    2010-01-29

    This background report describes the technical basis for the newly proposed American Nuclear Society (ANS) 5.4 standard, Methods for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuels. The proposed ANS 5.4 standard provides a methodology for determining the radioactive fission product releases from the fuel for use in assessing radiological consequences of postulated accidents that do not involve abrupt power transients. When coupled with isotopic yields, this method establishes the 'gap activity,' which is the inventory of volatile fission products that are released from the fuel rod if the cladding are breached.

  13. Fast-neutron interaction with the fission product 103Rh

    International Nuclear Information System (INIS)

    Neutron total and differential elastic- and inelastic-scattering cross sections of 103Rh are measured from ∼ 0.7 to 4.5 MeV (totals) and from ∼ 1.5 to 10 MeV (scattering) with sufficient detail to define the energy-averaged behavior of the neutron processes. Neutrons corresponding to excitations of groups of levels at 334 ± 13, 536 ± 10, 648 ± 25, 796 ± 20, 864 ± 22, 1120 ± 22, 1279 ± 60, 1481 ± 27 and 1683 ± 39 keV were observed. Additional groups at 1840 ± 79 and 1991 ± 71 key were tentatively identified. Assuming the target is a collective nucleus reasonably approximated by a simple one-phonon vibrator, spherical-optical, dispersive-optical, and coupled-channels models were developed from the data base with attention to the parameterization of the large inelastic-scattering cross sections. The physical properties of these models are compared with theoretical predictions and the systematics of similar model parameterizations in this mass region. In particular, it is shown that the inelastic-scattering cross section of the 103Rh fission product is large at the relatively low energies of applied interest

  14. Fission-Product Development Laboratory cell-decommissioning project plan

    International Nuclear Information System (INIS)

    The Fission Product Development Laboratory (FPDL) at Oak Ridge National Laboratory (ORNL) was a full-scale processing facility for separating megacurie quantities of 90Sr, 137Cs, and 144Ce for a variety of source applications, operating at full capacity from 1958 to 1975. Since facility shutdown, the inactive portions of the FPDL have been maintained in a protective storage mode as part of the ORNL Surplus Facilities Management Program (SFMP). Due to the significant radio-nuclide inventory remaining in the facility, the high surveillance and maintenance costs necessary to assure radionuclide containment, and the potential for reuse of the facility by other programs, the decommissioning of the inactive portions of the FPDL has been given a high priority by the SFMP. In response to this program direction, plans are being made for initiation of these activities in late FY 1983. This project plan has been prepared to satisfy the program documentation requirements for SFMP project planning. The plan outlines the scope of the proposed effort, describes the proposed methods of project accomplishment, and provides estimates of the project resource needs and schedule

  15. Results of recent ORNL fission product release tests

    International Nuclear Information System (INIS)

    The effects of time, high temperature, and atmosphere were explored in ORNL tests VI-2, VI-3, and VI-4. These tests were performed using vertically oriented segments of Zircaloy-clad UO2 fuel that had been irradiated to ∼42 MWd/kg U in the Belgian BR3 reactor. Tests in steam were conducted at temperature plateaus of 2,000, 2,300, and 2,700 K; test VI-4 was conducted in a hydrogen-helium atmosphere at 2,450 K. Results of test VI-2, which were run for 60 min at 2,300 K, showed that 63% of the fission product cesium had been released. The release rate for cesium, expressed as a fraction of the remaining inventory released per minute, decreased tenfold during the test. The fuel in test VI-3 was heated at 2,000 and 2,700 K for 20 min at each temperature. Essentially 100% of the cesium, krypton, and antimony were released. No measurable release of either cerium or europium was observed. In both VI-2 and VI-3, the steam oxidation of the Zircaloy cladding followed the Urbanick-Heidrick rate data. Modeling work shows that Booth diffusion coefficients (random diffusion from fuel grains) correlate all of the test results very well

  16. Treatment and solidification of high active fission product solutions

    International Nuclear Information System (INIS)

    On reprocessing spent fuel elements, > 97% of the fission products are found in the high active waste (HAW) solution. In order to avoid large amounts of sludge formation arising from phosphates produced by TBP degradation during evaporation and storage of these high level wastes, the suspended and dissolved TBP must be removed immediately from the HAW. It is proposed to separate the TBP by steam-stripping. The the HAW will be concentrated in an evaporator, the concentration factor depending on the amount of sludge formation and the heat content of the concentrate. These concentrates may be stored for short periods in stainless steel tanks. Acid concentration and waste volume may be further reduced by in-tank denitration and evaporation. For vitrification of the HAW liquid feed, ceramic melters are being developed universally. The first active plant to use a liquid feed ceramic melter is the German plant PAMELA, which is being built at Mol in Belgium, with an operational date of 1985

  17. TMI-2 (Three Mile Island) fission product inventory program: FY-85 status report

    Energy Technology Data Exchange (ETDEWEB)

    Langer, S; Croney, S T; Akers, D W; Russell, M L

    1986-11-01

    This report presents the status of the TMI-2 fission product inventory program through May 1985. The fission product inventory program is an assessment of the location of fission products distributed in the plant as a result of the TMI-2 accident. Included in this report are principal results of samples from the reactor building where most of the mobile fission products (i.e., radiocesium and iodine) are expected to be found. The data are now complete enough for most reactor components; therefore, it is possible to direct the balance of the examination and sampling program to areas and components where it is likely to be most productive. Those areas are the reactor core and the reactor building basement, with emphasis on the currently unsampled portions of the core.

  18. HTGR fuels and core development program. Quarterly progress report for the period ending May 31, 1976. [Graphite and fuels irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1976-06-30

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and the data are presented in tables, graphs, and photographs.

  19. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  20. HTGR fuels and core development program. Quarterly progress report for the period ending November 30, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-12-01

    The work reported here includes studies of basic fission product transport mechanisms, core graphite development and testing, and the development and testing of recyclable fuel systems. Materials studied include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  1. Heat and Fission Product Transport in a Molten U-Zr-O Pool With Crust

    International Nuclear Information System (INIS)

    Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry and the change of pool geometry during the numerical calculation was neglected. Results of the numerical calculation revealed that the peak temperature of the molten pool significantly decreased and most of the volatile fission products were released from the molten pool during the accident. (authors)

  2. Prompt γ-ray production in neutron-induced fission of 239Pu

    Science.gov (United States)

    Ullmann, J. L.; Bond, E. M.; Bredeweg, T. A.; Couture, A.; Haight, R. C.; Jandel, M.; Kawano, T.; Lee, H. Y.; O'Donnell, J. M.; Hayes, A. C.; Stetcu, I.; Taddeucci, T. N.; Talou, P.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Chyzh, A.; Gostic, J.; Henderson, R.; Kwan, E.; Wu, C. Y.

    2013-04-01

    Background: The prompt gamma-ray spectrum from fission is important for understanding the physics of nuclear fission, and also in applications involving fission. Relatively few measurements of the prompt gamma spectrum from 239Pu(n,f) have been published.Purpose: This experiment measured the multiplicity, individual gamma energy spectrum, and total gamma energy spectrum of prompt fission gamma rays from 239Pu(n,f) in the neutron energy range from thermal to 30 keV, to test models of fission and to provide information for applications.Method: Gamma rays from neutron-induced fission of 239Pu were measured using the DANCE gamma-ray calorimeter. Fission events were tagged by detecting fission products in a parallel-plate avalanche counter in the center of DANCE. The measurements were corrected for detector response using a geant4 model of DANCE. A detailed analysis for the gamma rays from the 1+ resonance complex at 10.93 eV is presented.Results: A six-parameter analytical parametrization of the fission gamma-ray spectrum was obtained. A Monte Carlo Hauser-Feshbach calculation provided good general agreement with the data, but some differences remain to be resolved.Conclusions: An analytic parametrization can be made of the gamma-ray multiplicity, energy distribution, and total-energy distribution for the prompt gamma rays following neutron-induced fission of 239Pu. This parametrization may be useful for applications. Modern Monte Carlo Hauser-Feshbach calculations can do a good job of calculating the fission gamma-ray emission spectrum, although some details remain to be understood.

  3. Fission product inventory calculation by a CASMO/ORIGEN coupling program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

  4. New Fission-Product Waste Forms: Development and Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Alexandra Navrotsky

    2010-07-30

    Research performed on the program “New Fission Product Waste Forms: Development and Characterization,” in the last three years has fulfilled the objectives of the proposal which were to 1) establish ceramic waste forms for disposing of Cs, Sr and minor actinides, 2) fully characterize the phase relationships, structures and thermodynamic and kinetic stabilities of promising waste forms, 3) establish a sound technical basis for understanding key waste form properties, such as melting temperatures and aqueous durability, based on an in-depth understanding of waste form structures and thermochemistry, and 4) establish synthesis, testing, scaleup and commercialization routes for wasteform implementation through out in-kind collaborations. In addition, since Cs and Sr form new elements by radioactive decay, the behavior and thermodynamics of waste forms containing different proportions of Cs, Sr and their decay products were discovered using non-radioactive analogues. Collaborations among researchers from three institutions, UC Davis, Sandia National Laboratories, and Shott Inc., were formed to perform the primary work on the program. The unique expertise of each of the members in the areas of waste form development, structure/property relationships, hydrothermal and high temperature synthesis, crystal/glass production, and thermochemistry was critical to program success. In addition, collaborations with the Brigham Young Univeristy, Ben Gurion University, and Los Alamos National Laboratory, were established for standard entropies of ceramic waste forms, sol-gel synthesis, and high temperature synthesis. This work has had a significant impact in a number of areas. First, the studies of the thermodynamic stability of the mineral analogues provided an important technical foundation for assessment the viability of multicomponent oxide phases for Cs and Sr removal. Moreover, the thermodynamic data discovered in this program established information on the reaction

  5. Results of fission products β decay properties measurement performed with a total absorption spectrometer

    Directory of Open Access Journals (Sweden)

    Zakari-Issoufou A.-A.

    2014-03-01

    Full Text Available β-decay properties of fission products are very important for applied reactor physics, for instance to estimate the decay heat released immediately after the reactor shutdown and to estimate the ν¯$\\bar \

  6. Fuels and fission products clean up for molten salt reactor of the incinerator type

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Gorbunov, V.; Zakirov, R. [RRC-Karchatov Institute, Moscow (Russian Federation)

    2000-07-01

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of plutonium, minor actinides and fission products, when the reactor and fission product cleanup unit are planned as an integral system. This contribution summarizes the reasons which led to selection of the salt compositions for the molten salt reactor of the TRU incinerator type (MSB). Special characteristics of behavior of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor (MSR) programs and relates then to the separation requirements for the MSB concept. Status and development needs in the thermodynamic properties of fluorides and fission product cleanup methods (with emphasis on actinides-lanthanides separation) are discussed. (authors)

  7. Fuels and fission products clean up for molten salt reactor of the incinerator type

    International Nuclear Information System (INIS)

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of plutonium, minor actinides and fission products, when the reactor and fission product cleanup unit are planned as an integral system. This contribution summarizes the reasons which led to selection of the salt compositions for the molten salt reactor of the TRU incinerator type (MSB). Special characteristics of behavior of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor (MSR) programs and relates then to the separation requirements for the MSB concept. Status and development needs in the thermodynamic properties of fluorides and fission product cleanup methods (with emphasis on actinides-lanthanides separation) are discussed. (authors)

  8. Fission product and actinide data evaluations for ENDF/B--V

    Energy Technology Data Exchange (ETDEWEB)

    Schenter, R.E.

    1978-05-01

    The planned content and performance of fission product and actinide nuclide evaluations for the ENDF/B-V collection of data are reviewed. Representative values of parameters for a few nuclides are shown. 10 figures, 5 tables. (RWR)

  9. The LANL C-NR counting room and fission product yields

    Energy Technology Data Exchange (ETDEWEB)

    Jackman, Kevin Richard [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

    2015-09-21

    This PowerPoint presentation focused on the following areas: LANL C-NR counting room; Fission product yields; Los Alamos Neutron wheel experiments; Recent experiments ad NCERC; and Post-detonation nuclear forensics

  10. User's manual of ART code for analyzing fission product transport behavior during core meltdown accident

    International Nuclear Information System (INIS)

    In a probabilistic risk assessment (PRA) it has been recognized that a core meltdown accident with a large amount of fission products released to the environment is a dominant contributor to public risk. For the evaluation of the risk, information about source terms are inevitable. In order to analyze fission product transport behavior and to evaluate source terms during a core meltdown accident, the ART code has been developed. The ART code has the following features: (1) It can treat fission product transport behavior both in a primary system and a containment system, (2) It models fission product transport caused by both gas flow and liquid flow, and (3) It includes a detailed model about transport behavior of aerosols which are released in quantity during a core meltdown accident. This report is a user's manual for the ART code and includes description of modeling, input/output data and a sample run. (author)

  11. Shielding calculation of a hot cell for the processing of fission products

    International Nuclear Information System (INIS)

    A dose rate estimation is made for an operator of a lead wall, fission products processing hot cell, in a distance of 50 cm from the emission source, at Brazilian Institute of Nuclear Engineering (IEN). (L.C.J.A.)

  12. Compilation of data related to fission products. I - Chain total yields

    International Nuclear Information System (INIS)

    As the theoretical study of the formation and evolution of fission products in a pile fuel requires the knowledge of a large number of data (fission product characteristics, parameters related to fission mechanism), and in the frame of such a type a study which aimed at taking, non only fuel irradiation conditions and fuel composition, but also the evolution of these features in time into account, the authors have been leaded to perform a large compilation of data required by the calculation, and also to make a choice among the available data. This volume gathers data related to the total yields of fission product chains. The first part contains chain total yields from different documents. These data deal with various energies and concern the following products: 233U, 235U, 238U, 239Pu, 241Pu. The second part proposes curves which, for 235U and 239Pu, give the total yields as a function of incident neutron energy

  13. Analysis of fission-product effects in a Fast Mixed-Spectrum Reactor concept

    International Nuclear Information System (INIS)

    The Fast Mixed-Spectrum Reactor (FMSR) concept has been proposed by BNL as a means of alleviating certain nonproliferation concerns relating to civilian nuclear power. This breeder reactor concept has been tailored to operate on natural uranium feed (after initial startup), thus eliminating the need for fuel reprocessing. The fissile material required for criticality is produced, in situ, from the fertile feed material. This process requires that large burnup and fluence levels be achievable, which, in turn, necessarily implies that large fission-product inventories will exist in the reactor. It was the purpose of this study to investigate the effects of large fission-product inventories and to analyze the effect of burnup on fission-product nuclide distributions and effective cross sections. In addition, BNL requested that a representative 50-group fission-product library be generated for use in FMSR design calculations

  14. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  15. Basic analysis of regularized series and products

    CERN Document Server

    Jorgenson, Jay A

    1993-01-01

    Analytic number theory and part of the spectral theory of operators (differential, pseudo-differential, elliptic, etc.) are being merged under amore general analytic theory of regularized products of certain sequences satisfying a few basic axioms. The most basic examples consist of the sequence of natural numbers, the sequence of zeros with positive imaginary part of the Riemann zeta function, and the sequence of eigenvalues, say of a positive Laplacian on a compact or certain cases of non-compact manifolds. The resulting theory is applicable to ergodic theory and dynamical systems; to the zeta and L-functions of number theory or representation theory and modular forms; to Selberg-like zeta functions; andto the theory of regularized determinants familiar in physics and other parts of mathematics. Aside from presenting a systematic account of widely scattered results, the theory also provides new results. One part of the theory deals with complex analytic properties, and another part deals with Fourier analys...

  16. Mass spectrometric study of the release of volatile fission products from irradiated LWR fuel

    International Nuclear Information System (INIS)

    The objective of these studies is to experimentally determine the chemical form and the rate of release of volatile fission product species from defected irradiated LWR reactor fuel pins. After release from the defected fuel pin the gaseous species immediately enters the ionizer of a quadrupole mass spectrometer thus ensuring that their chemical form is not likely to be changed prior to identification and measurement. These studies differ from prior studies in that: (1) the chemical form of the volatile fission products will be determined; and (2) the detection and measurement method does not depend on the radioactivity of the fission product element. Information on the chemical form of the released fission product species will enable a more accurate description of their transport and reaction in the primary system. These studies are also expected to yield information on the reaction of fission products after release from the fuel oxide with the zircaloy cladding. The results of these studies are expected to increase the understanding of the first step in the release of fission products by irradiated fuel and therefore help in the accurate prediction of source terms

  17. ORNL studies of fission product release under LWR severe accident conditions

    International Nuclear Information System (INIS)

    The large inventories of radioactive fission products in irradiated fuel represent the principal personnel hazards from nuclear reactors. A large fraction of the existing fission-product release data has been collected from experiments at Oak Ridge National Laboratory (ORNL). Tests of high-burnup light-water-reactor fuel, and also simulated fuel with fission-product tracers, have been conducted in an induction furnace at temperatures up to 2,700 K. In addition to the total releases, on-line release data for 85Kr and 137Cs at 1-min intervals throughout the tests provided release-rate values. The most valuable fission-product elements - krypton, iodine, and cesium - are released almost totally at the highest temperatures, with little effect of atmosphere, but the releases of fission product strontium, molybdenum, ruthenium, tellurium, antimony, barium, and europium are sensitive to atmosphere. Data for krypton and cesium releases have been used to develop the ORNL Diffusion Release Model, a simple, single-atom model that reliably predicts the release of volatile fission products. Studies of transport behavior and the chemical forms of released elements, as well as fuel melt progression, have been included also. 52 refs., 14 figs., 4 tabs

  18. Release of short-lived fission products from operating UO2 fuel under oxidizing conditions

    International Nuclear Information System (INIS)

    We have examined, using a sweep gas technique, the effects of oxidation on the release of short-lived fission products from UO2 fuel elements which defected, and from intact elements where water was deliberately added via the sweep gas. In both types of experiment, the short-lived fission products showed qualitatively similar behaviour. On initiating oxidation, there was a transient period during which the release of fission products depended on their inventory in the UO2. After the transient, diffusion-controlled steady state release resumed. Fission gases with short-lived precursors were released at about twice the pre-transient rate while those with long-lived precursors (Xe-133, -135m, -135) were released at about 4.5 times the pre-transient rate. These results suggest that, under oxidizing conditions, the effective diffusion coefficient for the iodines increases more than that of the rare gases

  19. Detecting special nuclear materials in containers using high-energy gamma rays emitted by fission products

    Science.gov (United States)

    Norman, Eric B.; Prussin, Stanley G.

    2007-10-02

    A method and a system for detecting the presence of special nuclear materials in a container. The system and its method include irradiating the container with an energetic beam, so as to induce a fission in the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  20. Heat and fission product transport in molten core material pool with crust

    International Nuclear Information System (INIS)

    Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention

  1. Photo-fission for the production of radioactive beams ALTO project

    Energy Technology Data Exchange (ETDEWEB)

    Essabaa, S. E-mail: essabaa@ipno.in2p3.fr; Arianer, J.; Ausset, P.; Bajeat, O.; Baronick, J.P.; Clapier, F.; Coacolo, L.; Donzaud, C.; Ducourtieux, M.; Gales, S.; Gardes, D.; Grialou, D.; Hosni, F.; Guillemaud-Mueller, D.; Ibrahim, F.; Junquera, T.; Lau, C.; Le Blanc, F.; Lefort, H.; Le Scornet, J.C.; Lesrel, J.; Mueller, A.C.; Obert, J.; Perru, O.; Potier, J.C.; Proust, J.; Pougheon, F.; Roussiere, B.; Rouviere, N.; Sauvage, J.; Sorlin, O.; Tkatchenko, A.; Verney, D.; Waast, B.; Rinolfi, L.; Rossat, G.; Forkel-Wirth, D.; Muller, A.; Bienvenu, G.; Bourdon, J.-C.; Garvey, T.; Jacquemard, B.; Omeich, M

    2003-05-01

    In order to probe neutron rich radioactive noble gases produced by photo-fission, a PARRNe-1 experiment (Production d'Atomes Radioactifs Riches en Neutrons) has been carried out at CERN. The incident electron beam of 50 MeV was delivered by the LIL machine: LEP Injector Linac. The experiment allowed us to compare under the same conditions two production methods of radioactive noble gases: fission induced by fast neutrons and photo-fission. The obtained results show that the use of the electrons is a promising mode to get intense neutron rich ion beams. After the success of this photo-fission experiment, a conceptual design for the installation at IPN Orsay of a 50 MeV electron accelerator close to the PARRNe-2 device has been worked out: ALTO Project. This work has started within a collaboration between IPNO, LAL (Laboratoire de l'Accelerateur Lineaire) and CERN groups.

  2. Photo-fission for the production of radioactive beams ALTO project

    CERN Document Server

    Essabaa, S; Ausset, P; Bajeat, O; Baronick, J P; Clapier, F; Coacolo, J L; Donzaud, C; Ducourtieux, M; Galas, S; Gardes, D; Grialou, D; Hosni, F; Guillemaud-Müller, D; Ibrahim, F; Junquera, T; Lau, C; Le Blanc, F; Lefort, H; Le Scornet, J C; Lesrel, J; Müller, A C; Obert, J; Perru, O; Potier, J C; Proust, J; Pougheon, F; Roussière, B; Rouvière, N; Sauvage, J; Sorlin, O; Tkatchenko, A; Verney, D; Waast, B; Rinolfi, Louis; Rossat, G; Forkel-Wirth, Doris; Müller, A; Bienvenu, G; Bourdon, J C; Garvey, Terence; Jacquemard, B; Omeich, M

    2003-01-01

    In order to probe neutron rich radioactive noble gases produced by photo-fission, a PARRNe-1 experiment (Production d'Atomes Radioactifs Riches en Neutrons) has been carried out at CERN. The incident electron beam of 50 MeV was delivered by the LIL machine: LEP Injector Linac. The experiment allowed us to compare under the same conditions two production methods of radioactive noble gases: fission induced by fast neutrons and photo-fission. The obtained results show that the use of the electrons is a promising mode to get intense neutron rich ion beams. After the success of this photo-fission experiment, a conceptual design for the installation at IPN Orsay of a 50 MeV electron accelerator close to the PARRNe-2 device has been worked out: ALTO Project. This work has started within a collaboration between IPNO, LAL (Laboratoire de l'Accelerateur Lineaire) and CERN groups.

  3. Experimental determination of the antineutrino spectrum of the fission products of {sup 238}U

    Energy Technology Data Exchange (ETDEWEB)

    Haag, Nils-Holger

    2013-10-09

    Fission of {sup 238}U contributes about 10 % to the antineutrino emission of a pressurized water reactor. In the present thesis, the beta spectrum of the fission products of {sup 238}U was determined in an experiment at the neutron source FRM II. This beta spectrum was subsequently converted into an antineutrino spectrum. This first measurement of the antineutrino spectrum supports all current and future reactor antineutrino experiments.

  4. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    The release of gaseous (krypton and xenon) and iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assumes precursor-corrected 'Booth diffusional release' in the UO2 and subsequent holdup in the fuel-to-sheath gap. Transport in the gap is separately modelled with a phenomenological rate constant (assuming release from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from possessing various states of defection are use in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small drilled hole. A second element was machined with 23 slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Comparison of measured release data with calculated values from the model yields estimates of empirical diffusion coefficients for the radioactive species in the UO2 (1.56 x 10-10 to 7.30 x 10-9 s-1), as well as escape rate constants (7.85 x 10-7 to 3.44 x 10-5 s-1) and diffusion coefficients (3.39 x 10-5 to 4.88 x 10-2 cm2/s) for these in the fuel-to-sheath gap. Analyses also enable identification of the various rate-controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is 'Booth diffusion'; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-heath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this work, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size)

  5. Basic principles of maximizing dental office productivity.

    Science.gov (United States)

    Mamoun, John

    2012-01-01

    To maximize office productivity, dentists should focus on performing tasks that only they can perform and not spend office hours performing tasks that can be delegated to non-dentist personnel. An important element of maximizing productivity is to arrange the schedule so that multiple patients are seated simultaneously in different operatories. Doing so allows the dentist to work on one patient in one operatory without needing to wait for local anesthetic to take effect on another patient in another operatory, or for assistants to perform tasks (such as cleaning up, taking radiographs, performing prophylaxis, or transporting and preparing equipment and supplies) in other operatories. Another way to improve productivity is to structure procedures so that fewer steps are needed to set up and implement them. In addition, during procedures, four-handed dental passing methods can be used to provide the dentist with supplies or equipment when needed. This article reviews basic principles of maximizing dental office productivity, based on the author's observations of business logistics used by various dental offices. PMID:22414506

  6. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

    International Nuclear Information System (INIS)

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual keff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

  7. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  8. Electrochemical dissolution of actinides and fission products in aqueous solutions: case of Mo2C

    International Nuclear Information System (INIS)

    UC and (U-Pu) mixed carbide are potential fuels for the new High Temperature Gas-cooled Reactors (HTGR) under study. The fuel reprocessing is to be reconsidered to provide simple actinide/fission products separation. To develop the methodological aspects, we have first studied the electrochemical dissolution of molybdenum carbide (MO2C) in basic media. The corrosion tests have shown no passivation in NaOH and carbonate buffer solution except in 4 M NaOH solution. The electrochemical dissolution is efficient in both media. Nevertheless, as predicted by voltametry, the dissolution rate calculated by weight loss of the MO2C pellet is function of the electrolysis potential: the rate increases with NaOH concentration, pH or electrolysis potential and the dissolution is more efficient in NaOH than in carbonate buffer solution. Finally, the oxidation potentials of MO2C in basic media were also determined with cavity micro-electrode and compared with those obtained with pellet. (authors)

  9. Fission-product yields for thermal-neutron fission of 243Cm determined from measurements with a high-resolution low-energy germanium gamma-ray detector

    International Nuclear Information System (INIS)

    Cumulative fission-product yields have been determined for 13 gamma rays emitted during the decay of 12 fission products created by thermal-neutron fission of 243Cm. A high-resolution low-energy germanium detector was used to measure the pulse-height spectra of gamma rays emitted from a 77-nanogram sample of 243Cm after the sample had been irradiated by thermal neutrons. Analysis of the data resulted in the identification and matching of gamma-ray energies and half-lives to individual radioisotopes. From these results, 12 cumulative fission product yields were deduced for radionuclides with half-lives between 4.2 min and 84.2 min. 7 references

  10. Uncertainties in fission-product decay-heat calculations

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, K.; Ohta, H.; Miyazono, T.; Tasaka, K. [Nagoya Univ. (Japan)

    1997-03-01

    The present precision of the aggregate decay heat calculations is studied quantitatively for 50 fissioning systems. In this evaluation, nuclear data and their uncertainty data are taken from ENDF/B-VI nuclear data library and those which are not available in this library are supplemented by a theoretical consideration. An approximate method is proposed to simplify the evaluation of the uncertainties in the aggregate decay heat calculations so that we can point out easily nuclei which cause large uncertainties in the calculated decay heat values. In this paper, we attempt to clarify the justification of the approximation which was not very clear at the early stage of the study. We find that the aggregate decay heat uncertainties for minor actinides such as Am and Cm isotopes are 3-5 times as large as those for {sup 235}U and {sup 239}Pu. The recommended values by Atomic Energy Society of Japan (AESJ) were given for 3 major fissioning systems, {sup 235}U(t), {sup 239}Pu(t) and {sup 238}U(f). The present results are consistent with the AESJ values for these systems although the two evaluations used different nuclear data libraries and approximations. Therefore, the present results can also be considered to supplement the uncertainty values for the remaining 17 fissioning systems in JNDC2, which were not treated in the AESJ evaluation. Furthermore, we attempt to list nuclear data which cause large uncertainties in decay heat calculations for the future revision of decay and yield data libraries. (author)

  11. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  12. Fission product transport in the high temperature gas-cooled reactor: Theory, program development and verification by recalculation of experiments

    International Nuclear Information System (INIS)

    The high temperature gascooled reactor (HTGR) reaches a special standard in safety because of its high temperature resistent fuel element. After all the possibility of fission product releases can not be excluded without further investigations for HTGRs. The mechanisms of fission product releases, which occur in case of such hypothetical events, are the subject of this work. The main focus of the investigation is how the fission products, which have been released, are re-adsorpted and prevented through this mechanism from being released in the environment. A strong effect of re-adsorption is expected, because experiments have shown that graphite, which is 100% of the core material, has an excellent capability to hold back fission products. With the program tools developed to calculate the fission product transport mechanisms, the corresponding experiments are recalculated and also fission product release calculations are carried out. (orig./HP)

  13. Basic contradiction of the international production cooperation

    Directory of Open Access Journals (Sweden)

    Rinas V. Kashbraziyev

    2015-12-01

    Full Text Available Objective to analyze the contradictions of international industrial cooperation as a driving force for its development. Methods identification and analysis of contradictions in international cooperation were carried out using systematic approach based on general scientific methods of theoretical and empirical research analysis synthesis scientific observation measurement and comparison. On the basis of generalization and analysis of the information contained in the statistical system of the Organization for economic cooperation and development and the global competitiveness reports of the world economic forum the author presents assessment of the level of technological and knowledgeintensity of the economies of certain developed and developing transition countries investment efficiency in science and research their influence on the technological level of production and the degree of technological sovereignty of the mentioned countries. Results the study of the industrialized countriesrsquo experience has shown that the production of hightech products is impossible without integration into a global cooperative network of industrial companies and research institutes. However being included into the global production chains and attracting advanced technologies of production marketing and management the national companies inevitably fall into dependence on foreign import supply. An economic axiom is formulated modern hightech production requires a dramatic expansion of international production cooperation. The main ontological contradiction of international industrial cooperation is revealed characterized by the impact on the improvement of the technological level of production and innovativeness of the national economy on the one hand and simultaneous strengthening of its dependence on foreign partners on the other hand. Scientific novelty on the basis of systematic approach the article reveals contradictions in international cooperation in the

  14. The Effect of Water Injection on the Fission Product Aerosol Behavior in Fukushima Unit 1

    International Nuclear Information System (INIS)

    The most important factor affects human health is fission product that is released from the plant. Fission products usually released with types of aerosol and vapor. The amount of released aerosols out of the plant is crucial, because it can be breathed by people. In this study, the best estimated scenario of Fukushima unit 1 accident was modeled with MELCOR. The amount of released fission product aerosols was estimated according to the amount of added water into reactor pressure vessel (RPV). The analysis of Fukushima unit 1 accident was conducted in view of fission product aerosol release using MELCOR. First of all, thermodynamic results of the plant were compared to the measured data, and then fission product aerosol (CsOH) behavior was calculated with changing the amount of water injection. Water injection affects the amount of aerosol which released into reactor building, because it decreases the temperature of deposition surface. In this study, only aerosol behavior was considered, further study will be conducted including hygroscopic model

  15. Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

    Directory of Open Access Journals (Sweden)

    Salahuddin Asif

    2013-01-01

    Full Text Available Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

  16. Theoretical and experimental studies of the neutron rich fission product yields at intermediate energies

    Directory of Open Access Journals (Sweden)

    Äystö J.

    2012-02-01

    Full Text Available A new method to measure the fission product independent yields employing the ion guide technique and a Penning trap as a precision mass filter, which allows an unambiguous identification of the nuclides is presented. The method was used to determine the independent yields in the proton-induced fission of 232Th and 238U at 25 MeV. The data were analyzed with the consistent model for description of the fission product formation cross section at the projectile energies up to 100 MeV. Pre-compound nucleon emission is described with the two-component exciton model using Monte Carlo method. Decay of excited compound nuclei is treated within time-dependent statistical model with inclusion of the nuclear friction effect. The charge distribution of the primary fragment isobaric chain was considered as a result of frozen quantal fluctuations of the isovector nuclear density. The theoretical predictions of the independent fission product cross sections are used for normalization of the measured fission product isotopic distributions.

  17. A theoretical study of volatile fission products release from oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Paraschiv, M.C.; Paraschiv, A. [Inst. for Nucl. Res., Pitesti (Romania); Grecu, V.V. [University of Bucharest, Faculty of Physics, P.O. Box MG-11, Bucharest (Romania)

    1999-11-01

    Treating the average volume grains as thermodynamically closed subsystems, a method to evaluate the volatile fission products migration at the grain boundary and their release in the void volume of the fuel elements is proposed. The method considers the phenomena of the intergranular bubble growth and interlinkage, grain growth and grain boundary resolution. Analytical solutions of the diffusion problem associated with the volatile fission products behaviour taking into account their direct yield from fission and from precursors simultaneously with the diffusion and decay, irradiation induced resolution and fuel grain growth, during a time-step varying irradiation history have also been derived. The results are very accurate and point out the strong effect of the boundary condition changes on the volatile fission products behaviour when the simultaneous effects of the intergranular bubble coalescence, the precursors, the irradiation induced resolution and grain growth are considered. Comparative analyses versus other similar models of the diffusion of only stable gas species of fission products are also presented. (orig.)

  18. Volatile fission product distributions in LWR spent fuel rods

    International Nuclear Information System (INIS)

    Results presented are from spent fuel characterizations being conducted by the Materials Characterization Center at Pacific Northwest Laboratory on a variety of spent commercial power reactor fuels designated as approved testing materials (ATMs). These ATMs have a variety of burnup levels and fission gas releases; they include fuel from both pressurized water and boiling water reactor designs. The purpose of this work is to provide a source of well-characterized spent fuel for testing in the U.S. Department of Energy Office of Civilian Radioactive Waste Management repository programs and, potentially, other programs

  19. Transmutation analysis considering and explicit fission product treatment based on a coupled Hammer-Technion and Cinder-2 system

    International Nuclear Information System (INIS)

    This work presents a study about neutron absorption in a typical PWR cell by considering an explicit treatment for the fission products. The proposed methodology to treat fission product neutron absorption in a lattice calculation combines the HAMMER-TECHNION and CINDER-2 codes. The fission product chain treatment considers nearly 99% of all original CINDER-2 neutron absorption chain treatment. Parallel to the explicit treatment, a cross section library in the HAMMER-TECHNION code multigroup structure for the fission products was generated using the ENDF/B-V fission product library and processed by NJOY and AMPX-II processing codes. The methodology validation was investigated against two available benchmarks and it was obtained excellent results for the K-Infinity (IAEA-TECDOC-233) as function of burnup and enrichment and for the aggregate quantity sup(σ)2200 in units of barns/fission cross sections (OKAZAKI and SOKOLOWSKI). This work contributed for a better understanding of the fission product neutron absorption in a typical PWR cell and showed that the explicit fission product treatment can be successfully achieved. Besides that the performance of the ENDF/B-V fission product library was accessed. (author)

  20. Results of fission products β decay properties measurement performed with a total absorption spectrometer

    Science.gov (United States)

    Zakari-Issoufou, A.-A.; Porta, A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Agramunt, J.; Äystö, J.; Bowry, M.; Bui, V. M.; Caballero-Folch, R.; Cano-Ott, D.; Eloma, V.; Estévez, E.; Farrelly, G. F.; Garcia, A.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez, A.; Podolyak, Zs.; Penttilä, H.; Regan, P. H.; Rissanen, J.; Rubio, B.; Weber, C.

    2014-03-01

    β-decay properties of fission products are very important for applied reactor physics, for instance to estimate the decay heat released immediately after the reactor shutdown and to estimate the bar ν flux emitted. An accurate estimation of the decay heat and the bar ν emitted flux from reactors, are necessary for purposes such as reactors operation safety and non-proliferation. In order to improve the precision in the prediction for these quantities, the bias due to the Pandemonium effect affecting some important fission product data has to be corrected. New measurements of fission products β-decay, not sensitive to this effect, have been performed with a Total Absorption Spectrometer (TAS) at the JYFL facility of Jyväskylä. An overview of the TAS technique and first results from the 2009 campaign will be presented.

  1. Fission product vapour - aerosol interactions in the containment: simulant fuel studies

    International Nuclear Information System (INIS)

    Experiments have been conducted in the Falcon facility to study the interaction of fission product vapours released from simulant fuel samples with control rod aerosols. The aerosols generated from both the control rod and fuel sample were chemically distinct and had different deposition characteristics. Extensive interaction was observed between the fission product vapours and the control rod aerosol. The two dominant mechanisms were condensation of the vapours onto the aerosol, and chemical reactions between the two components; sorption phenomena were believed to be only of secondary importance. The interaction of fission product vapours and reactor materials aerosols could have a major impact on the transport characteristics of the radioactive emission from a degrading core. (author)

  2. Investigation of the diffusion of atomic fission products in UC by density functional calculations

    Energy Technology Data Exchange (ETDEWEB)

    Bévillon, Émile, E-mail: emile.bevillon@yahoo.fr [IRSN, SEMIC, DPAM, LETR, Centre de Cadarache, 13115 Saint Paul Lez Durance (France); Ducher, Roland; Barrachin, Marc; Dubourg, Roland [IRSN, SEMIC, DPAM, LETR, Centre de Cadarache, 13115 Saint Paul Lez Durance (France)

    2013-03-15

    Activation energies of U and C atoms self-diffusion in UC, as well as activation energies of hetero-diffusion of fission products (FPs) are investigated by first-principles calculations. According to a previous study which showed a likely U site occupation was favoured for all the FPs, their diffusion is restricted to the uranium sublattice of UC in the present study. In this framework, long-range displacements are only possible through a concerted mechanism with a surrounding uranium vacancy. Using the apparent formation energies of the uranium vacancy defect calculated in our previous study and the classical approach used in UO{sub 2} by Andersson et al., the activation energies of the main fission products in the various stoichiometric domains have been calculated. The results are compared to those obtained with the five frequency model applied to two representative fission products, Xe and Zr. Interestingly, despite strong differences of formalism, both models provided similar activation energies.

  3. Investigation of the diffusion of atomic fission products in UC by density functional calculations

    Science.gov (United States)

    Bévillon, Émile; Ducher, Roland; Barrachin, Marc; Dubourg, Roland

    2013-03-01

    Activation energies of U and C atoms self-diffusion in UC, as well as activation energies of hetero-diffusion of fission products (FPs) are investigated by first-principles calculations. According to a previous study which showed a likely U site occupation was favoured for all the FPs, their diffusion is restricted to the uranium sublattice of UC in the present study. In this framework, long-range displacements are only possible through a concerted mechanism with a surrounding uranium vacancy. Using the apparent formation energies of the uranium vacancy defect calculated in our previous study and the classical approach used in UO2 by Andersson et al., the activation energies of the main fission products in the various stoichiometric domains have been calculated. The results are compared to those obtained with the five frequency model applied to two representative fission products, Xe and Zr. Interestingly, despite strong differences of formalism, both models provided similar activation energies.

  4. Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

    Science.gov (United States)

    Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

    2005-12-01

    Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

  5. Methodology for fission product release calculations during an ACR-1000 end-fitting failure event

    International Nuclear Information System (INIS)

    The ACR-1000® reactor enhances and retains the proven features of the CANDU® design such as the concept of the horizontal fuel channel core. At each end of a fuel channel, there is an end-fitting incorporating a feeder connection through which pressurized coolant enters and leaves the fuel channel, where 12 fuel bundles are inserted. The safety analysis cases include postulated end-fitting failure events to assess the fission product releases from all fuel bundles which would be ejected out of the channel and oxidized in the air-steam environment under decay power. This paper presents the methodology used in assessing the fuel behaviour and the fission product releases during a postulated end-fitting failure in an ACR-1000 reactor. After the end-fitting failure, the 12 fuel bundles are ejected out of the channel and drop onto the fuelling machine vault floor. The fuel bundles are likely heavily damaged by impact and would break into small clusters of elements or fragments. To calculate the fission product releases from an individual fragment, the transient fuel temperature is numerically solved by differential heat equations; the air oxidation model is chosen for the event accordingly; and the fission product inventory and releases are estimated by computer codes ORIGEN-S, CATHENA, ELESTRES and SOURCE-IST. Finally, the total fission product releases from all fragments into containment are calculated. This methodology has been developed for ACR-1000 safety analysis, which is also applicable to CANDU. With the new methodology, the transient releases from up to 150 fission products can be estimated as detail as in fragment. In this paper, a sample calculation is also provided to show the application of the methodology in ACR-1000 safety analysis for end-fitting failure. (author)

  6. Diffusion of fission products and radiation damage in SiC

    Science.gov (United States)

    Malherbe, Johan B.

    2013-11-01

    A major problem with most of the present nuclear reactors is their safety in terms of the release of radioactivity into the environment during accidents. In some of the future nuclear reactor designs, i.e. Generation IV reactors, the fuel is in the form of coated spherical particles, i.e. TRISO (acronym for triple coated isotropic) particles. The main function of these coating layers is to act as diffusion barriers for radioactive fission products, thereby keeping these fission products within the fuel particles, even under accident conditions. The most important coating layer is composed of polycrystalline 3C-SiC. This paper reviews the diffusion of the important fission products (silver, caesium, iodine and strontium) in SiC. Because radiation damage can induce and enhance diffusion, the paper also briefly reviews damage created by energetic neutrons and ions at elevated temperatures, i.e. the temperatures at which the modern reactors will operate, and the annealing of the damage. The interaction between SiC and some fission products (such as Pd and I) is also briefly discussed. As shown, one of the key advantages of SiC is its radiation hardness at elevated temperatures, i.e. SiC is not amorphized by neutrons or bombardment at substrate temperatures above 350 °C. Based on the diffusion coefficients of the fission products considered, the review shows that at the normal operating temperatures of these new reactors (i.e. less than 950 °C) the SiC coating layer is a good diffusion barrier for these fission products. However, at higher temperatures the design of the coated particles needs to be adapted, possibly by adding a thin layer of ZrC.

  7. Integral data testing of ENDF/B fission-product data and comparisons of ENDF/B with other fission product data files

    Energy Technology Data Exchange (ETDEWEB)

    LaBauve, R.J.; England, T.R.; George, D.C.

    1981-11-01

    Three experiments (one from Oak Ridge and two from Los Alamos), in which samples of /sup 235/U and /sup 238/Pu were irradiated with thermal neutrons and either the total, gamma-ray, or gamma- and beta-ray fission product decay-energies were measured as functions of cooling time, were selected for comparisons with calculations made using four different fission product data files. The data files used were (1) the ENDF/B-IV fission product file, (2) the ENDF/B-V fission product file, (3) a file derived by substituting decay energies from JNDC into the ENDF/B-V file, and (4) a file derived by substituting decay-energies and spectra from the UK data file into the ENDF/B-V file. Direct summation calculations and spectral comparisons of the experiments were made using these data files as input, and both types of calculational analyses yielded the same results; namely, all data files are deficient, but the JNDC-ENDF/B-V results for the gamma- and beta-ray total decay-energy agree best with experiments. In addition, spectral comparisons with experiment generally indicate that calculated gamma-ray decay-energies are relatively high for early cooling times and small gamma-ray energies; they are low for early cooling times and large gamma-ray energies. The opposite is somewhat the case for the beta-ray decay energies; that is, the calculations are generally low for small beta-ray energies and high for large energies.

  8. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    International Nuclear Information System (INIS)

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 20000C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

  9. Speciation of fission products in contaminated estuarine sediments by chemical elution techniques

    International Nuclear Information System (INIS)

    This paper describes the use of elution ion-exchange techniques using various ionic and complexing agents in order to elucidate the species of fission products sorbed onto contaminated estuarine sediment. The work concentrates on the fission products Cs-137, Ru-106, Zr-95, Nb-95 and Ce-144. The indications were that caesium was held mainly on inaccessible ion exchange sites; ruthenium appeared to be partially absorbed and partially held on anionic exchange sites; zirconium and niobium were sorbed chemically or physically in the form of complex hydrous oxides; cermium appeared to be in an ionic and easily complexible form on surface sites of the sediment

  10. Status of pseudo-fission-product cross-sections for fast reactors

    International Nuclear Information System (INIS)

    Within the framework of the Subgroup 17 (SG17) benchmark organized by a Working Party of the Nuclear Science Committee of the Nuclear Energy Agency (FR), a comparison of lumped or pseudo-fission-product cross-sections for fast reactors has been made. Several parameters have been compared: the one- group cross-sections and reactivity worths of the lumped nuclide for several partial absorption and scattering cross-sections, and the one-group cross sections of individual fission products. Graphs of the multi-group cross-sections and those of capture cross-sections for 27 nuclides have also been compared. (R.P.)

  11. Preparation of lumped fission product (FP) cross sections for a multigroup library

    International Nuclear Information System (INIS)

    A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author)

  12. Fission Product Transport Models Adopted in REFPAC Code for LOCA Conditions in PWR and WWER NPPS

    International Nuclear Information System (INIS)

    The report presents assumptions and physical models used for calculations of fission product releases from nuclear reactors, their behavior inside the containment and leakages to the environment after large break loss of coolant accident LB LOCA. They are the basis of code REFPAC (RElease of Fission Products under Accident Conditions), designed primarily to represent significant physical processes occurring after LB LOCA. The code describes these processes using three different models. Model 1 corresponds to established US and Russian practice, Model 2 includes all conservative assumptions that are in agreement with the actual state-of-the-art, and Model 3 incorporates formulae and parameter values actually used in EU practice. (author)

  13. The behaviour of selected fission products and actinides on UTEVA® resin

    International Nuclear Information System (INIS)

    The behaviour of selected fission product elements and actinides on UTEVA® resin in HCl and HNO3 media was determined by loading a mixed solution of Sr, Y, Zr, Mo, Ag, Cd, Cs, Ba, Ce, Eu, Tb, U, Np and Pu on to UTEVA® resin. The columns were eluted with decreasing concentrations of each acid. This investigation used stable elemental standards for the fission product elements and radioactive tracers for the actinide elements. The eluted fractions were analysed using ICP-OES and ICP-MS to determine the recovery of the elements across the fractions. A comparison using valency adjustment for the separation of Pu and Np is also reported. (author)

  14. Study on the Separation of 132Te From Fission Products by Precipitation

    Institute of Scientific and Technical Information of China (English)

    MAOGuo-shu; ZHANGSheng-dong; CUIAn-zhi; GUOJing-ru; YANGLei; DINGYou-qian

    2003-01-01

    In order to measure precisely the decay data of 132I, radiochemical purity 132I is directly or indirectly separated and prepared from fission products of 235U. The source of 132I which is separated directly has some isotope nuclides such as 131I, 133I, 135I etc. that interfere measurement of 132I by HPGe γ detector. The method of 132I separated indirectly has two steps: 1) Its mother nuclide 132Te is separated from fission products; 2) 132I is separated from 132Te. By this method, the interference of other isotope nuclides of 132I is greatly decreased and even eliminated.

  15. Experimental Determination of the Antineutrino Spectrum of the Fission Products of $^{238}$U

    CERN Document Server

    Haag, N; Hofmann, M; Oberauer, L; Potzel, W; Schreckenbach, K; Wagner, F M

    2013-01-01

    An experiment was performed at the scientific neutron source FRM II in Garching to determine the cumulative antineutrino spectrum of the fission products of $^{238}$U. This was achieved by irradiating target foils of natural uranium with a thermal and a fast neutron beam and recording the emitted $\\beta$-spectra with a gamma-suppressing electron-telescope. The obtained $\\beta$-spectrum of the fission products of $^{235}$U was normalized to the data of the magnetic spectrometer BILL of $^{235}$U. This method strongly reduces systematic errors in the $^{238}$U measurement. The $\\beta$-spectrum of $^{238}$U was converted into the corresponding antineutrino spectrum. The final $\\bar\

  16. TRIGA fuel enrichment verification based on the measurement of short-lived fission products

    Energy Technology Data Exchange (ETDEWEB)

    Peir, J.-J.; Liu, C.-C. [Nuclear Science and Technology Development Center, National Tsing Hua University, Hsinchu, Taiwan (China); Wang, T.-K. [Department of Engineering and System Science, National Tsing Hua University, Hsinchu, Taiwan (China)

    1999-06-01

    A method is developed to verify the {sup 235}U content of TRIGA fresh fuel using gamma-ray spectrometry of the short-lived fission products {sup 97}Zr/{sup 97}Nb, {sup 132}I and {sup 140}La. The short-lived fission-product activities can be established by irradiating the fuel in a nuclear reactor. Based on the measured activities, the {sup 235}U content can be deduced by iterative calculations. The aim of this work is to establish a calibration method for estimating the burnup values of the rod-type spent fuels without the need for detailed data on fuel irradiation history.

  17. TRIGA fuel enrichment verification based on the measurement of short-lived fission products

    International Nuclear Information System (INIS)

    A method is developed to verify the 235U content of TRIGA fresh fuel using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I and 140La. The short-lived fission-product activities can be established by irradiating the fuel in a nuclear reactor. Based on the measured activities, the 235U content can be deduced by iterative calculations. The aim of this work is to establish a calibration method for estimating the burnup values of the rod-type spent fuels without the need for detailed data on fuel irradiation history

  18. Most probable charge of fission products in proton-induced fission of sup 2 sup 3 sup 8 U and sup 2 sup 3 sup 2 Th

    CERN Document Server

    Kaji, D; Kudo, H; Fujita, M; Shinozuka, T; Fujioka, M

    2002-01-01

    The charge distributions of fission products in proton-induced fission of sup 2 sup 3 sup 8 U and sup 2 sup 3 sup 2 Th were measured in a wide mass range. The most probable charges lay on the proton-rich side in the light fragment region and on the proton-deficient side in the heavy one compared with the unchanged charge distribution hypothesis. This result implies that the charge polarization occurs in the fission process. The charge polarization was examined with respect to the ground-state Q values. The estimations by the Q values fairly well reproduced the experimental most probable charges. These results suggest that the fission path to the most favorable charge division may go through the most energetically favorable path at scission point. (author)

  19. Fission product release from ZrC-coated fuel particles during postirradiation heating at 1600 C

    International Nuclear Information System (INIS)

    Release behavior of fission products from ZrC-coated UO2 particles was studied by a postirradiation heating test at 1600 C (1873 K) for 4500 h and subsequent postheating examinations. The fission gas release monitoring and the postheating examinations revealed that no pressure vessel failure occurred in the test. Ceramographic observations showed no palladium attack and thermal degradation of ZrC. Fission products of 137Cs, 134Cs, 106Ru, 144Ce, 154Eu and 155Eu were released from the coated particles through the coating layers during the postirradiation heating. Diffusion coefficients of 137Cs and 106Ru in the ZrC coating layer were evaluated from the release curves based on a diffusion model. 137Cs retentiveness of the ZrC coating layer was much better than that of the SiC coating layer. ((orig.))

  20. Concentration-triggered fission product release from zirconia: consequences for nuclear safety

    Science.gov (United States)

    Gentils, A.; Thomé, L.; Jagielski, J.; Garrido, F.

    2002-02-01

    Crystalline oxide ceramics, more particularly zirconia and spinel, are promising matrices for plutonium and minor actinide transmutation. An important issue concerning these materials is the investigation of their ability to confine radiotoxic elements resulting from the fission of actinides. This letter reports the study of the release, upon annealing or irradiation at high temperature, of one of the most toxic fission product (Cs) in zirconia. The foreign species are introduced by ion implantation and the release is studied by Rutherford backscattering experiments. The results emphasize the decisive influence of the fission product concentration on the release properties. The Cs mobility in zirconia is strongly increased when the impurity concentration exceeds a threshold of the order of a few atomic per cent. Irradiation with medium-energy heavy ions is shown to enhance Cs outdiffusion with respect to annealing at the same temperature.

  1. Volatile fission product distributions in LWR fuel rods

    International Nuclear Information System (INIS)

    Results from this study are a part of spent fuel characterizations being conducted by the Materials Characterization Center (MCC) project at Pacific Northwest Laboratory on a variety of spent uranium oxide fuels designated as Approved Testing Materials (ATMs). These ATMs have a variety of burnup levels, fission gas releases, and include fuel from both pressurized water and boiling water reactor designs. The purpose of this work is to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) repository programs and potentially other programs. Details of these characterization studies for some of the ATMs are available. 7 refs., 8 figs., 1 tab

  2. HYPERFUSE: a hypervelocity inertial confinement system for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from a LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137Cs, 90Sr, 129I, 99Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,α), (n,γ), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R = 1.0 to 3.0) requirements. These studies also indicate that masses on the order of 1.0 g at densities of rho greater than or equal to 500.0 g/cm3 are required for a practical fusion-based fission product transmutation system

  3. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    Energy Technology Data Exchange (ETDEWEB)

    Asner, David M.; Burns, Kimberly A.; Campbell, Luke W.; Greenfield, Bryce A.; Kos, Marek S.; Orrell, John L.; Schram, Malachi; VanDevender, Brent A.; Wood, Lynn S.; Wootan, David W.

    2015-03-01

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

  4. Indonesia's current status for conversion of Mo-99 production to LEU fission

    International Nuclear Information System (INIS)

    Indonesia has a conversion program from HEU to LEU for producing Mo-99 from LEU foil target. Limited and restricted row material of HEU is the basic reasons to have conversion program in producing Mo-99 from LEU fission. The substitution of low-enriched uranium (LEU) metal foils for the HEU UO2 used in current target designs will be applied for production of Mo-99 Indonesia commercially. Batan has a joint research project with ANL to develop LEU-metal-foil target fabrication. Presented here is the current status of the experiment of foil fabrication using depleted uranium metal. The program of HEU will be reviewed as a background of the conversion program, and discussion will be focused on experiment results of foil target fabrication. The results show that foil targets resulted from the rolling has good characteristics. The surface of the foil is very smooth, and the grain has random orientation after heat treatment and quenching. Foil targets are fabricated from Ni foil-wrapped-DU foil inserted in between two concentric aluminum tubes, and the ends of the tubes are welded. Testing is conducted to assure that there is no leakage in the welded tubes. (author)

  5. Assessment of failed fuel and tramp uranium based on the activity of fission products in the primary circuit

    International Nuclear Information System (INIS)

    We have proposed a model for the nuclear fuel state of the operating power reactor from the physical characteristics of nuclear fission products which have been produced by nuclear reaction between neutron and uranium-235. The model equation for nuclear fission products release has been split into size independent steps: 1) calculation of the fission products generation inside the solid nuclear fuel, 2) release from the fuel to the fuel surface in three different ways, 3) release between the fuel surface and gap, 4) release from the defective nuclear fuel to the reactor coolant, 5) mass balance in the coolant taking into account the purification rate, 6) separation of fission products sources with two parts, i.e. fuel and tramp uranium. We have solved the equation of the model, calculated the activity of fission products released from the defected fuel to coolant and put the experimental activity data of the nuclear fission products in the primary coolant to determine the number of defective fuel and amount of tramp uranium by using the computer. The measurement and analysis of nuclear fission products in the primary coolant of nuclear power reactors have been carried out at the pressurized water reactor, Korea Nuclear Unit 2, 7 and 8. We have used the iodine isotopes among the nuclides of fission products. The analysis results have been well agreed with the results of diffusion model and of kinetics model. (author). 7 refs, 2 figs, 8 tabs

  6. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1976. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-24

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and, where appropriate, the data are presented in tables, graphs, and photographs.

  7. Relative yields of U-235 fission products measured in a high level radioactive sludge at Savannah River Site

    International Nuclear Information System (INIS)

    This paper presents measurements of the concentrations of 42 of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at Savannah River Site. The 42 fision products make up 98% of the waste sludge. We used inductively coupled plasma-mass spectroscopy for the analysis. The relative yields for most of the fission products are in complete agreement with the known relative yields for the beta decay chains of the two asymmetric branches of the slow neutron fission of U-235. Disagreements can be reconciled based on the chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses. This paper presents measurements of the concentrations of 42 (98%) of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at the Savannah River Site. We analyzed the sludge with inductively coupled plasma-mass spectroscopy. The relative yields for most of the fission products agree completely with the known relative vields for the beta decay chains of the two asymmetric: branches of the slow neutron fission of U-235. The chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses explain the differences in the measured and calculated results

  8. Method of reprocessing of irradiated nuclear fission products of the uranium, plutonium and thorium group

    International Nuclear Information System (INIS)

    A solvent extraction is used to separate irradiated nuclear fission materials of the group uranium, plutonium, thorium from radioactive fission products which are present together in an aqueous solution. An improvement on the known mehod is proposed in which a carboxylic nitrile, carboxylic ester, carboxylic amide, or a mixture of these substances is added to the organic phase which is mixed with a non-polar diluting agent as a polar modificator, where the modificators are derived from mono- or polycarboxylic acids or also from substituted carboxylic acids. Amyl acetate, N-N dimethyl caprylic acid amide, and adiponitrile are particularly suitable. (UW/LH)

  9. Measurement of the hydrogen yield in the radiolysis of water by dissolved fission products

    International Nuclear Information System (INIS)

    Hydrogen from the radiolysis of water by dissolved fission products is stripped from the solution and collected by bubbling CO2 through the solution. Quantitative measurements of the G value for hydrogen show that the yield is essentially the same as would be obtained by external gamma radiolysis of nonradioactive solutions of the same chemical composition. The hydrogen yield can be enhanced by addition of a hydrogen-atom donor, such as formic acid, to the solution. The yield of hydrogen from fission-waste solutions is discussed with respect to the question of whether it represents a significant energy source

  10. Simulation of neutron rich nuclei production through 239U fission at intermediates energies

    International Nuclear Information System (INIS)

    The theoretical part and some results obtained from a model realised for fission processes in wide range of mass-asymmetries are presented. The fission barriers are computed in a tridimensional configuration space using the Yukawa - plus - exponential macroscopic energies corrected within the Strutinsky procedure. It is assumed that channel probabilities are proportional with Gamow penetrabilities. The model is applied for the disintegration of the 239U in order to determine the relative yields for the production of neutron rich nuclei at diverse intermediate energies. (author)

  11. Behaviour of solid fission products in the HTGR coated fuel particles

    International Nuclear Information System (INIS)

    Results of profile measurements for volume concentrations of 134,137Cs, 144Ce, 155Eu, 106Ru and fissionable material in the HTGR coated fuel particles which have been subjected to standard tests in the temperature range of 1273-2133 K and at burnup up to 17% fima are presented. Values of the effective coefficients of cesium diffusion in kern and protective coating of fuel particles which were subjected to standard in-pile tests in spherical fuel elements at the temperature of 1273 K and the burnup up to 15% fima as well as the value of relative release of solid fission products from the samples studied are given

  12. Review of the book: Vasilenko, I.Ya. Toxicology of nuclear fission products

    International Nuclear Information System (INIS)

    Review on monograph of Vasilenko, I.Ya. Toxicology of nuclear fission (Moscow, Medicine, 1999) is presented. Data of longevity full-scale investigations during nuclear explosions on the Semipalatinsk test site are given. Classified, complex investigations into the effect of nuclear fission products mixtures on different kinds of laboratory animals are described, transfer of radiobiological researches to organism of man is scientific valid. The most complicate radiobiological problem of low dose is analyzed. The being investigated monograph contains unique scientific information and makes a heavy contribution in radiobiology

  13. Report on the Behavior of Fission Products in the Co-decontamination Process

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Leigh Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riddle, Catherine Lynn [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-30

    This document was prepared to meet FCT level 3 milestone M3FT-15IN0302042, “Generate Zr, Ru, Mo and Tc data for the Co-decontamination Process.” This work was carried out under the auspices of the Lab-Scale Testing of Reference Processes FCT work package. This document reports preliminary work in identifying the behavior of important fission products in a Co-decontamination flowsheet. Current results show that Tc, in the presence of Zr alone, does not behave as the Argonne Model for Universal Solvent Extraction (AMUSE) code would predict. The Tc distribution is reproducibly lower than predicted, with Zr distributions remaining close to the AMUSE code prediction. In addition, it appears there may be an intricate relationship between multiple fission product metals, in different combinations, that will have a direct impact on U, Tc and other important fission products such as Zr, Mo, and Rh. More extensive testing is required to adequately predict flowsheet behavior for these variances within the fission products.

  14. Nuclear transmutation strategies for management of long-lived fission products

    Indian Academy of Sciences (India)

    S Kailas; M Hemalatha; A Saxena

    2015-09-01

    Management of long-lived nuclear waste produced in a reactor is essential for long-term sustenance of nuclear energy programme. A number of strategies are being explored for the effective transmutation of long-lived nuclear waste in general, and long-lived fission products (LLFP), in particular. Some of the options available for the transmutation of LLFP are discussed.

  15. Thermochemical data for reactor materials and fission products: The ECN database

    International Nuclear Information System (INIS)

    The activities of the authors regarding the compilation of a database of thermochemical properties for reactor materials and fission products is reviewed. The evaluation procedures and techniques are outlined and examples are given. In addition, examples of the use of thermochemical data for the application in the field of Nuclear Technology are given. (orig.)

  16. Fission product partitioning in aerosol release from simulated spent nuclear fuel

    NARCIS (Netherlands)

    Di Lemma, F.G.; Colle, J.Y.; Rasmussen, G.; Konings, R.J.M.

    2015-01-01

    Aerosols created by the vaporization of simulated spent nuclear fuel (simfuel) were produced by laser heating techniques and characterised by a wide range of post-analyses. In particular attention has been focused on determining the fission product behaviour in the aerosols, in order to improve the

  17. A study of highly concentrated fission product salt loading into zeolite-A

    International Nuclear Information System (INIS)

    A study investigating the loading of highly contaminated electrorefiner salt into zeolite-4A is currently underway. The objective of the study is to optimize the absorption process in order to maximize fission product sorption into zeolite, which should result in reduction of waste associated with the pyrochemical processing of spent nuclear fuel. The study is based on both experimental and theoretical investigations to develop a fundamental understanding of the transport of LiCl-KCl and fission product chlorides in zeolite-A. A diffusion-limited sorption rate model has been formulated, while data have been collected using molten salt-zeolite contacting experiments. Solid-state salt-zeolite contacts have also been performed to probe single salt sorption characteristics in the absence of LiCl-KCl. Experimental data suggest that the rate of fission product sorption into the zeolite can be increased by pre-loading the zeolite with LiCl-KCl. Additionally, experiments involving variable zeolite particle size were performed in an effort to further support the theory that the salt sorption is diffusion-limited. X-ray fluorescence imaging of the cross section of a zeolite pellet after solid-state salt-zeolite contacting was performed to examine salt distribution through the zeolite pellet. The findings will be used to design practical processes that can be used for absorbing high fission product content salt into zeolite-A. (author)

  18. Method of Fission Product Beta Spectra Measurements for Predicting Reactor Anti-neutrino Emission

    CERN Document Server

    Asner, D M; Campbell, L W; Greenfield, B; Kos, M S; Orrell, J L; Schram, M; VanDevender, B; Wood, 1 L S; Wootan, D W

    2014-01-01

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron rich fission products that subsequently beta decay and emit electron anti-neutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to current precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent re-considerations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable i...

  19. Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-09-01

    This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive

  20. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  1. Fission product release in conditions of a spent fuel pool severe accident

    International Nuclear Information System (INIS)

    Full text: Depending on the residence time, fuel burnup, and fuel rack configuration, there may be sufficient decay heat for the fuel clad to heat up, swell, and burst in case of a loss of pool water. Initiating event categories can be: loss of offsite power from events initiated by severe weather, internal fire, loss of pool cooling, loss of coolant inventory, seismic event, aircraft impact, tornado, missile attack. The breach in the clad releases the radioactive gases present in the gap between the fuel and clad, what is called 'gap release'. If the fuel continues to heat up, the zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or zirconium and steam is exothermic. The energy released from the reaction, combined with the fuel's decay energy, can cause the reaction to become self-sustaining and ignite the zirconium. The increase in heat from the oxidation reaction can also raise the temperature in adjacent fuel assemblies and propagate the oxidation reaction. Simultaneously, the sintered UO2 pellets resulting from pins destroying are oxidized. Due to the self-disintegration of pellets by oxidation, fission gases and low volatile fission products are released. The release rate, the chemical nature and the amount of fission products depend on powder granulation distribution and environmental conditions. The zirconium burning and pellets self-disintegration will result in a significant release of spent fuel fission products that will be dispersed from the reactor site. (author)

  2. Fission rates measured using high-energy gamma-rays from short half-life fission products in fresh and spent nuclear fuel

    International Nuclear Information System (INIS)

    In recent years, higher discharge burn-ups and initial fuel enrichments have led to more and more heterogeneous core configurations in light water reactors (LWRs), especially at the beginning of cycle when fresh fuel assemblies are loaded next to highly burnt ones. As this trend is expected to continue in the future, the Paul Scherrer Institute has, in collaboration with the Swiss Association of Nuclear Utilities, swissnuclear, launched the experimental programme LIFE(at)PROTEUS. The LIFE(at)PROTEUS programme aims to better characterise interfaces between burnt and fresh UO2 fuel assemblies in modern LWRs. Thereby, a novel experimental database is to be made available for enabling the validation of neutronics calculations of strongly heterogeneous LWR core configurations. During the programme, mixed fresh and highly burnt UO2 fuel lattices will be investigated in the zero-power research reactor PROTEUS. One of the main types of investigations will be to irradiate the fuel in PROTEUS and measure the resulting fission rate distributions across the interface between fresh and burnt fuel zones. The measurement of fission rates in burnt fuel re-irradiated in a zero-power reactor requires, however, the development of new experimental techniques which are able to discriminate against the high intrinsic activity of the fuel. The principal goal of the present research work has been to develop such a new measurement technique. The selected approach is based on the detection of high-energy gamma-ray lines above the intrinsic background (i.e. above 2200 keV), which are emitted by short-lived fission products freshly created in the fuel. The fission products 88Kr, 142La, 138Cs, 84Br, 89Rb, 95Y, 90mRb and 90Rb, with half-lives between 2.6 min and 2.8 h, have been identified as potential candidates. During the present research work, the gamma-ray activity of short-lived fission products has, for the first time, been measured and quantitatively evaluated for re-irradiated burnt UO

  3. Background for Terrestrial Antineutrino Investigations: Radionuclide Distribution, Georeactor Fission Events, and Boundary Conditions on Fission Power Production

    OpenAIRE

    Herndon, J. Marvin; Edgerley, Dennis A.

    2005-01-01

    Estimated masses of fissioning and non-fissioning radioactive elements and their respective distributions within the Earth are presented, based upon the fundamental identity of the components of the interior 82% of the Earth, the endo-Earth, with corresponding components of the Abee enstatite chondrite meteorite. Within limits of existing data, the following generalizations concerning the endo-Earth radionuclides can be made: (1) Most of the K-40 may be expected to exist in combination with o...

  4. HEU and LEU MTR fuel elements as target materials for the production of fission molybdenum

    International Nuclear Information System (INIS)

    The processing of irradiated MTR-fuels for the production of fission nuclides for nuclear medicine presents a significantly increasing task in the field of chemical separation technology of high activity levels. By far the most required product is MO-99, the mother nuclide of Tc-99m which is used in over 90% of the organ function tests in nuclear medicine. Because of the short half life of Mo-99 (66 h) the separation has to be carried out from shortly cooled neutron irradiated U-targets. The needed product purity, the extremely high radiation level, the presence of fission gases like xenon-133 and of volatile toxic isotopes such as iodine-131 and its compounds in kCi-scale require a sophisticated process technology

  5. HYPERFUSE: a hypervelocity inertial confinement system for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137Cs, 90Sr, 129I, 99Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,α), (n,γ), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R=1.0 to 3.0) requirements

  6. Irradiation effects and behaviour of fission products in zirconia and spinel; Effets d'irradiation et comportement des produits de fission dans la zircone et le spinelle

    Energy Technology Data Exchange (ETDEWEB)

    Gentils, A

    2003-10-01

    Crystalline oxides, such as zirconia (ZrO{sub 2}) and spinel (MgAl{sub 2}O{sub 4}), are promising inert matrices for the transmutation of plutonium and minor actinides. This work deals with the study of the physico-chemical properties of these matrices, more specifically their behaviour under irradiation and their capacity to retain fission products. Irradiations at low energy and incorporation of stable analogs of fission products (Cs, I, Xe) into yttria-stabilized zirconia and magnesium-aluminate spinel single crystals were performed by using the ion implanter IRMA (CSNSM-Orsay). Irradiations at high energy were made on several heavy ion accelerators (GANIL-Caen, ISL-Berlin, HIL-Warsaw). The damage induced by irradiation and the release of fission products were monitored by in situ Rutherford Backscattering Spectrometry experiments. Transmission electron microscopy was also used in order to determine the nature of the damage induced by irradiation. The results show that irradiation of ZrO{sub 2} and MgAl{sub 2}O{sub 4} with heavy ions (about hundred keV and about hundred MeV) induces a huge structural damage in crystalline matrices. Total disorder (amorphization) is however never reached in zirconia, contrary to what is observed in the case of spinel. The results also emphasize the essential role played by the concentration of implanted species on their retention capacity. A dramatic release of fission products was observed when the concentration exceeds a threshold of a few atomic percent. Irradiation of implanted samples with medium-energy noble-gas ions leads to an enhancement of the fission product release. The exfoliation of spinel crystals implanted at high concentration of Cs ions is observed after a thermal treatment at high temperature. (author)

  7. 244CmO2/nat.-UO2 hybrid blanket with flat fission power production

    International Nuclear Information System (INIS)

    In the present study, 244CmO2 is mixed with nat.-UO2 for the purpose of power flattening in a hybrid blanket with a reasonably high energy multiplication factor. Also, the temporal variations of the fission power density (FPD) are observed during an 18-month plant operation period. The main conclusion drawn from this work is that it became possible to keep a flat fission power profile (FPP) over a very long plant operation period of 18 months by simply omitting the beryllium multiplier in the blanket and keeping the neutron spectrum fairly unchanged throughout the fission zone. This reduced the efforts for fuel management to a minimum. A further observation focused on only minor variations of the integral neutronic data over longer plant operation periods. Among others, the fission power generation increase is also very modest. This results in an optimum investment for the nonnuclear island. The blanket burns up high-level nuclear waste 244Cm effectively, with efficient electricity production and breeding of a new type of nuclear fuel 245Cm with very superior nuclear properties. Finally, a warning should be issued for the careful international safeguarding of such a hybrid plant due to the extremely high quality of the bred plutonium fuel

  8. FITPULS: a code for obtaining analytic fits to aggregate fission-product decay-energy spectra. [In FORTRAN

    Energy Technology Data Exchange (ETDEWEB)

    LaBauve, R.J.; George, D.C.; England, T.R.

    1980-03-01

    The operation and input to the FITPULS code, recently updated to utilize interactive graphics, are described. The code is designed to retrieve data from a library containing aggregate fine-group spectra (150 energy groups) from fission products, collapse the data to few groups (up to 25), and fit the resulting spectra along the cooling time axis with a linear combination of exponential functions. Also given in this report are useful results for aggregate gamma and beta spectra from the decay of fission products released from /sup 235/U irradiated with a pulse (10/sup -4/ s irradiation time) of thermal neutrons. These fits are given in 22 energy groups that are the first 22 groups of the LASL 25-group decay-energy group structure, and the data are expressed both as MeV per fission second and particles per fission second; these pulse functions are readily folded into finite fission histories. 65 figures, 11 tables.

  9. [Medical equipment product lines in basic pharmacies].

    Science.gov (United States)

    Macesková, B; Lipská, J

    2003-07-01

    Medical appliances dispensed in basic type pharmacies for cash or vouchers for medical or orthopedic appliances require expertise of pharmacists and laboratory assistants concerning the assortment, payment, construction of prices, conditions for prescription, ordering, properties, and functions of individual appliances. Using the method of frequency analysis, the analysis of data from five pharmacies within a period of three months (more than 17,000 records) revealed how individual subgroups of medical appliances and their concrete items are represented in both types of dispensation. The method of the semistructured questionnaire (10 respondents) was used to find what problems are encountered in dispensation, and which medical appliances and their subgroups are the sources of the problems. The respondents regard the contemporary level of knowledge concerning medical appliances gained in pregradual studies as insufficient.

  10. Contribution to decay heat calculation: fission product mean beta and gamma assessment

    International Nuclear Information System (INIS)

    Following a reactor shutdown, after the fission chain process has completely faded out, a significant quantity of energy (around seven per cent of the total power of the reactor) continues to be generated in the core. This is known as residual power or decay heat. The principal source of this energy is due to the radioactive decay of fission products and is at any time equal to the sum of the powers released by these different nuclei (P = Σ = Pi). Each of the powers Pi is the product of three terms: the concentration of the relevant nuclide, its decay constant and its mean decay energy. The evaluation of the first two term is straightforward. On the other hand the evaluation of the mean energies presents some difficulties due to a lack of data in beta and gamma spectra of some fission products. This study intends, after a critical analysis of the current method of evaluation of the mean energies, to propose a new model for this calculation. The new model tested on several well known nuclides, has been proved correct and precise. It has then been applied to approximatively sixty nuclides among the lesser known ones. The results obtained have lead to a better prediction of both beta and gamma ray components of the residual power. Consequently, this new model, which allows to take into account the lack of beta branching ratio corresponding to the highest levels of the product nucleus in the beta decay reaction, can be adopted to replace the current method, for calculation of the mean energies of fission products, especially in the case of the lesser known nuclides

  11. Investigation of crystallization in glasses containing fission products

    International Nuclear Information System (INIS)

    Five potential solidification products for high-level waste (four borosilicate glasses and one celsian glass ceramic) have been investigated in terms of crystallization. In all glasses and in the glass ceramic, crystallization, and recrystallization, respectively, were observed by heating above 7730K, however, at very different periods of time (0.1d greater than or equal to 100d). The noble metals precipitated into various phases. Crystal growth proceeded at the phase boundary glass-noble metal. In all products rare earth phases crystallized. Silicate phases rarely formed. The leach resistance (by the grain titration and Soxhlet tests) decreased after heat treatment in all cases. The changes were found to be within one order of magnitude for all products. 2 figures, 4 tables

  12. Development of industrial-scale fission {sup 99}Mo production process using low enriched uranium target

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Kon; Lee, Jun Sig [Radioisotope Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Beyer, Gerd J. [Grunicke Strasse 15, Leipzig (Germany)

    2016-06-15

    Molybdenum-99 ({sup 99}Mo) is the most important isotope because its daughter isotope, technetium-99m ({sup 99}mTc), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of {sup 99}Mo, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of {sup 99}Mo technology developments. Most of the industrial-scale {sup 99}Mo processes have been based on the fission of {sup 235}U. Recently, important issues have been raised for the conversion of fission {sup 99}Mo targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of {sup 99}Mo yield, caused by a significant reduction of {sup 235}U enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission {sup 99}Mo production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the {sup 99}Mo production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

  13. Exploratory study of fission product yields of neutron-induced fission of 235U , 238U , and 239Pu at 8.9 MeV

    Science.gov (United States)

    Bhatia, C.; Fallin, B. F.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E.; Bredeweg, T. A.; Fowler, M. M.; Moody, W.; Rundberg, R. S.; Rusev, G. Y.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

    2015-06-01

    Using dual-fission chambers each loaded with a thick (200 -400 -mg /c m2) actinide target of 235 ,238U or 239Pu and two thin (˜10 -100 -μ g /c m2) reference foils of the same actinide, the cumulative yields of fission products ranging from 92Sr to 147Nd have been measured at En= 8.9 MeV . The 2H(d ,n ) 3He reaction provided the quasimonoenergetic neutron beam. The experimental setup and methods used to determine the fission product yield (FPY) are described, and results for typically eight high-yield fission products are presented. Our FPYs for 235U(n ,f ) , 238U(n ,f ) , and 239Pu(n ,f ) at 8.9 MeV are compared with the existing data below 8 MeV from Glendenin et al. [Phys. Rev. C 24, 2600 (1981), 10.1103/PhysRevC.24.2600], Nagy et al. [Phys. Rev. C 17, 163 (1978), 10.1103/PhysRevC.17.163], Gindler et al. [Phys. Rev. C 27, 2058 (1983), 10.1103/PhysRevC.27.2058], and those of Mac Innes et al. [Nucl. Data Sheets 112, 3135 (2011), 10.1016/j.nds.2011.11.009] and Laurec et al. [Nucl. Data Sheets 111, 2965 (2010), 10.1016/j.nds.2010.11.004] at 14.5 and 14.7 MeV, respectively. This comparison indicates a negative slope for the energy dependence of most fission product yields obtained from 235U and 239Pu , whereas for 238U the slope issue remains unsettled.

  14. Fission products behaviour in UO2 submitted to nuclear severe accident conditions

    Science.gov (United States)

    Geiger, E.; Bès, R.; Martin, P.; Pontillon, Y.; Solari, P. L.; Salome, M.

    2016-05-01

    The objective of this work was to study the molybdenum chemistry in UO2 based materials, known as SIMFUELS. These materials could be used as an alternative to irradiated nuclear fuels in the study of fission products behaviour during a nuclear severe accident. UO2 samples doped with 12 stable isotopes of fission products were submitted to annealing tests in conditions representative to intermediate steps of severe accidents. Samples were characterized by SEM-EDS and XAS. It was found that Mo chemistry seems to be more complex than what is normally estimated by thermodynamic calculations: XAS spectra indicate the presence of Mo species such as metallic Mo, MoO2, MoO3 and Cs2MoO4.

  15. Radiochemical applications of insoluble sulfate columns. Analytical possibilities in the field of the fission product solutions

    International Nuclear Information System (INIS)

    In this paper we go on with our study of the heterogeneous ion-isotopic exchange in column. At present, we apply it to determine the radiochemical composition of the raw solutions used in the industrial recuperation of the long-lived fission products. The separation of the radioelements contained in these solutions is carried out mainly by making use of small columns, 1-3 cm height, of BaSO4 or SrSO4, under selected experimental conditions. These columns behave like a special type of inorganic exchangers, working by absorption or by ion-isotopic exchange depending on the cases,a nd they provide the means for the selective separation of several important fission products employing very small volumes of fixing and eluting solutions. (Author) 11 refs

  16. RAFT - A computer model for formation and transport of fission product aerosols in LWR primary systems

    International Nuclear Information System (INIS)

    A computer model RAFT (Reactor Aerosol Formation and Transport) has been developed to predict the size distribution and composition of the particles (aerosols) formed from condensation of the fission-product and control rod material vapors released in LWR accidents. The underlying theory of RAFT considers the processes of homogeneous and heterogeneous nucleation, aerosol agglomeration, and aerosol and vapor deposition, in conjunction with the equilibrium chemistry of the Cs-I-Te-O-H-Ag-In-Cd-inert gas system. Calculations using RAFT show that under most accident conditions, the particle size spectrum is determined primarily by the competition between the homogeneous and heterogeneous nucleation mechanisms rather than the agglomeration mechanism, and that direct vapor deposition on structural surfaces is an important mechanism for the scavenging of fission product vapours

  17. RAFT: a computer model for formation and transport of fission product aerosols in LWR primary systems

    International Nuclear Information System (INIS)

    A computer model RAFT (Reactor Aerosol Formation and Transport) has been developed to predict the size distribution and composition of the particles (aerosols) formed from condensation of the fission-product and control rod material vapors released in LWR accidents. The underlying theory of RAFT considers the processes of homogeneous and heterogeneous nucleation, aerosol agglomeration, and aerosol and vapor deposition, in conjunction with the equilibrium chemistry of the Cs-I-Te-O-H-Ag-In-Cd-inert gas system. Calculations using RAFT show that under most accident conditions, the particle size spectrum is determined primarily by the competition between the homogeneous and heterogeneous nucleation mechanisms rather than the agglomeration mechanism, and that direct vapor deposition on structural surfaces is an important mechanism for the scavenging of fission product vapors

  18. Fission product Pd-SiC interaction in irradiated coated particle fuels

    International Nuclear Information System (INIS)

    Silicon carbide is the main barrier to fission product release from coated particle fuels. Consequently, degradation of the SiC must be minimized. Electron microprobe analysis has identified that palladium causes corrosion of the SiC in irradiated coated particles. Further ceramographic and electron microprobe examinations on irradiated particles with kernels ranging in composition from UO2 to UC2, including PuO/sub 2 -x/ and mixed (Th, Pu) oxides, and in enrichment from 0.7 to 93.0% 235U revealed that temperature is the major factor affecting the penetration rate of SiC by Pd. The effects of kernel composition, Pd concentration, other fission products, and SiC properties are secondary

  19. Measurement of airborne fission products in Chapel Hill, NC, USA from the Fukushima I reactor accident

    CERN Document Server

    MacMullin, S; Green, M P; Henning, R; Holmes, R; Vorren, K; Wilkerson, J F

    2011-01-01

    We present measurements of airborne fission products in Chapel Hill, NC, USA, from 62 days following the March 11, 2011, accident at the Fukushima I Nuclear Power Plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products I-131 and Cs-137 were measured with maximum activities of 4.2 +/- 0.6 mBq/m^2 and 0.42 +/- 0.07 mBq/m^2 respectively. Additional activity from I-131, I-132, Cs-134, Cs-136, Cs-137 and Te-132 were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

  20. Swelling due to fission products and additives dissolved within the uranium dioxide lattice

    Energy Technology Data Exchange (ETDEWEB)

    Middleburgh, S.C., E-mail: simon.middleburgh@imperial.ac.uk [Centre for Nuclear Engineering, Department of Materials, Imperial College London, London SW7 2AZ (United Kingdom); Department of Materials and Fuel Rod Design, Westinghouse Electric Sweden AB, Vaesteras (Sweden); Grimes, R.W.; Desai, K.H. [Centre for Nuclear Engineering, Department of Materials, Imperial College London, London SW7 2AZ (United Kingdom); Blair, P.R.; Hallstadius, L.; Backman, K. [Department of Materials and Fuel Rod Design, Westinghouse Electric Sweden AB, Vaesteras (Sweden); Van Uffelen, P. [European Commission, Joint Research Centre, D-76344 Eggenstein-Leopoldeshafen (Germany)

    2012-08-15

    Simulations using empirical inter-atomic potentials have been used to predict the change in volume of the uranium dioxide lattice due to the accommodation of soluble fuel additives and fission products. The incorporation of divalent, trivalent and tetravalent cations are considered. The change in accommodation mechanism for aliovalent cations between UO{sub 2} and UO{sub 2+x} gives rise to markedly different defect volumes. Experimental data is in good agreement with the predictions made in this work, particularly swelling as a function of dopant concentration under different conditions. The predicted defect volumes have been combined to predict the change in lattice volume with burnup (fission product inventory) due to incorporation of these soluble species, which agrees well with swelling data from irradiated fuel.

  1. UKFPDD-2: a revised fission product decay data file in ENDF/B-IV format

    International Nuclear Information System (INIS)

    The UKFPDD-1 fission product decay data file has been revised, the main objective being to reduce, as much as possible, its content of theoretical data. For this purpose, a system of computer codes has been developed for the extraction of decay data from the Evaluated Nuclear Structure Data File (ENSDF) and conversion to the input format required by the processing code COGEND. In addition, some very recent data were transcribed directly from the journals into COGEND input format. The resulting data file, UKFPDD-2, is described. It consists of over 42,500 card images coded in ENDF/B-IV format. Spectral data are included for 390 of the 736 radioactive nuclides present in the file while theoretical half-life estimates are given for less than 200 fission products. By virtue of its greatly reduced theoretical content, UKFPDD-2 should replace UKFPDD-1 in all of its applications. (U.K.)

  2. Sodium aerosol release rate and nonvolatile fission product retention factor during a sodium-concrete reaction

    International Nuclear Information System (INIS)

    This paper reports on a series of tests conducted to study the mechanical release behavior of sodium aerosols containing nonvolatile fission products during a sodium-concrete reaction in which release behavior due to hydrodynamic breakup of the hydrogen bubble is predominant at the sodium pool surface. In the tests, nonradioactive materials, namely, strontium oxide, europium oxide, and ruthenium particles, whose sizes range from a few microns to several tens of microns, are used as nonvolatile fission product stimulants. The following results are obtained: The sodium aerosol release rate during the sodium-concrete reaction is larger than that of natural evaporation. The difference, however, becomes smaller with increasing sodium temperature: nearly ten times smaller at 400 degrees C and three times at 700 degrees C. The retention factors for the nonvolatile materials in the sodium pool increase to the range of 0.5 to 104 with an increase in the sodium temperature from 400 to 700 degrees C

  3. Development of assessment technology for hydrogen burn and fission product behavior in containment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. B.; Kim, J. T.; Ha, K. S.; Hong, S. W.; Song, Y. M.; Park, J. H.; Cho, Y. R.; Kang, H. S.

    2012-04-15

    Analysis tools for hydrogen burn was established to resolve the hydrogen issues in containment. To validate CFX commercial CFD(computational fluid dynamics) code, the hydrogen combustion experiments such as FLAME and ENACEFF for reactor containment were analyzed. And OpenFOAM hydrogen combustion code was developed and validated. Experiments for the flame propagation characteristics in IRWST and the run-up-distance for DDT(Deflagration to detonation transition) were performed and analytical model was evaluated to evaluation of the performance of hydrogen mitigation system, that is, PAR(Passive auto-catalistic re-combiner) To improvement of the fission product modelling in containment, separate analysis module for Iodine behavior and its application tool of K-IODIP (Korea IODIne Package) were developed. PHEBUS FPT-3 analysis was performed to validate MELCOR code. And also the characteristics of fission product behaviors in Future Reactors(GEN-IV) were compared.

  4. METHOD OF SEPARATING URANIUM, PLUTONIUM AND FISSION PRODUCTS BY BROMINATION AND DISTILLATION

    Science.gov (United States)

    Jaffey, A.H.; Seaborg, G.T.

    1958-12-23

    The method for separation of plutonium from uranium and radioactive fission products obtained by neutron irradiation of uranlum consists of reacting the lrradiated material with either bromine, hydrogen bromide, alumlnum bromide, or sulfur and bromine at an elevated temperature to form the bromides of all the elements, then recovering substantlally pure plutonium bromide by dlstillatlon in combinatlon with selective condensatlon at prescribed temperature and pressure.

  5. Measurement of reaction cross sections of fission products induced by DT neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Daisuke; Murata, Isao; Takahashi, Akito [Osaka Univ., Suita (Japan)

    1998-03-01

    With the view of future application of fusion reactor to incineration of fission products, we have measured the {sup 129}I(n,2n){sup 128}I reaction cross section by DT neutrons with the activation method. The measured cross section was compared with the evaluated nuclear data of JENDL-3.2. From the result, it was confirmed that the evaluation overestimated the cross section by about 20-40%. (author)

  6. Separation Sr and Ba From Fission Products Using Sr-spec Resin Column

    Institute of Scientific and Technical Information of China (English)

    WANG; Xiu-feng

    2013-01-01

    Sr and Ba are the IIA group elements,with the same outer electronic structure,and the very similar properties,so separation of the two elements becomes very difficult.The traditional separation method of Sr and Ba in fission products is repeated precipitation of BaCl2·H2O in hydrochloric acid-ether medium.Four times repeated precipitation steps are needed to ensure the decontamination factor of Sr to be better

  7. Review of fission product retention experiment results and application to the LWR design

    International Nuclear Information System (INIS)

    This paper examines the available literature on pipe retention of aerosols to determine the feasiblity of taking credit for attenuation of radioactive release from the plant through pipes when calculating off-site dose rates for plant licensing and emergency planning considerations. The results show that deposition in pipes can be significant. Experimental work on aerosol plugging shows that this phenomenon may provide the dominant mechanism for fission product retention

  8. Revaporisation of fission product deposits in the primary circuit and its impact on accident source term

    OpenAIRE

    BOTTOMLEY Paul; KNEBEL KEVIN; VAN WINCKEL Stefaan; HASTE Tim; Souvi, Sidi,; AUVINEN Ari; KALILAINEN J.; KÄRKELÄ Teemu

    2014-01-01

    Chemical revaporisation or physical resuspension of fission product deposits from the primary circuit is now recognised to be a major source term in the late phase of severe fuel degradation in a severe nuclear accident. These results come from tests carried out under different experimental projects in the European Commission (EC) Framework Programmes. These include the revaporisation tests carried out at the Transuranium Institute (ITU), Karlsruhe under the Fourth Framework Programme, the Ph...

  9. Resuspension of fission products during severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    This report investigates the influence of resuspension phenomena on the overall radiological source term of core melt accidents in a pressurized water reactor. A review of the existing literature is given and the literature data are applied to calculations of the source term. A large scatter in the existing data was found. Depending on the scenario and on the data set chosen for the calculations the relative influence of resuspended fission products on the source term ranges from dominant to negligible. (orig.)

  10. Separation of cesium from other fission products using ZrP-80-APW-20

    International Nuclear Information System (INIS)

    An ion exchange method for separation of Cs(137Cs) in the presence of long-lived fission products such as 95Zr, 95Nb etc., was developed. Selectively 137Cs was observed on the column whereas Zr and Nb which were complexed as oxalates passed through. 137Cs was eluted using 4 M nitric acid and 4 m ammonium nitrate solution. (author). 4 refs., 2 tabs

  11. The behaviour of fission products in fuel elements of the AVR-core

    International Nuclear Information System (INIS)

    The fuel elements of the THTR-1-type-, THTR-2-type-, and CFB-2-type reactor have an extraordinary retention capacity for all fission products. Pressed-carbide fuel elements, however, at a temperature above 3000C release considerable Sr- and Eu-activities as a result of diffusion through intact PyC-layers. For GLE-1 and GFB-1 fuel elements a considerable release of cesium has been observed which is caused by defective coated particles. (DG)

  12. Corrosion mechanisms of containment glasses for fission products

    International Nuclear Information System (INIS)

    After a review of nuclear energy production and waste vitrification principles, the aqueous corrosion mechanisms of the containment glasses and the various parameters affecting the corrosion are studied: effects of glass composition, temperature, lixiviation agent pH, lixiviation duration and mode. Conventional mass loss measurement and solution analyses are coupled to sophisticated surface analysis techniques. The hydrolyzed layer formation and the solubility limits are discussed. 87 figs., 30 tabs., 144 refs

  13. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    Science.gov (United States)

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  14. Fission Product Transport in TRISO Particle Layers under Operating and Off-Normal Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Van der Ven, Anton; Was, Gary; Wang, Lumin; Taheri, Mitra

    2014-07-07

    The objective of this project is to determine the diffusivity and chemical behavior of key fission products (ag, Cs, I. Te, Eu and Sr) through SiC and PyC both thermally, under irradiation, and under stress using FP introduction techniques that avoid the pitfalls of past experiments. The experimental approach is to create thin PyC-SiC couples containing the fission product to be studied embedded in the PyC layer. These samples will then be subjected to high temperature exposures in a vacuum and also to irradiation at high temperature, and last, to irradiation under stress at high temperature. The PyC serves as a host layer, providing a means of placing the fission product close to the SiC without damaging the SiC layer by its introduction or losing the FP during heating. Experimental measurements of grain boundary structure and distribution (EBSD, HRTEM, APT) will be used in the modeling effort to determine the qualitative dependence of FP diffusion coefficients on grain boundary orientation, temperature and stress.

  15. Comparison of neutron cross sections for selected fission products and isotopic composition analyses with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Gil, C. S.; Kim, J. D.; Jang, J. H.; Lee, Y. D. [KAERI, Taejon (Korea)

    2003-10-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI-BNL international collaboration have been compared with the ENDF/B-VI release 7. Also, the influence of the new evaluations on isotopic compositions of the fission products as a function of burnup has been analyzed through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69 group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including new evaluations in resonance region covering thermal region, and ENDF/B-VII expected including those in upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows maximum difference of 4.78% compared to ENDF/B-VI.7. However, the isotopic compositions of all fission products calculated with ENDF/B-VII shows no differences compared to ENDF/B-VI.7.

  16. Comparisons of Neutron Cross Sections and Isotopic Composition Calculations for Fission-Product Evaluations

    Science.gov (United States)

    Kim, Do Heon; Gil, Choong-Sup; Chang, Jonghwa; Lee, Yong-Deok

    2005-05-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI (Korea Atomic Energy Research Institute)-BNL (Brookhaven National Laboratory) international collaboration have been compared with ENDF/B-VI.7. Also, the influence of the new evaluations on the isotopic composition calculations of the fission products has been estimated through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69-group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including the new evaluations in the resonance region covering the thermal region, and the expected ENDF/B-VII including those in the upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows a maximum difference of 5.02% compared to ENDF/B-VI.7. However, the isotopic compositions of all the fission products calculated with the expected ENDF/B-VII show no differences when compared to ENDF/B-VI.7 for the thermal reactor benchmark cases.

  17. Linear Free Energy Correlations for Fission Product Release from the Fukushima-Daiichi Nuclear Accident

    Energy Technology Data Exchange (ETDEWEB)

    Abrecht, David G.; Schwantes, Jon M.

    2015-03-03

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes, et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the source of the radionuclides to be from active reactors rather than the spent fuel pool. Linear correlations of the form ln χ = -α (ΔGrxn°(TC))/(RTC)+β were obtained between the deposited concentration and the reduction potential of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn(TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2130 K and 2220 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, 151Sm through atmospheric venting and releases during the first month following the accident were performed, and indicate large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

  18. Factors affecting the corrosion of SiC layer by fission product palladium

    International Nuclear Information System (INIS)

    HTR is one of the advanced nuclear reactors which has inherent safety system, graphite moderated and helium gas cooled. In general, these reactors are designed with the TRISO coated particle consist of four coating layers that are porous pyrolytic carbon (PyC). inner dense PyC (IPyC), silicon carbide (SiC), and outer dense PyC (OPyC). Among the four coating layers, the SiC plays an important role beside in retaining metallic fission products, it also provides mechanical strength to fuel particle. However, results of post irradiation examination indicate that fission product palladium can react with and corrode SiC layer, This assessment is conducted to get the comprehension about resistance of SiC layer on irradiation effects, especially in order to increase the fuel bum-up. The result of this shows that the corrosion of SiC layer by fission product palladium is beside depend on the material characteristics of SiC, and also there are other factors that affect on the SiC layer corrosion. Fuel enrichment, bum-up, and irradiation time effect on the palladium flux in fuel kernel. While, the fuel density, vapour pressure of palladium (the degree depend on the irradiation temperature and kernel composition) effect on palladium migration in fuel particle. (author)

  19. From EXILL (EXogam at the ILL to FIPPS (FIssion Product Prompt γ-ray Spectrometer

    Directory of Open Access Journals (Sweden)

    Blanc A.

    2015-01-01

    Full Text Available Within the EXILL campaign a large and efficient cluster of Ge-detectors was installed around a very well collimated neutron beam. This has allowed to carry out rather complete spectroscopic studies close to the line of stability using the (n,γ reaction. Neutron rich isotopes were produced by neutron induced fission and prompt spectroscopy was carried out. The isotope selection in this setup was based on a partially known level scheme and the use of triple coincidences. The latter is limiting the statistical sensitivity in the case of weak production yields. Based on the experiences of these campaigns we are currently developing a new setup: FIPPS (FIssion Product Prompt Spectroscopy. This setup combines a collimated neutron beam, a highly efficient cluster of Ge detectors, a gas filled magnet and auxiliary detectors. The presence of the gas filled magnet will allow us to identify fission products directly and should give access to a new quality of studies if compared to the EXILL campaign. The EXILL campaign and the FIPPS project are presented.

  20. ACRR [Annular Core Research Reactor] fission product release tests: ST-1 and ST-2

    International Nuclear Information System (INIS)

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model. 8 refs., 8 figs., 2 tabs

  1. Linear free energy correlations for fission product release from the Fukushima-Daiichi nuclear accident.

    Science.gov (United States)

    Abrecht, David G; Schwantes, Jon M

    2015-03-01

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the initial source of the radionuclides to the environment to be from active reactors rather than the spent fuel pool. Linear correlations of the form In χ = −α ((ΔGrxn°(TC))/(RTC)) + β were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn (TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2015 and 2060 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, and 151Sm through atmospheric venting during the first month following the accident were obtained, indicating that large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores. PMID:25675358

  2. Linear free energy correlations for fission product release from the Fukushima-Daiichi nuclear accident.

    Science.gov (United States)

    Abrecht, David G; Schwantes, Jon M

    2015-03-01

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the initial source of the radionuclides to the environment to be from active reactors rather than the spent fuel pool. Linear correlations of the form In χ = −α ((ΔGrxn°(TC))/(RTC)) + β were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn (TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2015 and 2060 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, and 151Sm through atmospheric venting during the first month following the accident were obtained, indicating that large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

  3. Disposition of plutonium-239 via production of fission molybdenum-99.

    Science.gov (United States)

    Mushtaq, A

    2011-04-01

    A heritage of physical consequences of the U.S.-Soviet arms race has accumulated, the weapons-grade plutonium (WPu), which will become excess as a result of the dismantlement of the nuclear weapons under the arms reduction agreements. Disposition of Pu has been proposed by mixing WPu with high-level radioactive waste with subsequent vitrification into large, highly radioactive glass logs or fabrication into mixed oxide fuel with subsequent irradiation in existing light water reactors. A potential option may be the production of medical isotope molybdenum-99 by using Pu-239 targets.

  4. Neutron activation analysis (NAA), radioisotope production via neutron activation (PNA) and fission product gas-jet (GJA)

    Energy Technology Data Exchange (ETDEWEB)

    Gaeggeler, H.W. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-11-01

    Three different non-diffractive applications of neutrons are outlined, neutron activation analysis, production of radionuclides, mostly for medical applications, and production of short-lived fission nuclides with a so-called gas-jet. It is shown that all three devices may be incorporated into one single insert at SINQ due to their different requests with respect to thermal neutron flux. Some applications of these three facilities are summarized. (author) 3 figs., 1 tab., 8 refs.

  5. FIssion Product Prompt γ-ray spectrometer: Development of an instrumented gas-filled magnetic spectrometer at the ILL

    Science.gov (United States)

    Blanc, A.; Chebboubi, A.; Faust, H.; Jentschel, M.; Kessedjian, G.; Köster, U.; Materna, T.; Panebianco, S.; Sage, C.; Urban, W.

    2013-12-01

    Accurate thermal neutron-induced fission data are important for applications in reactor physics as well as for fundamental nuclear physics. FIPPS is the new FIssion Product Prompt γ-ray Spectrometer being developed at the Institut Laue Langevin for neutron-induced fission studies. FIPPS is based on the combination of a large Germanium detector array surrounding a fission target, a Time-Of-Flight detector and a Gas-Filled Magnet (GFM) to identify mass, nuclear charge and kinetic energy of one of the fission fragments. The GFM will be instrumented with a Time-Projection Chamber (TPC) for individual 3D tracking of the fragments. A conceptual design study of the new spectrometer is presented.

  6. Thermodynamics mechanisms of fission product retention in nuclear plants illustrated by the properties of the HTR reactor

    International Nuclear Information System (INIS)

    Starting from the first law of thermodynamics, the theoretical principles for the description of interactions between fission products and other materials are derived step by step, using fundamental terms such as phase equilibria, mixtures and solutions. Thereafter, the concepts of Onsager's theory of irreversible thermodynamics are introduced. They serve as an example of modelling fission product transport with special respect to thermochemical properties. In the last chapter real technical concepts for fission product retention are evaluated using thermodynamic criteria. A fine distinction is performed between barrier-, filter- and sinkmechanisms for retention-purposes. One important result is, that a barrier-concept alone doesn't meet the challenge of nuclear power operation without the probability of hazardous accidents. The work is finished by a proposal to improve the fission product retention capabilities of HTR fuel-elements in combination with a coating of the fuel-pebbles. (orig./DG)

  7. High temperature studies of simulant fission products: part IV

    International Nuclear Information System (INIS)

    A thermal gradient system has been used to study the interaction of caesium iodide with boric acid in various atmospheres over the temperature range from 400 to 10000C. Specific analytical techniques have been used to determine the reaction products, and differential thermal analysis-thermogravimetric studies have been undertaken to assess the reaction kinetics. Solid caesium iodide and molten boric acid react in a diffusion-controlled manner with an activation energy of 190 ± 30 kJ mol-1 to give complex caesium borates (eg Cs2B10O16) and hydrogen iodide. This volatile iodine species interacts with 304 stainless steel to produce nickel and iron-based iodides within the surface oxide layer of the metal. (author)

  8. Background for Terrestrial Antineutrino Investigations: Radionuclide Distribution, Georeactor Fission Events, and Boundary Conditions on Fission Power Production

    CERN Document Server

    Herndon, J M; Edgerley, Dennis A.

    2005-01-01

    Estimated masses of fissioning and non-fissioning radioactive elements and their respective distributions within the Earth are presented, based upon the fundamental identity of the components of the interior 82% of the Earth, the endo-Earth, with corresponding components of the Abee enstatite chondrite meteorite. Within limits of existing data, the following generalizations concerning the endo-Earth radionuclides can be made: (1) Most of the K-40 may be expected to exist in combination with oxygen in the silicates of the lower mantle, perhaps being confined to the upper region of the lower mantle where it transitions to the upper mantle; (2) Uranium may be expected to exist at the center of the Earth where it may undergo self-sustaining nuclear fission chain reactions, but there is a possibility that some non-fissioning uranium may be found scattered diffusely within the core floaters which are composed of CaS and MgS; and, (3) Thorium may be expected to occur within the core floaters at the core-mantle bound...

  9. Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

    OpenAIRE

    Salahuddin Asif; Iqbal Masood

    2013-01-01

    Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor....

  10. Energy from nuclear fission an introduction

    CERN Document Server

    De Sanctis, Enzo; Ripani, Marco

    2016-01-01

    This book provides an overview on nuclear physics and energy production from nuclear fission. It serves as a readable and reliable source of information for anyone who wants to have a well-balanced opinion about exploitation of nuclear fission in power plants. The text is divided into two parts; the first covers the basics of nuclear forces and properties of nuclei, nuclear collisions, nuclear stability, radioactivity, and provides a detailed discussion of nuclear fission and relevant topics in its application to energy production. The second part covers the basic technical aspects of nuclear fission reactors, nuclear fuel cycle and resources, safety, safeguards, and radioactive waste management. The book also contains a discussion of the biological effects of nuclear radiation and of radiation protection, and a summary of the ten most relevant nuclear accidents. The book is suitable for undergraduates in physics, nuclear engineering and other science subjects. However, the mathematics is kept at a level that...

  11. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James

    2012-12-19

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.

  12. Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

    International Nuclear Information System (INIS)

    A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D+ beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the three-dimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S(α, β) thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors. (authors)

  13. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    Science.gov (United States)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by "bottlenecks" in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

  14. In-pile release behavior of gaseous fission product from VHTR fuel

    International Nuclear Information System (INIS)

    Loose TRISO coated fuel particles and compacts fabricated under VHTR design were irradiated by the sweep gas capsules and the in-pile release behavior of gaseous fission products was studied. By the release from the intact fuel the activation energies of the release and the diffusion coefficients of krypton in the pyrocarbon layer were estimated. The relation between release and uranium contamination, and the release mechanism depending on the release level were made clear in the releases from both the intact and the failed fuels. Failure fraction of the loose coated fuel particles was estimated and the retention of the fission gas in the matrix was recognized by analysis of the release from the fuels containing the failed coated fuel particles

  15. Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

    Institute of Scientific and Technical Information of China (English)

    WANG Xin-Hua; GUO Hai-Ping; MOU Yun-Feng; ZHENG Pu; LIU Rong; YANG Xiao-Fei; YANG Jian

    2013-01-01

    A fusion-fission hybrid conceptual reactor is established.It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium.The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D+ beam of the Cockcroft-Walton neutron generator in direct current mode.The measured TPR distribution is compared with the calculated results obtained by the threedimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file.The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S(α,β) thermal scattering model,so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors.

  16. Release behavior of metallic fission products from HTGR fuel particles at 1600 to 1900 C

    International Nuclear Information System (INIS)

    Release behavior of metallic fission products from the Triso-coated UO2 particles was studied by postirradiation heating tests in the temperature range 1600 to 1900 C (1873 to 2173 K) and subsequent postheating examinations. The fission gas release monitoring and the postheating examinations revealed that no pressure vessel failure occurred in the tests. Ceramographic observations showed no palladium attack and thermal decomposition of SiC. 137Cs, 134Cs, 110mAg, 154Eu and 155Eu were released from the coated particles through the coating layers during postirradiation heating. The diffusion coefficient of 137Cs in the SiC layer was evaluated from the release curves based on a simple diffusion model assuming a one-layer coated particle. Fractional release measurements suggested that the diffusion coefficient of 110mAg in SiC be larger than that of 137Cs. (orig.)

  17. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    International Nuclear Information System (INIS)

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale

  18. Calculation of the decay power of fission products considering neutron capture transformation

    International Nuclear Information System (INIS)

    The decay power of fission products has been calculated taking into consideration the neutron capture transformation of each nuclide and its beta decay. The nuclear data library contains 1114 nuclides of which 144 are stable. Neutron capture transformation is considered for 59 nuclides, 31 of which are stable. The atom number of each nuclide is calculated analytically with code DCHAIN. The effect of neutron capture transformation in the decay power of fission products was examined by varying the neutron spectrum, neutron flux, fissioning nuclide, and irradiation and cooling time. From the results obtained the following were revealed: The effect of neutron capture increases with neutron flux and irradiation time, and it becomes salient beyond 105 sec in cooling time. It is small for less than the 104 sec which is important in the design of ECCS (emergency core cooling system) of a light-water reactor. In this region the decay power changes are small, less than 0.2%, by the neutron capture for the thermal fission of 235U irradiated for one year to thermal neutron flux 3 x 1013 n/cm2/sec. The effect of neutron capture has peaks around cooling time 106 sec and 108 sec; it is negligible beyond 109 sec. The changes in decay power are 2.4%, 10.5% and 0.2% at cooling time 106 sec, 108 sec and 109 sec, respectively, in the above irradiation. Around 106 sec, the change in decay power is mainly from the contributions of 134Cs (17%), sup(148m)Pm(60%) and 148Pm(14%). Around 108 sec 134Cs(98%) alone contributes to the change in decay power. (author)

  19. Diffusion of Zr, Ru, Ce, Y, La, Sr and Ba fission products in UO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Perriot, R., E-mail: rperriot@lanl.gov; Liu, X.-Y.; Stanek, C.R.; Andersson, D.A.

    2015-04-15

    The diffusivity of the solid fission products (FP) Zr (Zr{sup 4+}), Ru (Ru{sup 4+}, Ru{sup 3+}), Ce (Ce{sup 4+}), Y (Y{sup 3+}), La (La{sup 3+}), Sr (Sr{sup 2+}) and Ba (Ba{sup 2+}) by a vacancy mechanism has been calculated, using a combination of density functional theory (DFT) and empirical potential (EP) calculations. The activation energies for the solid fission products are compared to the activation energy for Xe fission gas atoms calculated previously. Apart from Ru, the solid fission products all exhibit higher activation energy than Xe. For all solid FPs except Y{sup 3+}, the migration of the FP has lower barrier than the migration of a neighboring U atom, making the latter the rate limiting step for direct migration. An indirect mechanism, consisting of two successive migrations around the FP, is also investigated. The calculated diffusivities show that most solid fission products diffuse with rates similar to U self-diffusion. However, Ru, Ba and Sr exhibit faster diffusion than the other solid FPs, with Ru{sup 3+} and Ru{sup 4+} diffusing even faster than Xe for T < 1200 K. The diffusivities correlate with the observed fission product solubility in UO{sub 2}, and the tendency to form metallic and oxide second phase inclusions.

  20. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    Science.gov (United States)

    Norman, Eric B.; Prussin, Stanley G.

    2009-01-27

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  1. Radioactive Ion Beam Production by Fast-Neutron-Induced Fission in Actinide Targets at EURISOL

    CERN Document Server

    Herrera-Martínez, Adonai

    The European Isotope Separation On-Line Radioactive Ion Beam Facility (EURISOL) is set to be the 'next-generation' European Isotope Separation On-Line (ISOL) Radioactive Ion Beam (RIB) facility. It will extend and amplify current research on nuclear physics, nuclear astrophysics and fundamental interactions beyond the year 2010. In EURISOL, the production of high-intensity RIBs of specific neutron-rich isotopes is obtained by inducing fission in large-mass actinide targets. In our contribution, the use of uranium targets is shown to be advantageous to other materials, such as thorium. Therefore, in order to produce fissions in U-238 and reduce the plutonium inventory, a fast neutron energy spectrum is necessary. The large beam power required to achieve these RIB levels requires the use of a liquid proton-to-neutron converter. This article details the design parameters of the converter, with special attention to the coupled neutronics of the liquid converter and fission target. Calculations performed with the ...

  2. Behaviour of fission products under severe PWR accident conditions. The VERCORS experimental programme-Part 3: Release of low-volatile fission products and actinides

    International Nuclear Information System (INIS)

    The VERCORS analytical programme consisted of a series of tests carried out on irradiated PWR fuel samples. The tests - funded jointly by EDF and IRSN - were carried out by the Commissariat a l'Energie Atomique (CEA) at their Grenoble site. They were performed in a hot cell belonging to the Active Materials Analysis Laboratory (LAMA). The general outline of the programme was set out in a first article (of a series of 3), which described the different levels of fission products (FP) volatility and their characteristics. This led to a classification into five main categories of volatility and/or behaviour: (1) Volatile FP including fission gases, iodine, caesium, antimony, tellurium, cadmium, rubidium and silver; (2) Semi-volatile FP, a category made up of molybdenum, rhodium, barium, palladium and technetium; (3) Low-volatile FP comprising ruthenium, cerium, strontium, yttrium, europium, niobium and lanthanum with generally low but significant release; (4) Non-volatile FP including zirconium, neodymium and praseodymium; and lastly (5) Actinides which group together uranium, plutonium, neptunium, americium and curium. The specific behaviour of fission gases and volatile FP is dealt with in the second article, which also includes the specific characteristics of volatile FP regarding transport. The main variables (i.e. temperature, which is the main variable at least until loss of sample geometry, oxidising-reducing conditions, burn-up, interactions with the cladding and/or the structural components, the nature of the fuel, and finally the state of the fuel) affecting the kinetics and/or the released fraction of these same FP could also be identified. This final article represents the Third Part of the series. It concerns the release of actinides and less volatile FP, in keeping with the classification by categories previously identified, which are as follows: (1) semi-volatile FP, comprising of Mo, Ba, Rh, Pd, Tc, (2) low-volatile FP, comprising of Sr, Y, Nb, Ru, La

  3. Seminar on Fission VI

    Science.gov (United States)

    Wagemans, Cyriel; Wagemans, Jan; D'Hondt, Pierre

    2008-04-01

    Topical reviews. Angular momentum in fission / F. Gönnenwein ... [et al.]. The processes of fusion-fission and quasi-fission of heavy and super-heavy nuclei / M. G. Itkis ... [et al.] -- Fission cross sections and fragment properties. Minor-actinides fission cross sections and fission fragment mass yields via the surrogate reaction technique / B. Jurado ... [et al.]. Proton-induced fission on actinide nuclei at medium energy / S. Isaev ... [et al.]. Fission cross sections of minor actinides and application in transmutation studies / A. Letourneau ... [et al.]. Systematics on even-odd effects in fission fragments yields: comparison between symmetric and asymmetric splits / F. Rejmund, M Caamano. Measurement of kinetic energy distributions, mass and isotopic yields in the heavy fission products region at Lohengrin / A. Bail ... [et al.] -- Ternary fission. On the Ternary [symbol] spectrum in [symbol]Cf(sf) / M. Mutterer ... [et al.]. Energy degrader technique for light-charged particle spectroscopy at LOHENGRIN / A. Oberstedt, S. Oberstedt, D. Rochman. Ternary fission of Cf isotopes / S. Vermote ... [et al.]. Systematics of the triton and alpha particle emission in ternary fission / C. Wagemans, S. Vermote, O. Serot -- Neutron emission in fission. Scission neutron emission in fission / F.-J. Hambsch ... [et al.]. At and beyond the Scission point: what can we learn from Scission and prompt neutrons? / P. Talou. Fission prompt neutron and gamma multiplicity by statistical decay of fragments / S. Perez-Martin, S. Hilaire, E. Bauge -- Fission theory. Structure and fission properties of actinides with the Gogny force / H. Goutte ... [et al.]. Fission fragment properties from a microscopic approach / N. Dubray, H. Goutte, J.-P. Delaroche. Smoker and non-smoker neutron-induced fission rates / I. Korneev ... [et al.] -- Facilities and detectors. A novel 2v2E spectrometer in Manchester: new development in identification of fission fragments / I. Tsekhanovich ... [et al

  4. Production, disposal, and relative toxicity of long-lived fission products and actinides in the radioactive wastes from nuclear fuel cycles

    International Nuclear Information System (INIS)

    Chapters are devoted to the following topics: predicted future development of nuclear energy in the German Federal Republic and in Western Europe, fuel cycle variations and production of fission products and actinides in the radioactive waste from reprocessed nuclear fuels, long-lived fission products and actinides in the waste streams from the reprocessing of nuclear fuels, relative toxicity index, presently preferred waste management concepts, and alternative concepts for the elimination of high-level wastes

  5. Deposition of fission and activation products after the Fukushima Dai-ichi nuclear power plant accident.

    Science.gov (United States)

    Shozugawa, Katsumi; Nogawa, Norio; Matsuo, Motoyuki

    2012-04-01

    The Great Eastern Japan Earthquake on March 11, 2011, damaged reactor cooling systems at Fukushima Dai-ichi nuclear power plant. The subsequent venting operation and hydrogen explosion resulted in a large radioactive nuclide emission from reactor containers into the environment. Here, we collected environmental samples such as soil, plant species, and water on April 10, 2011, in front of the power plant main gate as well as 35 km away in Iitate village, and observed gamma-rays with a Ge(Li) semiconductor detector. We observed activation products ((239)Np and (59)Fe) and fission products ((131)I, (134)Cs ((133)Cs), (137)Cs, (110m)Ag ((109)Ag), (132)Te, (132)I, (140)Ba, (140)La, (91)Sr, (91)Y, (95)Zr, and (95)Nb). (239)Np is the parent nuclide of (239)Pu; (59)Fe are presumably activation products of (58)Fe obtained by corrosion of cooling pipes. The results show that these activation and fission products, diffused within a month of the accident. PMID:22266366

  6. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Jason M. Harp; Paul A. Demkowicz

    2014-10-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10-4 to 10-5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.

  7. Study of the behaviour of tetracycline as fission products extracting agent

    International Nuclear Information System (INIS)

    Both spectrophotometric and potentiometric titration techniques were used to show the formation of complexes between tetracycline and the elements: zirconium, uranium, molybdenum, strontium, barium and ruthenium. It has been verified that tetracycline does not form complexes with cesium, tellurium and iodine. Those techniques have also been used to determine the sites on the tetracycline molecule at which ions may be bound. The behaviour of tetracycline as an extracting agent for those elements, as well as for niobium and technetium has been studied and the influence of the acidity of the aqueous phase upon extraction of the elements mentioned has been considered. Extraction experiments were carried out in the presence of chloride, perchlorate, nitrate and sulfate ions. Studies have been made to determine whether or not the complex extracted into organic phase is really the complex formed between tetracycline and the elements considered as well as to determine the time of shaking necessary so that the equilibrium between the phases is attained. Based on all information obtained from extraction experiments made for uranium and the fission products Zr-95, Nb-95, Ce-141, La-140, Ru-103, Ba-140 and Cs-137, the possibility of using tetracycline for separating those fission products from each other and from uranium has been studies and a scheme for simultaneous separation of those elements has been proposed. The same study has been made for I-131, Tc-99m, Mo-99, Te-132, Np-239 and uranium. The method described is applicable to the separation of some fission products existing in solutions at tracer levels, and not to be used in nuclear fuel reprocessing or any other industrial application. (Author)

  8. Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

    Science.gov (United States)

    Johnson, L.; Günther-Leopold, I.; Kobler Waldis, J.; Linder, H. P.; Low, J.; Cui, D.; Ekeroth, E.; Spahiu, K.; Evins, L. Z.

    2012-01-01

    Studies of the rapid aqueous release of fission products from UO 2 and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of 137Cs and 129I for a number of fuels irradiated to burn-ups of 50-75 MWd/kgU. In addition a method for analysis of leaching solutions for 79Se was developed. The results of the studies show that the fractional release of 137Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional 129I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association 137Cs and 129I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for 79Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that 79Se is not preferentially leached from spent fuel.

  9. Gamma emitting fission products in surface sediments of the Ravenglass estuary

    Energy Technology Data Exchange (ETDEWEB)

    Aston, S.R.; Stanners, D.A. (Lancaster Univ. (UK))

    1982-04-01

    The occurrence of some fission products from the Sellafield (formerly Windscale) nuclear fuel reprocessing facility has been determined for surface sediments from forty locations in the Ravenglass estuary, North-West England. The influence of the silt-sized fraction in the sediments on the geographic distribution of /sup 137/Cs is clearly important, and to a lesser extent also influences the distributions of /sup 106/Ru, /sup 134/Cs + /sup 95/Zr/Nb and /sup 144/Ce. The data are compared with recently published results reported by the Ministry of Agriculture, Fisheries and Food for a monitoring site in this estuary.

  10. Fission products plate-out analysis code in the HTGR: PLAIN

    International Nuclear Information System (INIS)

    Fission products plate-out analysis code in the HTGR, PLAIN, has been developed to calculate the plate-out distribution of high temperature gas cooled reactor. The code includes the following features: (1) The adsorption-desorption on the wall and the diffusion into the wall material are considered as a plate-out mechanism. (2) Mass transfer coefficient, adsorption rate, desorption rate, diffusion coefficient in the wall material and sublimation rate are calcualted in the code. (3) Numerical results are calculated by using Laplace inversion method. This report describes the analytical models, physical and chemical constants, numerical mehtod, users' manuals and the calculational results of experiments. (author)

  11. BIPAL - a data library for computing the burnup of fissionable isotopes and products of their decay

    International Nuclear Information System (INIS)

    The BIPAL databank contains data on 100 heavy metal isotopes starting with 206Tl and finishing with 253Es. Four are stable, the others are unstable. The following data are currently stored in the databank: the serial number and name of isotopes, decay modes and, for stable isotopes, the isotopic abundance (%), numbers of P decays and Q captures, numbers of corresponding final products, branching ratios, half-lives and their units, decay constants, thermal neutron captures, and fission cross sections, and other data (mainly alpha, beta and gamma intensities). The description of data and a printout of the BIPAL library are presented. (J.B.)

  12. ENDF/B-IV fission-product files: summary of major nuclide data

    International Nuclear Information System (INIS)

    The major fission-product parameters [sigma/sub th/, RI, tau/sub 1/2/, E-bar/sub β/, E-bar/sub γ/, E-bar/sub α/, decay and (n,γ) branching, Q, and AWR] abstracted from ENDF/B-IV files for 824 nuclides are summarized. These data are most often requested by users concerned with reactor design, reactor safety, dose, and other sundry studies. The few known file errors are corrected to date. Tabular data are listed by increasing mass number

  13. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W [Westinghouse Savannah River Co., Aiken, SC (USA); Hagrman, D L [EG and G Idaho, Inc., Idaho Falls, ID (USA); McClure, P R; Leonard, M T [Science Applications International Corp., Albuquerque, NM (USA)

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  14. Neutron capture cross-section of fission products in the European activation file EAF-3

    Energy Technology Data Exchange (ETDEWEB)

    Kopecky, J.; Delfini, M.G.; Kamp, H.A.J. van der; Gruppelaar, H.; Nierop, D. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands))

    1992-05-01

    This paper contains a description of the work performed to extend and revise the neutron capture data in the European Activation File (EAF-3) with emphasis on nuclides in the fission-product mass range. The starter was the EAF-1 data file from 1989. The present version, EAF/NG-3, contains (n,[gamma]) excitation functions for all nuclides (729 targets) with half-lives exceeding 1/2 day in the mass range from H-1 to Cm-248. The data file is equipped with a preliminary uncertainty file, that will be improved in the near future. (author). 19 refs.; 5 figs.; 3 tabs.

  15. Transport of fission products with a helium gas-jet at TRIGA-SPEC

    Science.gov (United States)

    Eibach, M.; Beyer, T.; Blaum, K.; Block, M.; Eberhardt, K.; Herfurth, F.; Geppert, C.; Ketelaer, J.; Ketter, J.; Krämer, J.; Krieger, A.; Knuth, K.; Nagy, Sz.; Nörtershäuser, W.; Smorra, C.

    2010-02-01

    A helium gas-jet system for the transport of fission products from the research reactor TRIGA Mainz has been developed, characterized and tested within the TRIGA-SPEC experiment. For the first time at TRIGA Mainz carbon aerosol particles have been used for the transport of radionuclides from a target chamber with high efficiency. The radionuclides have been identified by means of γ-spectroscopy. Transport time, efficiency as well as the absolute number of transported radionuclides for several species have been determined. The design and the characterization of the gas-jet system are described and discussed.

  16. HYPERFUSE: a novel inertial confinement system utilizing hypervelocity projectiles for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137Cs or 90Sr. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n, 2n), (n, α), etc.) that convert the long lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product

  17. Design, construction, and testing of a 20000C furnace and fission product collection system

    International Nuclear Information System (INIS)

    An induction furnace, capable of operation at 20000C in steam, was developed to conduct product release tests. The test specimen and steam atmosphere are contained in a stabilized ZrO2 furnace tube, which is heated by a concentric susceptor of either tungsten or graphite. A two-color optical pyrometer and high-temperature thermocouples are used for temperature measurement. The furnace has operated reliably for periods up to 30 min at test temperatures of 1400 to 20000C with steam flowing at approx. 1 L/min. The apparatus for collecting the released fission products includes a TGT, an aerosol deposition sampler, a series of glass fiber filters, and heated charcoal. A steam condenser and cooled charcoal, for inert gas adsorption, are located further downstream. The principal analytical techniques used for fission product identification and measurement are: (1) gamma spectroscopy, for all radionuclides on all test components; (2) spark-source mass spectrometry, for all elements, primarily deposits on the thermal gradient tube and filters; and (3) neutron activation, for iodine on selected test components. 16 references, 26 figures, 7 tables

  18. Immobilization of fission products in low-temperature ceramic waste forms

    International Nuclear Information System (INIS)

    Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically bonded phosphate ceramics (CBPCs) for use in solidifying and stabilizing low-level mixed wastes. The focus of this work is development of CBPCs for use with fission-product wastes generated from high-level waste (HLW) tank cleaning or other decontamination and decommissioning activities. The volatile fission products such as Tc, Cs, and Sr removed from HLW need to be disposed of in a low-temperature immobilization system. Specifically, this paper reports on the solidification and stabilization of separated 99Tc from Los Alamos National Laboratory's complexation-elution process. Using rhenium as a surrogate form technetium, we fabricated CBPC waste forms by acid-base reactions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with 35 wt.% waste loading. Standard leaching tests such as ANS 16.1 and PCT were conducted on the final waste forms. In addition, stability of the waste forms in aqueous environments was evaluated by long-term water-immersion tests

  19. Monte Carlo Models for the Production of beta-delayed Gamma Rays Following Fission of Special Nuclear Materials

    Energy Technology Data Exchange (ETDEWEB)

    Pruet, J; Prussin, S; Descalle, M; Hall, J

    2004-02-03

    A Monte Carlo method for the estimation of {beta}-delayed {gamma}-ray spectra following fission is described that can accommodate an arbitrary time-dependent fission rate and photon collection history. The method invokes direct sampling of the independent fission yield distributions of the fissioning system, the branching ratios for decay of individual fission products and the spectral distributions for photon emission for each decay mode. Though computationally intensive, the method can provide a detailed estimate of the spectrum that would be recorded by an arbitrary spectrometer, and can prove useful in assessing the quality of evaluated data libraries, for identifying gaps in these libraries, etc. The method is illustrated by a first comparison of calculated and experimental spectra from decay of short-lived fission products following the reactions {sup 235}U(n{sub th}, f) and {sup 239}Pu(n{sub th}, f). For general purpose transport calculations, where detailed consideration of the large number of individual {gamma}-ray transitions in a spectrum may be unnecessary, it is shown that an accurate and simple parameterization of a {gamma}-ray source function can be obtained. These parametrizations should provide high-quality average spectral distributions that should prove useful in calculations describing photons escaping from thick attenuating media.

  20. Production and study of fission fragments, from Lohengrin to Alto; Production et etude des fragments de fission, de Lohengrin a Alto

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, F

    2005-06-15

    The study of nuclei far from stability is constitutive of the history of nuclear physics at its very beginning and has been making considerable great strides since then. The study of these nuclei give the opportunity to reach new information on the nuclear structure and thus to measure the solidity of our knowledge on nuclear matter and its validity when it is pushed to its limits. The reaction selected for the production of exotic nuclei in the framework of the PARRNe program is the fission of uranium 238. The nuclei produced have an intermediate mass and are very rich in neutrons. The technique to recover them in order to accelerate them is the thick target method called also the Isol technique. The installation of the ancient Lep injector at the Tandem line in Orsay (IPN) is expected to increase by a factor 100 the production rate of exotic nuclei in the PARRNe program, it is the Alto project. The work presented here concerns studies carried out at the Lohengrin spectrometer installed at the ILL in Grenoble, and at the Tandem installation in Orsay. This document is divided into 4 parts: 1) in flight techniques at Lohengrin, 2) the Isol technique, 3) magic numbers in the domain N=50, and 4) the Alto project.

  1. Investigation of Adsorption Behavior of Ba and Other Fission Products on the Sr·spc Chromatographic Column by Static Method

    Institute of Scientific and Technical Information of China (English)

    YANG; Lei; MA; Peng; YANG; Su-liang; LIANG; Xiao-hu

    2012-01-01

    <正>Adsorption behavior of Ba, Cs and some other fission products on the Sr·spc resin has been investigated for the purpose of extracting 141Ba from the fission product. Sr·spc resin with the main functional group of 18-crown-6 ether was purchased from US. Eichrom Company. Tracers of Ba, Cs and some other fission products were acquired from an irradiated U target.

  2. Chemical thermodynamics of Cs and Te fission product interactions in irradiated LMFBR mixed-oxide fuel pins

    Science.gov (United States)

    Adamson, M. G.; Aitken, E. A.; Lindemer, T. B.

    1985-02-01

    A combination of fuel chemistry modelling and equilibrium thermodynamic calculations has been used to predict the atom ratios of Cs and Te fission products (Cs:Te) that find their way into the fuel-cladding interface region of irradiated stainless steel-clad mixed-oxide fast breeder reactor fuel pins. It has been concluded that the ratio of condensed, chemically-associated Cs and Te in the interface region,Čs:Te, which in turn determines the Te activity, is controlled by an equilibrium reaction between Cs 2Te and the oxide fuel, and that the value of Čs:Te is, depending on fuel 0:M, either equal to or slightly less than 2:1. Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), the observed out-of-pile Cs:Te thresholds for FCCI (4˜:1) and FPLME (2˜:1) have been rationalized in terms of Cs:Te thermochemistry and phase equilibria. Also described in the paper is an updated chemical evolution model for reactive/volatile fission product behavior in irradiated oxide pins.

  3. Education in Basic Skills and Training for Productive Work

    Science.gov (United States)

    Labarca, Guillermo

    1998-09-01

    The success of global policies and strategies aimed at training for productive work depends to a large extent on the level of development of basic skills among the work force and, likewise, training costs will vary according to the level of general preparation of those entering on the process. In view of the close relationship between the structure of the school system, the development of basic skills and actual training, different options are available to resolve imbalances between training for productive employment and previous basic education. Our conclusions are that training cannot replace basic education, that the process of technological change goes hand in hand with an increased demand for workers with a high level of education, that substituting training in specific skills for good basic education is not the most efficient option, and that one of the favorable effects of primary education is that it facilitates after- school training. This article seeks to identify certain dimensions of human resource training which are often overlooked in relation to both basic skills and specific training proper: namely, the imbalances existing between vocational training and previous education, and the options available for correcting them.

  4. Partitioning of selected fission products from irradiated oxide fuel induced by thermal treatment

    International Nuclear Information System (INIS)

    The release of fission products (FPs) from spent nuclear fuel (SNF) has been studied as a function of the temperature and redox conditions. The present paper concerns essentially the high temperature separation of Cs and Sr from irradiated pressurized (PWR) and boiling water reactor (BWR) fuel of different burn-up levels with use of an in-house designed system for inductive vaporization (InVap). Using thermodynamic calculations with the Module of Fission Product Release (MFPR) code along with annealing experiments on SNF in the InVap it was shown that the speciation of Cs and Sr, hence their release behavior at high temperature, is sensitive to the redox conditions during thermal treatment. It was demonstrated that annealing conditions in the InVap can be adjusted in the way to promote the release of selected FPs without significant loss of the fuel matrix or actinides: complete release of Cs and I was achieved during treatment of irradiated fuel at 1800 °C under reducing atmosphere (0.7% H2/Ar mixture). The developed partitioning procedure can be used for the SNF pretreatment as an advanced head-end step in the hydrometallurgical or pyrochemical reprocessing technology

  5. Breast cancer: evidence for a relation to fission products in the diet.

    Science.gov (United States)

    Sternglass, E J; Gould, J M

    1993-01-01

    To establish the possible relation between breast cancer mortality and low doses of radiation due to fission products in the environment, the mortality rates in the nine census regions of the United States for the years 1984-1988 were correlated with the cumulative airborne releases from all the nuclear plants in each region for the period 1970-1987. A high correlation coefficient of 0.91 was obtained for a logarithmic dependence on the total releases, consistent with an indirect action via free-radical oxygen at very low dose rates, in contrast to a direct action on DNA at high dose rates, explaining the wide differences in risk per unit dose obtained in earlier studies. The recent temporal changes of breast cancer rates in the New York metropolitan area including nearby Connecticut, Westchester, and Long Island were examined in relation to the releases from nearby nuclear plants and found to be consistent with a dominant role of short-lived fission products in drinking water and fresh milk. The results support a major role for nuclear plant releases in industrial countries in the recent rises of breast and other forms of cancers not related to smoking, especially among older persons, and strongly support the need to replace nuclear reactors with more benign ways to generate electricity.

  6. Distribution of fission products in Peach Bottom HTGR fuel element E11-07

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Bate, L.C.

    1977-04-01

    This is the second in a projected series of six post-irradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements. Element E11-07, the subject of this report, received an equivalent of 701 full-power days of irradiation prior to scheduled withdrawal. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a /sup 137/Cs inventory of 17 Ci in the graphite sleeve and 8.3 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides /sup 134/Cs, /sup 110m/Ag, /sup 60/Co, and /sup 154/Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the distribution of the beta emitters /sup 3/H, /sup 14/C, and /sup 90/Sr were obtained at six axial locations, four within the fueled region and one each above and below. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. These profiles reveal an increased degree of penetration of /sup 134/Cs, relative to /sup 137/Cs, evidently due to a longer time spent as xenon precursor. In addition to fission product distribution, the appearance of the element components was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed.

  7. Fission product release profiles from spherical HTR fuel elements at accident temperatures

    International Nuclear Information System (INIS)

    With the construction of the cold finger apparatus, a new method has been developed to determine fission product release profiles during heating tests of irradiated spherical fuel elements. 22 fuel elements with modern TRISO particles have been tested in the temperature range of 1500-25000C. In addition, experiments were done on seven UO2 samples at 1400 to 18000C. For heating times up to 100 hours at the maximum temperature, the following results were obtained: silver is the only fission product to be released at 1200-16000C by diffusion through intact SiC, but is of low significance in accident scenarios; caesium, iodine, strontium and noble gas releases up to 16000C are solely due to various forms of contamination. At 1700-18000C, corrosion-induced SiC defects cause the release of Cs, Sr, I/Xe/Kr. Above 20000C, thermal decomposition of the silicon carbide layer sets in, while pyrocarbons still remain intact. Around 16000C, the accident specific contribution of caesium, strontium, iodine and noble gas release is negligible. (orig./HP)

  8. A facility for recovery of uranium from the waste solutions of fission molybdenum production - AMOR II

    International Nuclear Information System (INIS)

    A facility is described for the recovery of the enriched uranium from the waste solutions of the fission molybdenum production. To tackle this task a solvent extraction process for the separation of uranium from fission products by tri-n-butylphosphate, diluted in tetrachloro-ethen has been developed. Owing to the special composition of the feed solution and the technical conditions the well-known PUREX-process had to be modified. The extraction and reextraction of uranium is carried out by a new mixer-settler unit which works without mechanical moving parts. The whole facility including the immediate storage tanks is installed in four semi-hot cells, shielded with 24 cm steel stones. The needed plant throughput has been assumed as 5 kg U (35% enriched) per year. It will be able to process this amount in 10-12 weeks (process time: 7 h/day). The facility was put into operation in 1985. The content of nuclear material is regularly being examined by the International Atomic Energy Agency according to the non-proliferation agreement. (author)

  9. Corrosion behavior of 9CrODS steel by simulated fission product cesium and tellurium

    International Nuclear Information System (INIS)

    Out-of-pile FCCI tests for 9CrODS steel were performed at 973 K by using simulated fission products Cs and Te under the oxygen potential in equilibrium with Fe/FeO and Cr/Cr2O3. Al2O3 powder were inserted to reduce a concentration of the Cs and Te in the system; its molar fraction is Cs:Te:Al2O3 = 1:1:1000. From EPMA and XRD analyses, Cr2O3 was formed at the most outer layer, which significantly suppressed the fission product corrosion. Cr2Te3 was also produced at the outer layer and interior of 9CrODS steel through liquid Te migration along grain boundaries. It was demonstrated the corrosion depth of 9CrODS steel is between PNC-FMS and PNC316, which were tested as reference. The Cs and Te assisted corrosion of 9CrODS steel was thermodynamically analyzed through the formation of Cs2O, Cs3CrO4, Cr2O3 and Cr2Te3

  10. Airborne measurements of fission product fall-out. An investigation of possibilities and problems

    Energy Technology Data Exchange (ETDEWEB)

    Hovgaard, J.; Korsbech, U.

    1992-12-01

    During 1993 the Danish Emergency Management Agency will install an airborne {gamma}-ray detector system for area survey of contamination with radioactive nuclides - primarily fission products that may be released during a heavy accident at a nuclear power plant or from accidents during transport of radioactive material. The equipment is based on 16 liter NaI(TI) crystals and multichannel analysers from Exploranium (Canada). A preliminary investigation of the possibilities for detection of low and high level contamination - and the problems that may be expected during use of the equipment, and during interpretation of the measured data, is described. Several days after reactor shut-down some of the nuclides can be identified directly from the measured spectrum, and contamination levels may be determined within a factor two. After several weeks, most fission products have decayed. Concentrations and exposure rates can be determined with increasing accuracy as time passes. Approximate calibration of the equipment for measurements of surface contamination and natural radioactivity can be performed in the laboratory. Further checks of equipment should include accurate measurements of the spectrum resolution. Detectors should be checked individually, and all together. Further control of dead time and pulse pile-up should be performed. Energy calibration, electronics performance and data equipment should be tested against results from the original calibration. (AB).

  11. Automated system for selective fission product separations; decays of /sup 113 -115/Pd

    Energy Technology Data Exchange (ETDEWEB)

    Meikrantz, D.H.; Gehrke, R.J.; McIsaac, L.D.; Baker, J.D.; Greenwood, R.C.

    1981-01-01

    A microcomputer controlled radiochemical separation system has been developed for the isolation and study of fission products with half-lives of approx. >= 10 s. The system is based upon solvent extraction with three centrifugal contactors coupled in series, which provides both rapid and highly efficient separations with large decontamination factors. This automated system was utilized to study the radioactive decays of /sup 113 -115/Pd via solvent extraction of the Pd-dimethylglyoxime complex from /sup 252/Cf fission products. As a result of this effort, ..gamma..-rays associated with the decay of approx. equal to 90-s sup(113,113m)Pd, 149-s /sup 114/Pd and 47-s /sup 115/Pd have been identified. The isotopic assignments to each of these Pd radioactivities have been confirmed from observation of the growth and decay curves of their respective Ag daughters. In addition, previously unreported Ag ..gamma..-rays have been assigned; one to the decay of 69-s /sup 113/Ag, and two to the decay of 19-s /sup 115/Ag.

  12. Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

    International Nuclear Information System (INIS)

    Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion

  13. Development of separation process for transuranium elements and some fission products using new extractants and adsorbents

    International Nuclear Information System (INIS)

    Separation process for transuranium elements (TRU = Am, Cm, Np and Pu) and some fission products (Sr, Cs and Mo) has been developed at Japan Atomic Energy Agency using new innovative extractants and adsorbents to improve the partitioning process from the viewpoints of the economy and the reduction of secondary wastes. Phosphorus-free compounds consisting of carbon, hydrogen, oxygen and nitrogen (CHON principle) were applied to the separation steps for TRU, Cs and Sr by using solvent extraction or extraction chromatography. At the first step, TRU and rare-earth elements (RE) are recovered from high-level liquid waste by solvent extraction with N,N,N',N'-tetra-dodecyl-diglycolamide (TDdDGA). Trivalent actinides Am and Cm, are separated from RE at the next step by extraction chromatography using N,N'-dioctyl-N,N'- diphenyl-pyridine-2,6-dicarboxy-amide (Oct-PDA). Heat-generating fission products Cs and Sr are separated from the raffinate of the TDdDGA extraction step by extraction chromatography using calix-crown derivatives for Cs and crown ether derivatives for Sr, sequentially. Finally, Mo is separated by adsorption with an iron oxide adsorbent. This paper presents research and development results concerning the separation process. (authors)

  14. Thermochemical Study on the Sulfurization of Fission Products in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    The thermodynamic behavior of the sulfurization of Nd, and Eu element, which are contained in spent nuclear fuel as fission products was investigated through collection and properties analysis of thermodynamic data in sulfurization of uranium oxides, thermodynamic properties analysis for the oxidation and reduction of fission products, and test and analysis for sulfurization characteristics of Nd and Eu oxide. And also, analysis on thermodynamic data, such as M-O-S phase stability diagram and changes of Gibbs free energy for sulfurization of uranium and Nd2O3 and Eu2O3 were carried out. Nd2O3 and Eu2O3 are sulfurized into Nd2O2S and Eu2O2S or NdySx and EuySx at a range of 400 to 450 .deg. C, while uranium oxides, such as UO2 and U3O8 remain unreacted up to 450 .deg. C Formation of UOS at 500 .deg. C is initiated by sulfurization of uranium oxides. Hence, reaction temperature for the sulfurization of the Nd2O3 and Eu2O3 was selected as a 450 .deg. C

  15. Separation of cesium-137 from uranium fission products via a NeoflonR column supporting tetraphenylboron

    International Nuclear Information System (INIS)

    Cesium is a member of the Group I alkali metals, very reactive earth metals that react vigorously with both air and water. The chemistry of cesium is much like the chemistry of neighboring elements on the periodic table, potassium and rubidium. This close relation creates many problems in plant-life exposed to cesium because it is so easily confused for potassium, an essential nutrient to plants. Radioactive 134Cs and 137Cs are also chemically akin to potassium and stable cesium. Uptake of these radioactive isotopes from groundwater by plant-life destroys the plant-life and can potentially expose humans to the radioactive affects of 134Cs and 137Cs. Much experimental work has been focused on the separation of 137Cs from uranium fission products. In previous experimental work performed a column consisting of Kel-F supporting tetraphenylboron (TPB) was utilized to separate 137Cs from uranium fission products. It is of interest at this time to attempt the separation of 134Cs from 0.01M EDTA using the same method and Neoflon in the place of Kel-F as the inert support. The results of this experiment give a separation efficiency of 88% and show a linear relationship between the column bed length and the separation efficiency obtained. (author)

  16. Energy from nuclear fission(*

    Directory of Open Access Journals (Sweden)

    Ripani M.

    2015-01-01

    Full Text Available The main features of nuclear fission as physical phenomenon will be revisited, emphasizing its peculiarities with respect to other nuclear reactions. Some basic concepts underlying the operation of nuclear reactors and the main types of reactors will be illustrated, including fast reactors, showing the most important differences among them. The nuclear cycle and radioactive-nuclear-waste production will be also discussed, along with the perspectives offered by next generation nuclear assemblies being proposed. The current situation of nuclear power in the world, its role in reducing carbon emission and the available resources will be briefly illustrated.

  17. Energy from nuclear fission()

    Science.gov (United States)

    Ripani, M.

    2015-08-01

    The main features of nuclear fission as physical phenomenon will be revisited, emphasizing its peculiarities with respect to other nuclear reactions. Some basic concepts underlying the operation of nuclear reactors and the main types of reactors will be illustrated, including fast reactors, showing the most important differences among them. The nuclear cycle and radioactive-nuclear-waste production will be also discussed, along with the perspectives offered by next generation nuclear assemblies being proposed. The current situation of nuclear power in the world, its role in reducing carbon emission and the available resources will be briefly illustrated.

  18. Computation of fission product distribution in core and primary circuit of a high temperature reactor during normal operation

    International Nuclear Information System (INIS)

    The fission product release during normal operation from the core of a high temperature reactor is well known to be very low. A HTR-Modul-reactor with a reduced power of 170 MWth is examined under the aspect whether the contamination with Cs-137 as most important nuclide will be so low that a helium turbine in the primary circuit is possible. The program SPTRAN is the tool for the computations and siumlations of fission product transport in HTRs. The program initially developed for computations of accident events has been enlarged for computing the fission product transport under the conditions of normal operation. The theoretical basis, the used programs and data basis are presented followed by the results of the computations. These results are explained and discussed; moreover the consequences and future possibilities of development are shown. (orig./HP)

  19. The Fission-Based  99Mo Production Process ROMOL-99 and Its Application to PINSTECH Islamabad

    Directory of Open Access Journals (Sweden)

    Rudolf Muenze

    2013-01-01

    Full Text Available An innovative process for fission based 99Mo production has been developed under Isotope Technologies Dresden (ITD GmbH (former Hans Wälischmiller GmbH (HWM, Branch Office Dresden, and its functionality has been tested and proved at the Pakistan Institute of Nuclear Science and Technology (PINSTECH, Islamabad. Targets made from uranium aluminum alloy clad with aluminum were irradiated in the core of Pakistan Research Reactor-1 (PARR-1. In the mean time more than 50 batches of fission molybdenum-99 (99Mo have been produced meeting the international purity/pharmacopoeia specifications using this ROMOL-99 process. The process is based on alkaline dissolution of the neutron irradiated targets in presence of NaNO3, chemically extracting the 99Mo from various fission products and purifying the product by column chromatography. This ROMOL-99 process will be described in some detail.

  20. The Outlook for Some Fission Products Utilization with the Aim to Immobilize Long-Lived Radionuclides

    International Nuclear Information System (INIS)

    The prospects for development of nuclear power are intimately associated with solving the problem of safe management and removal from the biosphere of generated radioactive wastes. The most suitable material for fission products and actinides immobilization is the crystalline ceramics. By now numerous literature data are available concerning the synthesis of a large range of various materials with zirconium-based products. It worth mentioning that zirconium is only one of fission products accumulated in the fuel in large amounts. The development of new materials intended for HLW immobilization will allow increasing of radionuclides concentration in solidified product so providing costs reduction at the stage of subsequent storage. At the same time the idea to use for synthesis of compounds, suitable as materials for long-term storage or final disposal of rad-wastes some fission products occurring in spent fuel in considerable amount and capable to form insoluble substances seems to be rather attractive. In authors opinion in the nearest future one can expect the occurrence of publications proposing the techniques allowing the use of 'reactor's zirconium, molybdenum or, perhaps, technetium as well, with the aim of preparing materials suitable for long-lived radionuclides storage or final disposal. The other element, which is generated in the reactor and worth mentioning, is palladium. The prospects for using palladium are defined not only by its higher generation in the reactor, but by a number of its chemical properties as well. It is evident that the use of natural palladium with the purpose of radionuclides immobilization is impossible due to its high cost and deficiency). In author's opinion such materials could be used as targets for long-lived radionuclides transmutation as well. The object of present work was the study on methods that could allow to use 'reactor' palladium with the aim of long-lived radionuclides such as I-129 and TUE immobilization. In the

  1. New antineutrino energy spectra predictions from the summation of beta decay branches of the fission products.

    Science.gov (United States)

    Fallot, M; Cormon, S; Estienne, M; Algora, A; Bui, V M; Cucoanes, A; Elnimr, M; Giot, L; Jordan, D; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Taín, J L; Yermia, F; Zakari-Issoufou, A-A

    2012-11-16

    In this Letter, we study the impact of the inclusion of the recently measured beta decay properties of the (102;104;105;106;107)Tc, (105)Mo, and (101)Nb nuclei in an updated calculation of the antineutrino energy spectra of the four fissible isotopes (235,238)U and (239,241)Pu. These actinides are the main contributors to the fission processes in pressurized water reactors. The beta feeding probabilities of the above-mentioned Tc, Mo, and Nb isotopes have been found to play a major role in the γ component of the decay heat of (239)Pu, solving a large part of the γ discrepancy in the 4-3000 s range. They have been measured by using the total absorption technique, insensitive to the pandemonium effect. The calculations are performed by using the information available nowadays in the nuclear databases, summing all the contributions of the beta decay branches of the fission products. Our results provide a new prediction of the antineutrino energy spectra of (235)U, (239,241)Pu, and, in particular, (238)U for which no measurement has been published yet. We conclude that new total absorption technique measurements are mandatory to improve the reliability of the predicted spectra.

  2. A Research Program for Fission Product/Dust Transport in HTGR’s

    Energy Technology Data Exchange (ETDEWEB)

    Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States)

    2016-02-01

    High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuel was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.

  3. New antineutrino energy spectra predictions from the summation of beta decay branches of the fission products

    CERN Document Server

    Fallot, M; Estienne, M; Algora, A; Bui, V M; Cucoanes, A; Elnimr, M; Giot, L; Jordan, D; Martino, J; Onillon, A; Porta, A; Pronost, G; Taín, J L; Yermia, F; Zakari-Issoufou, A -A

    2012-01-01

    In this paper, we study the impact of the inclusion of the recently measured beta decay properties of the $^{102;104;105;106;107}$Tc, $^{105}$Mo, and $^{101}$Nb nuclei in an updated calculation of the antineutrino energy spectra of the four fissible isotopes $^{235, 238}$U, and $^{239,241}$Pu. These actinides are the main contributors to the fission processes in Pressurized Water Reactors. The beta feeding probabilities of the above-mentioned Tc, Mo and Nb isotopes have been found to play a major role in the $\\gamma$ component of the decay heat of $^{239}$Pu, solving a large part of the $\\gamma$ discrepancy in the 4 to 3000\\,s range. They have been measured using the Total Absorption Technique (TAS), avoiding the Pandemonium effect. The calculations are performed using the information available nowadays in the nuclear databases, summing all the contributions of the beta decay branches of the fission products. Our results provide a new prediction of the antineutrino energy spectra of $^{235}$U, $^{239,241}$Pu ...

  4. Fission-product releases to the primary system of EBR-II from April 1975 to March 1977

    Energy Technology Data Exchange (ETDEWEB)

    So, B Y.C.; Lambert, J D.B.; Kirn, F S; Armstrong, J R; Ebersole, E R; Laug, M T

    1979-05-01

    This report describes the 14 releases of fission products that occurred in EBR-II from April 1975 to March 1977. Each release was readily detected, and all but one (in a driver-fuel subassembly) was identified with a particular subassembly. Xenon tagging was the primary method of identification, although other methods were used where appropriate. Methods of monitoring and identifying fission-product sources are discussed, and each release and identification is described. Effects of breached elements on plant availability were minimal in this period. From all evidence, cladding breaching on elements in EBR-II continues to be a benign process.

  5. Fission-product releases to the primary system of EBR-II from April 1977 to May 1978

    Energy Technology Data Exchange (ETDEWEB)

    So, B Y.C.; Gross, K C; Lambert, J D.B.; Kirn, F S; Ebersole, E R; Laug, M T

    1979-07-01

    Suspected fission-product releases from 18 subassemblies between April 1977 and May 1978 are described. Postirradiation examinations on 15 of the suspect subassemblies confirmed that all contained one or more breached elements. Except for two untagged subassemblies, xenon tagging was the primary method of identification, although other methods were used where appropriate. Methods to monitor and identify fission-product sources are discussed. Problems encountered with multiple-element breaches are described. Overall, the effects of breached elements on plant availability were minimal during this reporting period. From all evidence, cladding breaching on elements in EBR-II continues to be a benign process.

  6. Fission product release from UO/sub 2/ under LWR accident conditions: recent data compared with review values

    International Nuclear Information System (INIS)

    Studies of fission product release from commercial LWR fuel at temperatures up to 20000C in steam have shown that >50% of the Kr, I, and Cs may be released within 20 min. These data are in fairly good agreement with the results of a previous NRC review, but the influences of specific test/accident conditions other than temperature (which is the most important variable) on the behavior of these and other fission products are apparent. In particular, chemical effects related to the extent of cladding oxidation may dominate, as in the case of tellurium. 14 refs., 5 figs., 2 tabs

  7. Fission product chemistry in severe nuclear reactor accidents, specialists' meeting at JRC-Ispra, 15-17 January 1990

    International Nuclear Information System (INIS)

    A specialists' meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions). (author)

  8. Fission product release and microstructure changes during laboratory annealing of a very high burn-up fuel specimen

    Science.gov (United States)

    Hiernaut, J.-P.; Wiss, T.; Colle, J.-Y.; Thiele, H.; Walker, C. T.; Goll, W.; Konings, R. J. M.

    2008-07-01

    A commercial PWR fuel sample with a local burn-up of about 240 MWd/kgHM was annealed in a Knudsen cell mass spectrometer system with a heating rate of 10 K/min up to 2750 K at which temperature the sample was completely vaporized. The release of fission gases and fission products was studied as a function of temperature. In one of the runs the heating was interrupted successively at 900, 1500 and 1860 K and at each step a small fragment of the sample was examined by SEM and analysed by energy dispersive electron probe microanalysis. The release behaviour of volatile, gaseous and other less volatile fission products is presented and analysed with the EFFUS program and related to the structural changes of the fuel.

  9. Cavity Ring-Down Spectroscopy for Gaseous Fission Products Trace Measurements in Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Safety and availability are key issues of the generation IV reactors. Hence, the three radionuclide confinement barriers, including fuel cladding, must stay tight during the reactor operation. During the primary gaseous failure, fission products xenon and krypton are released. Their fast and sensitive detection guarantees the first confinement barrier tightness. In the frame of the French ASTRID project, an optical spectroscopy technique - Cavity Ring Down Spectroscopy (CRDS) - is investigated for the gaseous fission products measurement. A dedicated CRDS set-up is needed to detect the rare gases with a commercial laser. Indeed, the CRDS is coupled to a glow discharge plasma, which generates a population of metastable atoms. The xenon plasma conditions are optimized to 110 Pa and 1.3 W (3 mA). The production efficiency of metastable Xe is then 0.8 %, stable within 0.5% during hours. The metastable number density is proportional to the xenon over argon molar fraction. The spectroscopic parameters of the strong 823.16 nm xenon transition are calculated and/or measured in order to optimize the fit of the experimental spectra and make a quantitative measurement of the metastable xenon. The CRDS is coupled to the discharge cell. The laser intensity inside the cavity is limited by the optical saturation process, resulting from the strong optical pumping of the metastable state. The resulting weak CRDS signal requires a fast and very sensitive photodetector. A 600 ppt xenon molar fraction was measured by CRDS. With the present set-up, the detection limits are estimated from the baseline noise to approximately 20 ppt for each even isotope, 60 ppt for the 131Xe and 55 ppt for the 129Xe. This sensitivity matches the specifications required for gaseous leak measurement; approximately 100 ppt for 133Xe (4 GBq/m3) and 10 ppb for stable isotopes. The odd isotopes are selectively measured, whereas the even isotopes overlap, a spectroscopic feature that applies for stable or

  10. Laboratory-Scale Bismuth Phosphate Extraction Process Simulation To Track Fate of Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Serne, R. JEFFREY; Lindberg, Michael J.; Jones, Thomas E.; Schaef, Herbert T.; Krupka, Kenneth M.

    2007-02-28

    Recent field investigation that collected and characterized vadose zone sediments from beneath inactive liquid disposal facilities at the Hanford 200 Areas show lower than expected concentrations of a long-term risk driver, Tc-99. Therefore laboratory studies were performed to re-create one of the three processes that were used to separate the plutonium from spent fuel and that created most of the wastes disposed or currently stored in tanks at Hanford. The laboratory simulations were used to compare with current estimates based mainly on flow sheet estimates and spotty historical data. Three simulations of the bismuth phosphate precipitation process show that less that 1% of the Tc-99, Cs-135/137, Sr-90, I-129 carry down with the Pu product and thus these isotopes should have remained within the metals waste streams that after neutralization were sent to single shell tanks. Conversely, these isotopes should not be expected to be found in the first and subsequent cycle waste streams that went to cribs. Measurable quantities (~20 to 30%) of the lanthanides, yttrium, and trivalent actinides (Am and Cm) do precipitate with the Pu product, which is higher than the 10% estimate made for current inventory projections. Surprisingly, Se (added as selenate form) also shows about 10% association with the Pu/bismuth phosphate solids. We speculate that the incorporation of some Se into the bismuth phosphate precipitate is caused by selenate substitution into crystal lattice sites for the phosphate. The bulk of the U daughter product Th-234 and Np-237 daughter product Pa-233 also associate with the solids. We suspect that the Pa daughter products of U (Pa-234 and Pa-231) would also co-precipitate with the bismuth phosphate induced solids. No more than 1 % of the Sr-90 and Sb-125 should carry down with the Pu product that ultimately was purified. Thus the current scheme used to estimate where fission products end up being disposed overestimates by one order of magnitude the

  11. Evaluation of six decontamination processes on actinide and fission product contamination

    International Nuclear Information System (INIS)

    In-situ decontamination technologies were evaluated for their ability to: (1) reduce equipment contamination levels to allow either free release of the equipment or land disposal, (2) minimize residues generated by decontamination, and (3) generate residues that are compatible with existing disposal technologies. Six decontamination processes were selected. tested and compared to 4M nitric acid, a traditional decontamination agent: fluoroboric acid (HBF4), nitric plus hydrofluoric acid, alkaline persulfate followed by citric acid plus oxalic acid, silver(II) plus sodium persulfate plus nitric acid, oxalic acid plus hydrogen peroxide plus hydrofluoric acid, and electropolishing using nitric acid electrolyte. The effectiveness of these solutions was tested using prepared 304 stainless steel couponds contaminated with uranium, plutonium, americium, or fission products. The decontamination factor for each of the solutions and tests conditions were determined; the results of these experiments are presented

  12. The decay heat of fission products and actinides of the SNR-300

    International Nuclear Information System (INIS)

    The report describes the computer code RASPA, which calculates the build-up and decay of fission products and actinides. The verification of the code and its library has been performed by comparison with theoretical and experimental results of other authors, whereby a good agreement has been achieved. Furthermore, an error analysis has shown, that the error of the calculated decay heat, which is induced by uncertainties of nuclear data, is less than 10 % up to decay times of one month. The results of calculations of the time dependent decay heat and the gamma source strength in various zones of the cores Mark-Ia and Mark-II of the SNR-300 are documented and discussed in detail

  13. PADLOC: a one-dimensional computer program for calculating coolant and plateout fission product concentrations. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Hudritsch, W.W.; Smith, P.D.

    1977-11-01

    The one-dimensional computer program PADLOC is designed to analyze steady-state and time-dependent plateout of fission products in an arbitrary network of pipes. The problem solved is one of mass transport of impurities in a fluid, including the effects of sources in the fluid and in the plateout surfaces, convection along the flow paths, decay, adsorption on surfaces (plateout), and desorption from surfaces. These phenomena are governed by a system of coupled, nonlinear partial differential equations. The solution is achieved by (a) linearizing the equations about an approximate solution, employing a Newton Raphson iteration technique, (b) employing a finite difference solution method with an implicit time integration, and (c) employing a substructuring technique to logically organize the systems of equations for an arbitrary flow network.

  14. Fuel behaviour and fission product release in the Power Pulse 1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, R.S.; Belov, A.I.; Gauthier, M.D.; Peplinskie, R.T.; Buchanan, C.A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2010-07-01

    The Power Pulse 1 experiment was intended to determine fuel behaviour and fission-product releases under conditions approaching and above the upper limit of fuel behaviour in Loss-of-Coolant Accident (LOCA) power pulses (formation of molten material). Eleven tests were performed on sections cut from two CANDU power reactor fuel elements: one in typical discharged condition, the other after operation in defected condition. The samples were pre-heated in a slow flow of helium before applying direct electric heating; some samples were subjected to power pulses, while others were not. Formation of molten material and columnar grains, limited loss of grain cohesion, and release of {sup 85}Kr were observed in these tests. (author)

  15. Cherenkov light detection as a velocity selector for uranium fission products at intermediate energies

    Science.gov (United States)

    Yamaguchi, T.; Enomoto, A.; Kouno, J.; Yamaki, S.; Matsunaga, S.; Suzaki, F.; Suzuki, T.; Abe, Y.; Nagae, D.; Okada, S.; Ozawa, A.; Saito, Y.; Sawahata, K.; Kitagawa, A.; Sato, S.

    2014-12-01

    The in-flight particle separation capability of intermediate-energy radioactive ion (RI) beams produced at a fragment separator can be improved with the Cherenkov light detection technique. The cone angle of Cherenkov light emission varies as a function of beam velocity. This can be exploited as a velocity selector for secondary beams. Using heavy ion beams available at the HIMAC synchrotron facility, the Cherenkov light angular distribution was measured for several thin radiators with high refractive indices (n = 1.9 ~ 2.1). A velocity resolution of ~10-3 was achieved for a 56Fe beam with an energy of 500 MeV/nucleon. Combined with the conventional rigidity selection technique coupled with energy-loss analysis, the present method will enable the efficient selection of an exotic species from huge amounts of various nuclides, such as uranium fission products at the BigRIPS fragment separator located at the RI Beam Factory.

  16. LMFBR safety: Task 10 - characterization of sodium fires and fission product

    International Nuclear Information System (INIS)

    The objectives of this project are to: develop a computer program for calculating two-dimensional, transient, natural convection phenomena such as those arising from various sodium spill accidents in Liquid Metal Fast Breeder Reactor (LMFBR) heat transfer equipment vaults, head compartments, containment buildings, and secondary heat transfer systems; develop experimental programs and conduct tests that will characterize the behavior of sodium, sodium oxide, fuel, fission product, and other aerosols as they might be generated by various postulated LMFBR accidents; determine by analysis and experiment the generation and transport of these aerosols; and determine the effects of an accident in an LMFBR involving fuel melting by contacting molten UO2 (a fuel simulant) with stainless steel, sodium, concrete, and various sacrificial materials

  17. Fuel behaviour and fission product release in the Power Pulse 1 experiment

    International Nuclear Information System (INIS)

    The Power Pulse 1 experiment was intended to determine fuel behaviour and fission-product releases under conditions approaching and above the upper limit of fuel behaviour in Loss-of-Coolant Accident (LOCA) power pulses (formation of molten material). Eleven tests were performed on sections cut from two CANDU power reactor fuel elements: one in typical discharged condition, the other after operation in defected condition. The samples were pre-heated in a slow flow of helium before applying direct electric heating; some samples were subjected to power pulses, while others were not. Formation of molten material and columnar grains, limited loss of grain cohesion, and release of 85Kr were observed in these tests. (author)

  18. Technical bases for estimating fission product behavior during LWR accidents. Technical report

    International Nuclear Information System (INIS)

    The objective of this report is to provide the Nuclear Regulatory Commission and the public with a description of the best technical information currently available for estimating the release of radioactive material during postulated reactor accidents, and to identify where gaps exist in our knowledge. This report focuses on those low probability-high consequence accidents involving severe damage to the reactor core and core meltdown that dominate the risk to the public. Furthermore, in this report particular emphasis is placed on the accident behavior of radioactive iodine, as (1) radioiodine is predicted to be a major contributor to public exposure, (2) current regulatory accident analysis procedures focus on iodine, and (3) several technical issues have been raised recently about the magnitude of iodine release. The generation, transport, and attenuation of aerosols were also investigated in some detail to assess their effect on fission product release estimates and to determine the performance of engineered safety features under accident conditions exceeding their design bases

  19. THAI experiments on hydrogen and fission product behavior in the LWR containment during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, Sanjeev [Becker Technologies GmbH, Berlin (Germany)

    2012-03-15

    In case of a severe accident in a nuclear reactor, a large amount of hydrogen and fission products may be generated by interactions of the core melt with containment structures and water. The behaviour and the distribution of hydrogen and the fission products can be affected by a variety of chemical and physical phenomena taking place inside the containment. These phenomena involve the disciplines of thermal hydraulics, hydrogen distribution and deflagrations, fission products chemistry and material interactions, aerosol physics, and effectiveness of mitigation measures among others. Predictions of the consequences of a severe accident by conventional Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes need to be based on large scale coupled-phenomenon experiments to minimize the scale effect in extrapolation to reactor safety analysis purpose. In this context, an extensive experimental program on nuclear severe accident has been pursuing at the THAI (Thermal-hydraulics, Hydrogen, Aerosol, and Iodine) test facility for many years. Main component of the facility is a 60-m{sup 3} stainless steel vessel, 9.2m high and 3.2m in diameter, with exchangeable internals for multi-compartment investigations. The test facility is operated by Becker Technologies under sponsorship of the German Federal Ministry of Economics and Technology. Since its construction in 2000, sponsorship of the German Federal Ministry of Economics and Technology. Since its construction in 2000, THAI facility has been engaged in the field of reactor safety in the frame of various national (THAI-I to THAI-IV) and international programs (OECD-THAI and THAI2 projects). Additionally, experimental data has been provided in the frame of several International Standard Problems (ISP 41, 46, 47, and 49) for code validation exercises. The THAI test facility allows investigating various accident scenarios, ranging from turbulent free convection to stagnant stratified containment atmospheres, and

  20. Ab initio modelling of the behaviour of point defects and fission products in nuclear fuel

    International Nuclear Information System (INIS)

    The aim of this work is to determine precisely the mechanisms of formation and migration of defects and fission products as well as the associated energies. Examples on uranium dioxide UO2 (standard nuclear fuel) and on uranium carbide UC (potential fuel for new generation reactors) are given. The obtained results are discussed and compared with the experimental results carried out. The ab initio method used is the Projector Augmented-Wave (PAW) method based on the density functional theory. The particular electronic properties of actinides are especially studied because, on account of their 5f orbitals more or less localized around the nucleus, it is difficult to model the actinide compounds by the DFT method. In particular, the modelling of the exchange-correlation interaction of the 5f electrons of UO2 requires approximations (as GGA+U) beyond those more currently used in ab initio calculations (LDA or GGA). (O.M.)

  1. Basic steps in the production of the African catfish seed

    OpenAIRE

    Ondhoro, C.C.; Mwanja, T.M.

    2014-01-01

    A typical production cycle for African catfish farming begins with a selection of fingerlings or juvenile fish of good quality for brood stock development. Fish are selected from a family or grow out stock basing on records of the origin,age, strain and performance history of the parents or from the wild in this brochure, we explain the basic steps and requirements a farmer needs in order to achieve good results in the hatchery.

  2. Fission product plateout and liftoff in the MHTGR primary system: A review

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P. (Oak Ridge National Lab., TN (USA))

    1991-04-01

    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.

  3. Fission product plateout and liftoff in the MHTGR primary system: A review

    International Nuclear Information System (INIS)

    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

  4. Distribution of fission products in Peach Bottom HTGR fuel element E14-01

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Fairchild, L.L.

    1977-08-01

    The third in a projected series of six postirradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements is presented. Element E14-01, the subject of the report, was one of the 60 driver elements (out of a total of 804) that contained zirconium boride pellets within a hollow spine. It was also one of the few predimensioned elements, which therefore allowed accurate determination of dimensional change due to irradiation service. The element received an equivalent of 897 full-power days irradiation prior to scheduled termination of Core 2 operation. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a /sup 137/Cs inventory of 0.24 Ci in the graphite sleeve and 0.047 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides /sup 134/Cs, /sup 110m/Ag, /sup 60/Co, and /sup 154/Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the beta-emitters /sup 3/H, /sup 14/C, and /sup 90/Sr were obtained at three axial locations within the fueled region of the element. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. In addition to fission product distributions, the appearance of the component parts of the element was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed.

  5. Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

    Energy Technology Data Exchange (ETDEWEB)

    Cols, H.; Cristini, P.; Marques, R.

    1997-08-01

    The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing high enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.

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

    Science.gov (United States)

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

    2016-07-01

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

  7. Development of analytical methods relating to aerosol and fission product release from hot and boiling sodium pools

    International Nuclear Information System (INIS)

    Analytical methods are described for (a) sodium; (b) the following anions of sodium aerosols: OH-, CO2- and HCO3-; (c) fission products Cs and Sr. For sodium, the ion selective electrode was used. The anions were determined by a titration method using phenolphthalein and methyl orange as indicators. Atomic absorption spectroscopy was used for Cs and Sr. (U.K.)

  8. Thermodynamics of fission products in dispersion fuel designs - first principles modeling of defect behavior in bulk and at interfaces

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang-yand [Los Alamos National Laboratory; Uberuaga, Blas P [Los Alamos National Laboratory; Nerikar, Pankaj [Los Alamos National Laboratory; Sickafus, Kurt E [Los Alamos National Laboratory; Stanek, Chris R [Los Alamos National Laboratory

    2009-01-01

    Density functional theory (DFT) calculations of fission product (Xe, Sr, and Cs) incorporation and segregation in alkaline earth metal oxides, HfO{sub 2} and UO{sub 2} oxides, and the MgO/(U, Hf, Ce)O{sub 2} interfaces have been carried out. In the case of UO{sub 2}, the calculations were performed using spin polarization and with a Hubbard U term characterizing the on-sit Coulomb repulsion between the localized 5f electrons. The fission product solution energies in bulk UO{sub 2{+-}x} have been calculated as a function of non-stoichiometry x, and were compared to that in MgO. These calculations demonstrate that the fission product incorporation energies in MgO are higher than in HfO{sub 2}. However, this trend is reversed or reduced for alkaline earth oxides with larger cation sizes. The solution energies of fission products in MgO are substantially higher than in UO{sub 2{+-}x}, except for the case of Sr in the hypostoichiometric case. Due to size effects, the thermodynamic driving force of segregation for Xe and Cs from bulk MgO to the MgO/fluorite interface is strong. However, this driving force is relatively weak for Sr.

  9. Atomic scale mobility of the volatile fission products Xe, Kr and I in cubic SiC.

    Science.gov (United States)

    Cooper, M W D; Kelly, S; Bertolus, M

    2016-06-22

    The migration barriers for the vacancy-assisted migration of fission products in 3C-SiC are reported and analysed in the context of the five frequency model, which enables one to calculate an effective diffusion coefficient from elementary mechanisms. Calculations were carried out using the nudged elastic band method (NEB) with interatomic forces determined from density functional theory (DFT). Justification for treating vacancy-assisted fission product migration as limited to the FCC carbon sublattice is based on the stability of carbon vacancies, unfavourable silicon vacancy formation and the accommodation of fission products on the carbon sublattice. Results show that for most Fermi levels within the band gap the activation energy for I exceeds that of Xe which exceeds that of Kr. Results also indicate that activation energies are higher near the conduction edge, thus, implying that enhanced fission product retention can be achieved through n-type doping of 3C-SiC, which limits the availability of the migration mediating carbon vacancies. PMID:27282287

  10. Substrate preference of citrus naringenin rhamnosyltransferases and their application to flavonoid glycoside production in fission yeast.

    Science.gov (United States)

    Ohashi, Takao; Hasegawa, Yuka; Misaki, Ryo; Fujiyama, Kazuhito

    2016-01-01

    Flavonoids, which comprise a large family of secondary plant metabolites, have received increased attention in recent years due to their wide range of features beneficial to human health. One of the most abundant flavonoid skeletons in citrus species is the flavanone naringenin, which is accumulated as glycosides containing terminal rhamnose (Rha) after serial glycosylation steps. The linkage type of Rha residues is a determining factor in the bitterness of the citrus fruit. Such Rha residues are attached by either an α1,2- or an α1,6-rhamnosyltransferase (1,2RhaT or 1,6RhaT). Although the genes encoding these RhaTs from pummelo (Citrus maxima) and orange (Citrus sinensis) have been functionally characterized, the details of the biochemical characterization, including the substrate preference, remain elusive due to the lack of availability of the UDP-Rha required as substrate. In this study, an efficient UDP-Rha in vivo production system using the engineered fission yeast expressing Arabidopsis thaliana rhamnose synthase 2 (AtRHM2) gene was constructed. The in vitro RhaT assay using the constructed UDP-Rha revealed that recombinant RhaT proteins (Cm1,2RhaT; Cs1,6RhaT; or Cm1,6RhaT), which were heterologously produced in fission yeast, catalyzed the rhamnosyl transfer to naringenin-7-O-glucoside as an acceptor. The substrate preference analysis showed that Cm1,2RhaT had glycosyl transfer activity toward UDP-xylose as well as UDP-Rha. On the other hand, Cs1,6RhaT and Cm1,6RhaT showed rhamnosyltransfer activity toward quercetin-3-O-glucoside in addition to naringenin-7-O-glucoside, indicating weak specificity toward acceptor substrates. Finally, naringin and narirutin from naringenin-7-O-glucoside were produced using the engineered fission yeast expressing the AtRHM2 and the Cm1,2RhaT or the Cs1,6RhaT genes as a whole-cell-biocatalyst. PMID:26433966

  11. Substrate preference of citrus naringenin rhamnosyltransferases and their application to flavonoid glycoside production in fission yeast.

    Science.gov (United States)

    Ohashi, Takao; Hasegawa, Yuka; Misaki, Ryo; Fujiyama, Kazuhito

    2016-01-01

    Flavonoids, which comprise a large family of secondary plant metabolites, have received increased attention in recent years due to their wide range of features beneficial to human health. One of the most abundant flavonoid skeletons in citrus species is the flavanone naringenin, which is accumulated as glycosides containing terminal rhamnose (Rha) after serial glycosylation steps. The linkage type of Rha residues is a determining factor in the bitterness of the citrus fruit. Such Rha residues are attached by either an α1,2- or an α1,6-rhamnosyltransferase (1,2RhaT or 1,6RhaT). Although the genes encoding these RhaTs from pummelo (Citrus maxima) and orange (Citrus sinensis) have been functionally characterized, the details of the biochemical characterization, including the substrate preference, remain elusive due to the lack of availability of the UDP-Rha required as substrate. In this study, an efficient UDP-Rha in vivo production system using the engineered fission yeast expressing Arabidopsis thaliana rhamnose synthase 2 (AtRHM2) gene was constructed. The in vitro RhaT assay using the constructed UDP-Rha revealed that recombinant RhaT proteins (Cm1,2RhaT; Cs1,6RhaT; or Cm1,6RhaT), which were heterologously produced in fission yeast, catalyzed the rhamnosyl transfer to naringenin-7-O-glucoside as an acceptor. The substrate preference analysis showed that Cm1,2RhaT had glycosyl transfer activity toward UDP-xylose as well as UDP-Rha. On the other hand, Cs1,6RhaT and Cm1,6RhaT showed rhamnosyltransfer activity toward quercetin-3-O-glucoside in addition to naringenin-7-O-glucoside, indicating weak specificity toward acceptor substrates. Finally, naringin and narirutin from naringenin-7-O-glucoside were produced using the engineered fission yeast expressing the AtRHM2 and the Cm1,2RhaT or the Cs1,6RhaT genes as a whole-cell-biocatalyst.

  12. Management of radioactive waste from 99Mo production by nuclear fission

    International Nuclear Information System (INIS)

    Brazil intends to build a facility for the 99Mo production through 235U fission, once this radioisotope is largely used in nuclear medicine. This study aimed at estimating the physical, chemical and radiological characteristics of radioactive waste expected to be generated in that facility, and to provide theoretical subsides that can be used on the definition of a proper waste management system. Two production scenarios were established and the radioisotope inventories of the wastes were calculated by Scale®. From the chemical processing of the uranium targets the wastes were characterized on their chemical and radiological features. MicroShield® was used to determine the activity concentrations up to three months of 99Mo production. In addition, this work presents dose rate calculation for several sizes of shielding and different amount of wastes, collected in a proper package for in-site transportation. Radionuclides responsible for higher doses were identified in order to facilitate choosing the most appropriate method for managing the wastes after their chemical separation and before their storage. These results are part of what is expected on radioactive wastes at a 99Mo production facility and might help on the development of the waste management planning for that facility. (author)

  13. Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Postirradiation heating tests of TRISO-coated UO2 particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of 85Kr, 110mAg, 134Cs, 137Cs, and 154Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of 110mAg, 134Cs, 137Cs, and 154Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations

  14. Study of the production of neutron-rich isotope beams issuing from fissions induced by fast neutrons; Etude de la production de faisceaux riches en neutrons par fission induite par neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Lau, Ch

    2000-09-15

    This work is a contribution to the PARRNe project (production of radioactive neutron-rich isotopes). This project is based on the fission fragments coming from the fission of 238-uranium induced by fast neutrons. The fast neutron flux is produced by the collisions of deutons in a converter. Thick targets of uranium carbide and liquid uranium targets have been designed in order to allow a quick release of fission fragments. A device, able to trap on a cryogenic thimble rare gas released by the target, has allowed the production of radioactive nuclei whose half-life is about 1 second. This installation has been settled to different deuton accelerators in the framework of the European collaboration SPIRAL-2. A calibration experiment has proved the feasibility of fixing an ISOL-type isotope separator to a 15 MV tandem accelerator, this installation can provide 500 nA deutons beams whose energy is 26 MeV and be a valuable tool for studying fast-neutron induced fission. Zinc, krypton, rubidium, cadmium, iodine, xenon and cesium beams have been produced in this installation. The most intense beams reach 10000 nuclei by micro-coulomb for 26 MeV deutons. An extra gain of 2 magnitude orders can be obtained by using a more specific ion source and by increasing the thickness of the target. Another extra gain of 2 magnitude orders involves 100 MeV deutons.

  15. Status of the French research program for actinides and fission products partitioning and transmutation

    International Nuclear Information System (INIS)

    currently presented to French Ministries of Research and Industry and to the National Parliament which plans to pass a new waste management law in 2006 asking for new prospects for P and T further implementation. The massive research programme on enhanced separation, conducted by CEA and supported by broad international cooperation, has recently achieved some vital progress. Based on real solutions derived from the La Hague process, the CEA demonstrated in 2001 the lab-scale feasibility of extracting minor actinides and some fission products (I, Cs and Tc) using an hydrometallurgical process. Then, the 2002-2005 program has encompassed technological demonstration of the selected liquid-liquid process, with representative equipment which have been set up for this purpose in new shielded cells inside the Atalante facility. CEA also conducted programmes proving the feasibility of the elimination of minor actinides and fission products by transmutation: fabrication of specific targets and fuels for transmutation test in the HFR and Phenix reactors, neutronics and technology studies for critical reactors and ADS developments. The scenario studies aimed at examining the possibilities of reducing significantly the final waste inventory and at quantifying the inventories of plutonium, minor actinides and certain long-lived fission products in various nuclear-power-plant geometries; they also allowed to verify the feasibility at the level of the cycle facilities and fuels involved in those scenarios. (author)

  16. Development of an evaluation method of fission product release fraction from High Temperature Gas-cooled Reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Minato, Kazuo; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-11-01

    The High Temperature Gas-cooled Reactor (HTGR) uses coated particles as fuel. Current coated particle is a microsphere of fuel kernel with TRISO coatings. The TRISO coatings consist of a low-density, porous pyrolytic carbon (PyC) buffer layer adjacent to the spherical fuel kernel, followed by an inner isotropic PyC layer, a SiC layer and a final (outer) PyC layer. An evaluation method of fission product release behavior during the normal operation was developed. Key issues of fission gas release model were: (1) fission gas releases from matrix contamination uranium and through-coatings failed particle were separately modeled and (2) burnup and fast neutron irradiation effects were newly considered. For metallic fission product, fractional release of cesium from coated fuel particles was investigated by comparing measured data in an irradiation test which contained three kinds of fuel particles; artificially bored particles simulating through-coatings failed particles, as-manufactured SiC-failed particles and intact particles. Through the comparison of measured and calculated fractional releases, an equivalent diffusion coefficient of SiC layer in the SiC-failed particle was introduced. This report describes the developed model together with validation result of the release model. (author)

  17. Determination of long-lived fission products and actinides in Savannah River site HLW sludge and glass for waste acceptance

    International Nuclear Information System (INIS)

    Savannah River Site (SRS) is currently immobilizing the radioactive, caustic, high-level waste sludge in Tank 51 into a borosilicate glass for disposal in a geologic repository. A requirement for repository acceptance is that SRS report the concentrations of certain fission product and actinide radionuclides in the glass. This paper presents measurements of many of these concentrations in both Tank 51 sludge and the final glass. The radionuclides were measured by inductively coupled plasma - mass spectrometry and α, β, and γ counting methods. Examples of the radionuclides are Sr-90, Cs-137, U-238, Pu-239, and Cm-244. Concentrations in the glass are 3.1 times lower due to dilution of the sludge with a nonradioactive glass forming frit in the vitrification process. Results also indicated that in both the sludge and glass the relative concentrations of the long lived fission products insoluble in caustic area in proportion to their yields from the fission of U-235 in the SRS reactors. This allowed the calculation of a fission yield scaling factor. This factor in addition to the sludge dilution factor can be used to estimate concentrations of waste acceptance radionuclides that cannot be measured in the glass

  18. Determination of long-lived fission products and actinides in Savannah River Site HLW sludge and glass for waste acceptance

    International Nuclear Information System (INIS)

    Savannah River Site (SRS) is immobilizing the radioactive, high-level waste sludge in Tank 51 into a borosilicate glass for disposal in a geologic repository. A requirement for repository acceptance is that SRS report the concentrations of certain fission product and actinide radionuclides in the glass. This paper presents measurements of many of these concentrations in both Tank 51 sludge and the final glass. The radionuclides were measured by inductively coupled plasma mass spectrometry and α, β, and γ counting methods. Examples of the radionuclides are 90Sr, 137Cs, 238U and , 239Pu. Concentrations in the glass are 3.1 times lower due to dilution of the sludge with a nonradioactive glass forming frit in the vitrification process. Results also indicated that in both the sludge and glass the relative concentrations of the long lived fission products insoluble in caustic are in proportion to their yields from the fission of 235U waste in the SRS reactors. This allowed the calculation of a fission yield scaling factor. This factor in addition to the sludge dilution factor can be used to estimate concentrations of waste acceptance radionuclides that cannot be measured in the glass. Examples of these radionuclides are 79Se, 93Zr, and 107Pd. (author)

  19. Competition of fusion and quasi-fission in the reactions leading to production of the superheavy elements

    OpenAIRE

    Veselsky, M.

    2003-01-01

    The mechanism of fusion hindrance, an effect observed in the reactions of cold, warm and hot fusion leading to production of the superheavy elements, is investigated. A systematics of transfermium production cross sections is used to determine fusion probabilities. Mechanism of fusion hindrance is described as a competition of fusion and quasi-fission. Available evaporation residue cross sections in the superheavy region are reproduced satisfactorily. Analysis of the measured capture cross se...

  20. Energy Dependence of Neutron-Induced Fission Product Yields of 235U, 238U and 239Pu Between 0.5 and 14.8 MeV

    Science.gov (United States)

    Gooden, Matthew; Tornow, Werner; Tonchev, Anton; Vieira, Dave; Wilhelmy, Jerry; Arnold, Charles; Fowler, Malcolm; Stoyer, Mark

    2014-09-01

    Under a joint collaboration between TUNL-LANL-LLNL, a set of absolute fission product yield measurements have been performed. The energy dependence of a number of cumulative fission products between 0.5 and 14.8 MeV have been measured using quasi-monoenergetic neutron beams for three actinide targets, 235U, 238U and 239Pu, between 0.5 and 14.8 MeV. The FPYs were measured by a combination of activation utilizing specially designed dual-fission chambers and γ-ray counting. The dual-fission chambers are back-to-back ionization chambers encasing a target with thin deposits of the same target isotope in each chamber. This method allows for the direct measurement of the fission rate in the activation target with no reference to the fission cross-section, reducing uncertainties. γ-ray counting was performed on well-shield HPGe detectors over a period of 2 months per activation to properly identify fission products. Reported are absolute cumulative fission product yields for incident neutron energies of 0.5, 1.37, 2.4, 4.6 and 14.8 MeV.

  1. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-09-01

    Safety tests were conducted on fourteen fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800°C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show different trends in the prediction of the fractional release depending on the species, and it leads to different conclusions regarding the diffusivities used in the modeling of fission product transport in TRISO-coated particles: • For silver, the diffusivity in silicon carbide (SiC) might be over-estimated by a factor of at least 102 to 103 at 1600°C and 1700°C, and at least 10 to 102 at 1800°C. The diffusivity of silver in uranium oxy-carbide (UCO) might also be over-estimated, but the available data are insufficient to allow definitive conclusions to be drawn. • For cesium, the diffusivity in UCO might be over-estimated by a factor of at least 102 to 103 at 1600°C, 105 at 1700°C, and 103 at 1800°C. The diffusivity of cesium in SiC might also over-estimated, by a factor of 10 at 1600°C and 103 at 1700°C, based upon the comparisons between calculated and measured release fractions from intact particles. There is no available estimate at 1800°C since all the compacts heated up at 1800°C contain particles with failed SiC layers whose release dominates the release from intact particles. • For strontium, the diffusivity in SiC might be over-estimated by a factor of 10 to 102 at 1600 and 1700°C, and 102 to 103 at 1800°C. These

  2. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  3. Nuclear Data Requirements for the Production of Medical Isotopes in Fission Reactors and Particle Accelerators

    CERN Document Server

    Garland, M A; Talbert, R J; Mashnik, S G; Wilson, W B

    1999-01-01

    Through decades of effort in nuclear data development and simulations of reactor neutronics and accelerator transmutation, a collection of reaction data is continuing to evolve with the potential of direct applications to the production of medical isotopes. At Los Alamos the CINDER'90 code and library have been developed for nuclide inventory calculations using neutron-reaction (En < 20 MeV) and/or decay data for 3400 nuclides; coupled with the LAHET Code System (LCS), irradiations in neutron and proton environments below a few GeV are tractable; additional work with the European Activation File, the HMS-ALICE code and the reaction models of MCNPX (CEM95, BERTINI, or ISABEL with or without preequilibrium, evaporation and fission) have been used to produce evaluated reaction data for neutrons and protons to 1.7 GeV. At the Pacific Northwest National Laboratory, efforts have focused on production of medical isotopes and the identification of available neutron reaction data from results of integral measuremen...

  4. Determination of relative krypton fission product yields from 14 MeV neutron induced fission of 238U at the National Ignition Facility

    Science.gov (United States)

    Edwards, E. R.; Cassata, W. S.; Velsko, C. A.; Yeamans, C. B.; Shaughnessy, D. A.

    2016-11-01

    Precisely-known fission yield distributions are needed to determine a fissioning isotope and the incident neutron energy in nuclear security applications. 14 MeV neutrons from DT fusion at the National Ignition Facility induce fission in depleted uranium contained in the target assembly hohlraum. The fission yields of Kr isotopes (85m, 87, 88, and 89) are measured relative to the cumulative yield of 88Kr and compared to previously tabulated values. The results from this experiment and England and Rider are in agreement, except for the 85mKr/88Kr ratio, which may be the result of incorrect nuclear data.

  5. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDUR 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    International Nuclear Information System (INIS)

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  6. Experimental simulation of irradiation effects on thermomechanical behaviour of UO2 fuel: Impact of solid and gaseous fission products

    International Nuclear Information System (INIS)

    Predictive simulation of thermomechanical behaviour of nuclear fuel has to take into account irradiation effects. Fission Products (FP) can modify the thermomechanical behaviour of UO2. During this thesis, differentiation was made between fission products which create a solid solution with UO2 and gaseous products, generating pressurized bubbles. SIMFUELS containing gadolinium oxide and pressurized argon bubbles were manufactured, respectively by conventional process and by Gas Pressure Sintering. Brittle and ductile behaviour of UO2 was investigated, under experimental conditions representative of Pellet-Cladding Interaction (PCI), respectively with 3 points bending tests and compressive creep tests. Investigation of brittle behaviour of UO2 showed that fracture is mainly controlled by natural defects, like porosities, acting like starting points for cracks propagation. Addition of simulates fission products increase the brittle-to-ductile transition temperature of UO2, up to 400-500 C regarding FP in solid solution, and up to 200 C for gaseous products. Fission products although reduce fracture stresses, by a factor between 1.5 and 4, respectively for gas bubbles and solid solutions. Decrease of fracture stress is linked to an increase of microstructural defects due the solid solution and to pressurized bubbles located at grain boundaries. Pellets were tested under compressive solicitation at high temperatures. Experimental results of creep tests are well represented by Norton laws. Creep controlling mechanisms are evidenced by microstructural analysis performed on pellets at different strains. On the basis of calculations made for fuels having the same microstructures than the SIMFUELs, a creep factor is determined. It revealed a strong hardening effect of the solid solution, due to the fact that the added elements anchor the dislocations, whereas pressurized bubbles showed a coupling between hardening and softening effects. (author)

  7. Investigations of the mass and charge distribution of fission products from the 238U(n14,f) reaction by direct Ge(Li) method

    International Nuclear Information System (INIS)

    The fission yields can be measured by the well-known activation method if it is taken into account that the fission process results in 5-6 nuclides in an isobaric chain. The method which is based only on the gamma-spectrometric measurement of the irradiated fissioning sample is referred to as the direct Ge(Li) method for fission yield measurement. The thesis contains detailed description of the direct Ge(Li) method. The method was tested by the measurement of cumulative yields of 47 fission products and independent yields of 7 products in the reaction of 238U(n14,f). These are the members of 37 mass chains in the A=83-149 mass number region. The half-lives of the studied products are in the range of Tsub(1/2)=102-109 s; the gamma spectrometric method was improved by extending its applicability to the measurement of short-lived products. Applying short irradiation time (5 min) the yields of 16 fission products with half-lives shorter than 1 hour could be measured. The lowest measured partial fission cross sections (yields) are in the order of 1 mb (0.1%). The accuracy of the yield measured by the direct Ge(Li) method is as high as or higher than that obtained radiochemically, especially for the products measured by many intensive gamma lines. (author)

  8. Primary system fission product release and transport. A state-of-the-art report to the committee on the safety of nuclear installations

    International Nuclear Information System (INIS)

    This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena and presents major conclusions on the state of the art

  9. Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

    International Nuclear Information System (INIS)

    This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena and presents major conclusions on the state of the art

  10. Evaporation release behavior of volatile fission products from liquid sodium pool to the inert cover gas

    Energy Technology Data Exchange (ETDEWEB)

    Nakagiri, T.; Miyahara, S. [Oarai Engineering Center, Power Reactor and Nuclear Fuel Development Corp., Oaraimachi, Ibaraki (Japan)

    1996-12-01

    In fuel failure of sodium cooled fast breeder reactors, released volatile fission products (VFPs) such as iodine and cesium from the fuel will be dissolved into the liquid sodium coolant and transferred to the cover vaporization. In the cover gas system of the reactor, natural convection occurs due to temperature differences between the sodium pool and the gas phase. The release rates of VFPs together with sodium vaporization are considered to be controlled by the convection. In this study, three analytical models are developed and examined to calculate the transient release rates using the equilibrium partition coefficients of VFPs. The calculated release rates are compared with experimental results for sodium and sodium iodide. The release rate of sodium is closest to the calculation by the heterogeneous nucleation theory. The release rate of sodium iodide obtained from the experiment is between the release rates calculated by the model based on heat-and-mass transfer analogy and the Hill`s theory. From this study, it is confirmed that the realistic release rate of sodium is able to be calculated by the model based on the heterogeneous nucleation theory. The conservative release rate of sodium iodide is able to be calculated by the model based on the Hill`s theory using the equilibrium partition coefficient of sodium iodide. (author) 7 figs., 1 tab., 3 refs.

  11. Fast-neutron interaction with the fission product {sup 103}Rh

    Energy Technology Data Exchange (ETDEWEB)

    Smith, A.B. [Argonne National Lab., IL (United States)]|[Arizona Univ., Tucson, AZ (United States); Guenther, P.T. [Argonne National Lab., IL (United States)

    1993-09-01

    Neutron total and differential elastic- and inelastic-scattering cross sections of {sup 103}Rh are measured from {approximately} 0.7 to 4.5 MeV (totals) and from {approximately} 1.5 to 10 MeV (scattering) with sufficient detail to define the energy-averaged behavior of the neutron processes. Neutrons corresponding to excitations of groups of levels at 334 {plus_minus} 13, 536 {plus_minus} 10, 648 {plus_minus} 25, 796 {plus_minus} 20, 864 {plus_minus} 22, 1120 {plus_minus} 22, 1279 {plus_minus} 60, 1481 {plus_minus} 27 and 1683 {plus_minus} 39 keV were observed. Additional groups at 1840 {plus_minus} 79 and 1991 {plus_minus} 71 key were tentatively identified. Assuming the target is a collective nucleus reasonably approximated by a simple one-phonon vibrator, spherical-optical, dispersive-optical, and coupled-channels models were developed from the data base with attention to the parameterization of the large inelastic-scattering cross sections. The physical properties of these models are compared with theoretical predictions and the systematics of similar model parameterizations in this mass region. In particular, it is shown that the inelastic-scattering cross section of the {sup 103}Rh fission product is large at the relatively low energies of applied interest.

  12. Separation of the rare-earth fission product poisons from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  13. Gas emission from the UO2 samples, containing fission products and burnable absorber

    Science.gov (United States)

    Kopytin, V. P.; Baranov, V. G.; Burlakova, M. A.; Tenishev, A. V.; Kuzmin, R. S.; Pokrovskiy, S. A.; Mikhalchik, V. V.

    2016-04-01

    The process gas released from the fuel pellets of uranium fuel during fuel burn-up reduces the thermal conductivity of the rod-shell gap, enhances hydrogen embrittlement of the cladding material, causes it's carbonization, as well as transport processes in the fuel. In this study a technique of investigating the thermal desorption of gases from the UO2 fuel material were perfected in the temperature range 300-2000 K for uniform sample heating rate of 15 K/min in vacuum. The characteristic kinetic dependences are acquired for the gas emission from UO2 samples, containing simulators of fission products (SFP) and the burnable neutron absorber (BNA). Depending on the amount of SFP and BNA contained in the sample thermal desorption gas spectra (TDGS) vary. The composition of emitted gas varies, as well as the number of peaks in the TDGS and the peaks shift to higher temperatures. This indicates that introduction of SFPs and BNA alters the sample material structure and cause the creation of so- called traps which have different bonding energies to the gases. The traps can be a grid of dislocations, voids, and contained in the UO2 matrix SFP and BNA. Similar processes will occur in the fuel pellets in the real conditions of the Nuclear Power Plant as well.

  14. Water reactor fuel behaviour and fission products release in off-normal and accident conditions

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held at the IAEA Headquarters in Vienna from 10 to 13 November 1986. Thirty participants from 17 countries and an international organization attended the meeting. Eighteen papers were presented from 13 countries and one international organization. The meeting was composed of four sessions and covered subjects related to: physico-chemical properties of core materials under off-normal conditions, and their interactions up to and after melt-down (5 papers); core materials deformation, relocation and core coolability under (severe) accident conditions (4 papers); fission products release: including experience, mechanisms and modelling (5 papers); power plant experience (4 papers). A separate abstract was prepared for each of these 18 papers. Four working groups covering the above-mentioned topics were held to discuss the present status of the knowledge and to develop recommendations for future activities in this field. Refs, figs and tabs

  15. Modeling of constituent redistribution and fission product migration in fast reactor U-Pu-Zr fuel

    International Nuclear Information System (INIS)

    Radial constituent redistribution in a fast reactor U-Pu-Zr fuel is an important phenomenon that occurs because the fuel alloy has thermal gradients and poly-phase fields at the typical operation temperature. In a typical temperature range (500-700degC), Zr moves from the mid-radius region to the fuel center region and the fuel surface region. Because of this phenomenon, the homogeneous fuel at beginning of life turns into locally heterogeneous fuel. Most of the thermophysical properties change locally, as does fuel performance. Fuel constituent redistribution of U-Pu-Zr is modeled by treating Pu as sessile element and therefore by assuming a pseudo-binary system. Fission product lanthanides (LA) migrate to the fuel surface. LA migration appears to be due both to direct vapor transport and diffusion through the fuel matrix. Large pores at the low Zr zone and fuel periphery may support for LA precipitates. LA diffusion through Pu also contributes to observed LA migration. Because Pu is relatively sessile, however, LA migration by diffusion through the fuel matrix is relatively small. Upon migration to the fuel surface, LA and Pu react with Fe-base alloy cladding such as HT9 and D9 whereas U and Zr do not. The LA and Pu reaction with cladding is via interdiffusion. (author)

  16. Fission product release profiles from spherical HTR fuel elements at accident temperatures

    International Nuclear Information System (INIS)

    A total of 22 fuel elements with modern TRISO particles has been tested in the temperature range 1500-25000C. Additionally, release profiles of iodine and other isotopes have been obtained with seven UO2 samples at 1400-18000C. For heating times up to 100 hours at the maximum temperature, the following results are pertinent to HTR accident conditions: Ag 110 m is the only fission products to be released at 1200-16000C by diffusion through intact SiC, but it is of low significance in accident assessments; cesium, iodine, strontium, and noble gas releases up to 16000C are solely due to various forms of contamination; at 1700-18000C, corrosion induced SiC defects cause the release of Cs, Sr, I/Xe/Kr; above 20000C, thermal decomposition of the silicon carbide layer sets in while pyrocarbons still remain intact. Around 16000C, the accident specific contribution of cesium, strontium, iodine, and noble gases is negligible. (orig./HP)

  17. TRAFIC, a computer program for calculating the release of metallic fission products from an HTGR core

    Energy Technology Data Exchange (ETDEWEB)

    Smith, P.D.

    1978-02-01

    A special purpose computer program, TRAFIC, is presented for calculating the release of metallic fission products from an HTGR core. The program is based upon Fick's law of diffusion for radioactive species. One-dimensional transient diffusion calculations are performed for the coated fuel particles and for the structural graphite web. A quasi steady-state calculation is performed for the fuel rod matrix material. The model accounts for nonlinear adsorption behavior in the fuel rod gap and on the coolant hole boundary. The TRAFIC program is designed to operate in a core survey mode; that is, it performs many repetitive calculations for a large number of spatial locations in the core. This is necessary in order to obtain an accurate volume integrated release. For this reason the program has been designed with calculational efficiency as one of its main objectives. A highly efficient numerical method is used in the solution. The method makes use of the Duhamel superposition principle to eliminate interior spatial solutions from consideration. Linear response functions relating the concentrations and mass fluxes on the boundaries of a homogeneous region are derived. Multiple regions are numerically coupled through interface conditions. Algebraic elimination is used to reduce the equations as far as possible. The problem reduces to two nonlinear equations in two unknowns, which are solved using a Newton Raphson technique.

  18. An analytical solution to time-dependent fission-product diffusion in an HTGR core

    International Nuclear Information System (INIS)

    An analytical time-dependent fission-product diffusion model is solved for the fuel-moderator regions of a high temperature gas-cooled reactor (HTGR) during a hypothetical loss of forced circulation (LOFC) accident. A conservative approximate 1-D model is developed for the fuel and moderator regions, represented in cylindrical and slab geometries, from consideration of the hexagonal fuel-element symmetry. Transport is assumed along the shortest diffusion path and the concentration change across the fuel-moderator interface is approximated by a jump condition. The model is solved by construction of the Green's functions for the Laplace-transformed equations and identification of the pole structure. The concentration and current inverse Laplace transforms are obtained by the Cauchy residue theorem in each region for cubic piecewise polynomial initial conditions. A computer program was developed and validated to evaluate the solution, serve as a benchmark for more sophisticated numerical models and to investigate 90Sr diffusion during a hypothetical LOFC. (author)

  19. Hot beta particles in the lung: Results from dogs exposed to fission product radionuclides

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, F.F.; Griffith, W.C.; Hobbs, C.H. [and others

    1995-12-01

    The Chernobyl nuclear reactor accident resulted in the release of uranium dioxide fuel and fission product radionuclides into the environment with the fallout of respirable, highly radioactive particles that have been termed {open_quotes}hot beta particles.{close_quotes} There is concern that these hot beta particles (containing an average of 150-20,000 Bq/particle), when inhaled and deposited in the lung, may present an extraordinary hazard for the induction of lung cancer. We reviewed data from a group of studies in dogs exposed to different quantities of beta-emitting radionuclides with varied physical half-lives to determine if those that inhaled hot beta particles were at unusual risk for lung cancer. This analysis indicates that the average dose to the lung is adequate to predict biologic effects of lung cancer for inhaled beta-emitting radionuclides in the range of 5-50 Gy to the lung and with particle activities in the range of 0.10-50 Bq/particle.

  20. Study on collection efficiency of fission products by spray: Experimental device and modelling

    Energy Technology Data Exchange (ETDEWEB)

    Ducret, D.; Roblot, D.; Vendel, J. [Institut de Protection et de Surete Nucleaire, Gif-Sur-Yvette (France); Billarand, Y. [ECCO Pharmacie et Chimie, Neuilly (France)

    1997-08-01

    Consequences of an hypothetical overheating reactor accident in nuclear power plants can be limited by spraying cold water drops into containment building. The spray reduces the pressure and the temperature levels by condensation of steam and leads to the washout of fission products (aerosols and gaseous iodine). The present study includes a large program devoted to the evaluation of realistic washout rates. An experimental device (named CARAIDAS) was designed and built in order to determine the collection efficiency of aerosols and iodine absorption by drops with representative conditions of post-accident atmosphere. This experimental device is presented in the paper and more particularly: (1) the experimental enclosure in which representative thermodynamic conditions can be achieved, (2) the monosized drops generator, the drops diameter measurement and the drops collector, (3) the cesium iodide aerosols generator and the aerosols measurements. Modelling of steam condensation on drops aerosols collection and iodine absorption are described. First experimental and code results on drops and aerosols behaviour are compared. 8 refs., 18 figs.

  1. Influence of remaining fission products in low-decontaminated fuel on reactor core characteristics

    International Nuclear Information System (INIS)

    Design study of core, fuel and related fuel cycle system with low-decontaminated fuel has been performed in the framework of the feasibility study (F/S) on commercialized fast reactor cycle systems. This report summarizes the influence on core characteristics of remaining fission products (FPs) in low-decontaminated fuel related to the reprocessing systems nominated in F/S phase I. For simple treatment of the remaining FPs in core neutronics calculation the representative nuclide method parameterized by the FP equivalent coefficient and the FP volume fraction was developed, which enabled an efficient evaluation procedure. As a result of the investigation on the sodium cooled fast reactor with MOX fuel designed in fiscal year 1999, it was found that the pyrochemical reprocessing with molten salt (the RIAR method) brought the largest influence. Nevertheless, it was still within the allowable range. Assuming an infinite-times recycling, the alternations in core characteristics were evaluated as follows: increment of burnup reactivity by 0.5%Δk/kk', decrement of breeding ratio by 0.04, increment of sodium void reactivity by 0.1x10-2Δk/kk' and decrement of Doppler constant (in absolute value) by 0.7x10-3 Tdk/dT. (author)

  2. Electrochemical separation of actinides and fission products in molten salt electrolyte

    Science.gov (United States)

    Gay, R. L.; Grantham, L. F.; Fusselman, S. P.; Grimmett, D. L.; Roy, J. J.

    1995-09-01

    Molten salt electrochemical separation may be applied to accelerator-based conversion (ABC) and transmutation systems by dissolving the fluoride transport salt in LiCl-KCl eutectic solvent. The resulting fluoride-chloride mixture will contain small concentrations of fission product rare earths (La, Nd, Gd, Pr, Ce, Eu, Sm, and Y) and actinides (U, Np, Pu, Am, and Cm). The Gibbs free energies of formation of the metal chlorides are grouped advantageously such that the actinides can be deposited on a solid cathode with the majority of the rare earths remaining in the electrolyte. Thus, the actinides are recycled for further transmutation. Rockwell and its partners have measured the thermodynamic properties of the metal chlorides of interest (rare earths and actinides) and demonstrated separation of actinides from rare earths in laboratory studies. A model is being developed to predict the performance of a commercial electrochemical cell for separations starting with PUREX compositions. This model predicts excellent separation of plutonium and other actinides from the rare earths in metal-salt systems.

  3. Structural study and properties of peraluminous formulations for the fission products and minor actinides confinement

    International Nuclear Information System (INIS)

    In this work, peraluminous glasses (lack of alkaline and alkaline earth ions regarding aluminum) are under study to assess the potentiality of these matrices to confine fission products and minor actinides (FPA) at higher rate than current R7T7 glass (18,5 wt % FPA). The first part of this work aims at studying the physical and chemical properties of complex peraluminous glasses containing increasing FPA rate (18.5 to 32 wt %) to compare them with the specifications. The very low crystallization tendency of complex glasses containing up to 22.5 wt % as well as the very good chemical durability observed are major assets. The other part focuses on the lanthanides incorporation in simplified glass compositions in the SiO2-B2O3-Al2O3-Na2O-CaO-Ln2O3 system (Ln = Nd or La). The glass homogeneity and devitrification tendency are investigated at different scales by XRD, SEM, TEM and structural techniques such as NMR (MAS, MQMAS, REDOR, HMQC, DHMQC) and neodymium optical spectroscopy that appear very powerful to determine the lanthanides structural role regarding aluminum and describe more precisely the structural organization of peraluminous network, as still unknown in such systems. The glass homogeneity was demonstrated in a large composition domain and new structural data were put in evidence at high lanthanides content. (author)

  4. Fission-product chemistry in severe reactor accidents: Review of relevant integral experiments

    International Nuclear Information System (INIS)

    The attenuation of the radioactive fission-product emission from a severe reactor accident will depend on a combination of chemical, physical and thermal-hydraulic effects. Chemical species stabilised under the prevailing conditions will determine the extent of aerosol formation and any subsequent interaction, so defining the magnitude and physical forms of the eventual release into the environment. While several important integral tests have taken place in recent years, these experiments have tended to focus on the generation of mass-balance and aerosol-related data to test and validate materials-transport codes rather than study the impact of important chemical phenomena. This emphasis on thermal hydraulics, fuel behaviour and aerosol properties has occurred in many test (e.g. PBF, DEMONA, Marviken-V, LACE and ACE). Nevertheless, the generation and reaction of the chemical species in all of these programmes determined the transport properties of the resulting vapours and aerosols. Chemical effects have been studied in measurements somewhat subsidiary to the main aims of the tests. This work has been reviewed in detail with respect to Marviken-V, LACE, ACE and Falcon. Specific issues remain to be addressed, and these are discussed in terms of the proposed Phebus-FB programme. (author). 58 refs, 9 figs, 1 tab

  5. Modelling and simulation the radioactive source-term of fission products in PWR type reactors

    International Nuclear Information System (INIS)

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  6. Neutron cross sections of 28 fission product nuclides adopted in JENDL-1

    International Nuclear Information System (INIS)

    This is the final report concerning the evaluated neutron cross sections of 28 fission product nuclides adopted in the first version of Japanese Evaluated Nuclear Data Library (JENDL-1). These 28 nuclides were selected as being most important for fast reactor calculations, and are 90Sr, 93Zr, 95Mo, 97Mo, 99Tc, 101Ru, 102Ru, 103Rh, 104Ru, 105Pd, 106Ru, 107Pd, 109Ag, 129I, 131Xe, 133Cs, 135Cs, 137Cs, 143Nd, 144Ce, 144Nd, 145Nd, 147Pm, 147Sm, 149Sm, 151Sm, 153Eu and 155Eu. The status of the experimental data was reviewed over the whole energy range. The present evaluation was performed on the basis of the measured data with the aid of theoretical calculations. The optical and statical models were used for evaluation of the smooth cross sections. An improved method was developed in treating the multilevel Breit-Wigner formula for the resonance region. Various physical parameters and the level schemes, adopted in the present work are discussed by comparing with those used in the other evaluations such as ENDF/B-IV, CEA, CNEN-2 and RCN-2. Furthermore, the evaluation method and results are described in detail for each nuclide. The evaluated total, capture and inelastic scattering cross sections are compared with the other evaluated data and some recent measured data. Some problems of the present work are pointed out and ways of their improvement are suggested. (author)

  7. Composition of high fission product wastes resulting from future reprocessing of commercial nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Swanson, J.L

    1986-07-01

    Pacific Northwest Laboratory studies, aimed at defining appropriate glass compositions for future disposal of high-level wastes, have developed composition ranges for the waste that will likely result during reprocessing of Light Water Reactor (LWR) and Liquid Metal Reactor (LMR) fuels. The purpose of these studies was to provide baseline waste characterizations for possible future commercial high-level waste so that waste immobilization technologies (e.g., vitrification) can be studied. Ranges in waste composition are emphasized because the waste will vary with time as different fuels are reprocesses, because choice of process chemicals is nuclear, and because fuel burnups will vary. Consequently, composition ranges are based on trends in fuel reprocessing procedures and on achievable burnups in operating reactors. In addition to the fission product and actinide elements, which are the primary hazardous materials in the waste, likely composition ranges are given for inert elements that may be present in the waste. These other elements may be present because of being present in the fuel, because of being added as process chemical during reprocessing, because of being added during equipment decontamination, or because of corrosion of plant equipment and/or fuel element cladding. This report includes a discussion of the chemicals added in variation of the PUREX process, which is likely to remain the favored reprocessing technique for commercial nuclear fuels. Consideration is also given to a pyrochemical process proposed for the reprocessing of some LMR fuels.

  8. Modeling of thermal hydraulic behavior and fission product releases in degraded cores

    International Nuclear Information System (INIS)

    When core material reaches melting conditions severe degradation of the core geometry occurs. Data available on the core behavior in a severely degraded state suggest that extensive blockage of the flow channels would occur. If a sufficient bypass is available for the gas flow, such as in the LOFT LP-FP-2 test, severe retardation of the hydrogen and fission product sources from the degraded channel is suggested from the available data. This phenomena is expected to occur in an LWR core and should be considered by core models that are used for severe accident analysis. In the MAAP code it is done by preventing gas flow through molten core regions. Good agreement is obtained with all relevant data that are directly applicable to LWR accident conditions. A more mechanistic model for the freezing of core material and its effect on the coolant channel geometry is currently being investigated by the US Department of Energy Advanced Reactor Severe Accident Program (ARSAP). 8 refs., 2 figs

  9. Distribution of fission products in graphite sleeves and blocks of the eleventh and twelfth OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    The 11th and 12th fuel assemblies were irradiated in an in-pile gas loop, OGL-1, installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Distribution of fission products in the graphite sleeves and blocks of the assemblies was measured by gamma-ray spectrometry. The 11th fuel assembly was aimed at testing the irradiation performance of mass product fuels in trial manufacturing of the first charge fuel for the High Temperature Engineering Test Reactor (HTTR) in relatively short irradiation, and the 12th assembly in long-term irradiation. The 12th assembly attained a burnup approximately as high as that of the HTTR driver fuel design. In the graphite sleeve of the 11th assembly, high concentration peaks of fission products were found in the axial distribution. Exposure of failed fuel particles was not detected on the surface of fuel compacts, while fissures of graphite matrix at overcoat boundaries were observed on the surface. These results led to a presumption that fission products, which were released from failed particles located inside of the fuel compact, was easily transported through the fissures of the matrix to the inner surface of the sleeve. In the graphite sleeve of the 12th assembly, 110mAg was detected together with other fission products of 137Cs, 134Cs etc. Silver-110m showed characteristic distribution: this nuclides was less concentrated at the region with highly concentrated 60Co which is presumed to have been transported from melted sheath material of thermocouples to the graphite sleeve. It was inferred from the distribution that the transport behavior of 110mAg had been influenced by co-sorption or by pore structure change in the graphite material of the sleeve, which had been induced by metallic elements including cobalt. (author)

  10. Distribution of fission products in graphite sleeves and blocks of the eleventh and twelfth OGL-1 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Fukuda, Kousaku; Kikuchi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tsuruta, Harumichi

    1994-06-01

    The 11th and 12th fuel assemblies were irradiated in an in-pile gas loop, OGL-1, installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Distribution of fission products in the graphite sleeves and blocks of the assemblies was measured by gamma-ray spectrometry. The 11th fuel assembly was aimed at testing the irradiation performance of mass product fuels in trial manufacturing of the first charge fuel for the High Temperature Engineering Test Reactor (HTTR) in relatively short irradiation, and the 12th assembly in long-term irradiation. The 12th assembly attained a burnup approximately as high as that of the HTTR driver fuel design. In the graphite sleeve of the 11th assembly, high concentration peaks of fission products were found in the axial distribution. Exposure of failed fuel particles was not detected on the surface of fuel compacts, while fissures of graphite matrix at overcoat boundaries were observed on the surface. These results led to a presumption that fission products, which were released from failed particles located inside of the fuel compact, was easily transported through the fissures of the matrix to the inner surface of the sleeve. In the graphite sleeve of the 12th assembly, {sup 110m}Ag was detected together with other fission products of {sup 137}Cs, {sup 134}Cs etc. Silver-110m showed characteristic distribution: this nuclides was less concentrated at the region with highly concentrated {sup 60}Co which is presumed to have been transported from melted sheath material of thermocouples to the graphite sleeve. It was inferred from the distribution that the transport behavior of {sup 110m}Ag had been influenced by co-sorption or by pore structure change in the graphite material of the sleeve, which had been induced by metallic elements including cobalt. (author).

  11. Determination of Fission Product Yields of 235U, 238U and 239Pu for Neutron Energies from 0.5 to 14.8 MeV

    Science.gov (United States)

    Gooden, Matthew; Arnold, Charles; Becker, John; Bhatia, Chitra; Bhike, Megha; Fowler, Malcolm; Howell, Calvin; Kelley, John; Stoyer, Mark; Tonchev, Anton; Tornow, Werner; Vieira, Dave; Wilhelmy, Jerry

    2014-03-01

    A joint TUNL-LANL-LLNL collaboration has been formed to study the issue of possible energy dependences for certain fission product isotopes. Work has been carried out at the TUNL 10 MV Tandem accelerator which produces nearly mono-energetic neutrons via either 2H(d,n)3He,3H(d,n)4He,or3H(p,n)3He reactions. Three dual fission ionization chambers dedicated to 235U, 238U and 239Pu thick target foils and thin monitor foils respectively, were exposed to the neutron beams. After irradiation, thick target foils were gamma counted over a period of 1-2 months and characteristic gamma rays from fission products were recorded using HPGe detectors at TUNL's low background counting area. Using the dual fission chambers, relative fission product yield were determined at a high precision of 2-3 % as well as absolute fission product yields at a lower precision of 5-6 %. Preliminary results will be presented for a number of fission product isotopes over the incident neutron energy range of 0.5 to 14.8 MeV.

  12. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 3: Fission-Product Transport and Dose PIRTs

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert Noel [ORNL

    2008-03-01

    This Fission Product Transport (FPT) Phenomena Identification and Ranking Technique (PIRT) report briefly reviews the high-temperature gas-cooled reactor (HTGR) FPT mechanisms and then documents the step-by-step PIRT process for FPT. The panel examined three FPT modes of operation: (1) Normal operation which, for the purposes of the FPT PIRT, established the fission product circuit loading and distribution for the accident phase. (2) Anticipated transients which were of less importance to the panel because a break in the pressure circuit boundary is generally necessary for the release of fission products. The transients can change the fission product distribution within the circuit, however, because temperature changes, flow perturbations, and mechanical vibrations or shocks can result in fission product movement. (3) Postulated accidents drew the majority of the panel's time because a breach in the pressure boundary is necessary to release fission products to the confinement. The accidents of interest involved a vessel or pipe break, a safety valve opening with or without sticking, or leak of some kind. Two generic scenarios were selected as postulated accidents: (1) the pressurized loss-of-forced circulation (P-LOFC) accident, and (2) the depressurized loss-of-forced circulation (D-LOFC) accidents. FPT is not an accident driver; it is the result of an accident, and the PIRT was broken down into a two-part task. First, normal operation was seen as the initial starting point for the analysis. Fission products will be released by the fuel and distributed throughout the reactor circuit in some fashion. Second, a primary circuit breach can then lead to their release. It is the magnitude of the release into and out of the confinement that is of interest. Depending on the design of a confinement or containment, the impact of a pressure boundary breach can be minimized if a modest, but not excessively large, fission product attenuation factor can be introduced into

  13. History and Actual State of Non-HEU Fission-Based Mo-99 Production with Low-Performance Research Reactors

    Directory of Open Access Journals (Sweden)

    S. Dittrich

    2013-01-01

    Full Text Available Fifty years ago, one of the worldwide first industrial production processes to produce fission-Mo-99 for medical use had been started at ZfK Rossendorf (now: HZDR, Germany. On the occasion of this anniversary, it is worth to mention that this original process (called LITEMOL now together with its target concept used at that time can still be applied. LITEMOL can be adapted very easily to various research reactors and applied at each site, which maybe still of interest for very small-scale producers. Besides this original process, two further and actually proven processes are suitable as well and recommended for small-scale LEU fission Mo-99 production also. They are known under the names KSA/KSS COMPACT and ROMOL LITE and will be described below.

  14. Heated uranium tetrafluoride target system to release non-rare gas fission products for the TRISTAN isotope separator

    International Nuclear Information System (INIS)

    The development of a heated uranium tetrafluoride target system for the TRISTAN isotope separator to release non-rare gas fission products is presented. Off-line experiments indicated that fluorides of As, Se, Br, Kr, Zr, Nb, Mo, Tc, Ru, Sb, Te, I and Xe could be volatilized, but except for Br, Kr, I and Xe, none of these elements was observed after mass separation in the on-line experiments. The results of the on-line experiments indicated a very low level of hydride contamination at ambient temperature and consequently, uranium tetrafluoride replaced uranyl stearate as the primary gaseous fission product target. Possible reasons for the failure of the heated target system to yield non-rare gas activities are discussed and suggestions for designing a new heated target system are presented

  15. Stainless steel-zirconium alloy waste forms for metallic fission products and actinides during treatment of spent nuclear fuel

    International Nuclear Information System (INIS)

    Stainless steel-zirconium waste form alloys are being developed for the disposal of metallic wastes recovered from spent nuclear fuel using an electrometallurgical process developed by Argonne National Laboratory. The metal waste form comprises the fuel cladding, noble metal fission products and other metallic constituents. Two nominal waste form compositions are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels. The noble metal fission products are the primary source of radiation and their contribution to the waste form radioactivity has been calculated. The disposition of actinide metals in the waste alloys is also being explored. Simulated waste form alloys were prepared to study the baseline alloy microstructures and the microstructural distribution of noble metals and actinides, and to evaluate corrosion performance

  16. TRANCS, a computer code for calculating fission product release from high temperature gas-cooled reactor fuel, (2)

    International Nuclear Information System (INIS)

    This report describes the calculation procedure of the TRANCS code, which deals with fission product transport in fuel rod of high temperature gas-cooled reactor (HTGR). The fundamental equation modeled in the code is a cylindrical one-dimensional diffusion equation with generation and decay terms, and the non-stationary solution of the equation is obtained numerically by a finite difference method. The generation terms consist of the diffusional release from coated fuel particles, recoil release from outer-most coating layer of the fuel particle and generation due to contaminating uranium in the graphite matrix of the fuel compact. The decay term deals with neutron capture as well as beta decay. Factors affecting the computation error has been examined, and further extention of the code has been discussed in the fields of radial transport of fission products from graphite sleeve into coolant helium gas and axial transport in the fuel rod. (author)

  17. Cement As a Waste Form for Nuclear Fission Products: The Case of 90Sr and Its Daughters

    OpenAIRE

    Dezerald, Lucile; Kohanoff, Jorge J.; Correa, Alfredo A.; Caro, Alfredo; Pellenq, Roland J.-M.; Ulm, Franz J.; Saúl, Andrés

    2015-01-01

    One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DF...

  18. A physical description of fission product behavior fuels for advanced power reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  19. Fission product release from UO2 during irradiation. Diffusion data and their application to reactor fuel pins

    International Nuclear Information System (INIS)

    Release of fission product species from UO2, and to a limited extent from (U, Pu)02 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO2 powder 20% 235U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x1016 neutrons m-2s-1 providing a heat rating within the samples of 34.5 MW/teU

  20. The behaviour of fission product caesium in LMFBR primary circuits: experiments in a small stainless steel loop containing circulating sodium

    International Nuclear Information System (INIS)

    The available information which can be used to determine the likely behaviour of fission product caesium, released from defective fuel pins, in the primary circuit of a sodium-cooled fast reactor has been reviewed and assessed. For a relatively small number of fuel pin failures over a reactor lifetime 137Cs is potentially a predominant contributor to dose rates on out-of-core steel components if the adsorption of caesium on these components is sufficiently high. Shortcomings are identified in the existing data on caesium adsorption from liquid sodium onto austenitic steel surfaces. This has enabled the definition of those areas of work which are necessary before the behaviour of fission product caesium in the primary circuit, and the possible requirement of any alleviatory measures, can be predicted with precision and confidence. Experiments are described from which the mechanism of adsorption of caesium onto steel surfaces was elucidated, thereby enabling the prediction of the likely sorption behaviour of fission product caesium on components in the primary circuits of LMFBRs. The potential use of cold-trapping as a means of removing caesium from sodium was also investigated and found to be of no significant value; alternative methods for caesium removal are discussed. The development of methods for decontamination of steelwork which has become contaminated by caesium sorption onto chromium (III) oxides is described, although it is not clear that the levels of contamination which would arise by this mechanism under normal LMFBR operating conditions would necessitate such treatment. (author)

  1. A generalized method for characterization of 235U and 239Pu content using short-lived fission product gamma spectroscopy

    Science.gov (United States)

    Knowles, Justin; Skutnik, Steven; Glasgow, David; Kapsimalis, Roger

    2016-10-01

    Rapid nondestructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the Oak Ridge National Laboratory High Flux Isotope Reactor Neutron Activation Analysis facility has developed a generalized nondestructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and makes use of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a complete characterization of isotopic identification, mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% recovery bias have been conducted on standards of 235U and 239Pu as low as 12 ng in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 198 ng of fissile mass with less than 7% recovery bias. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. It is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation facilities, and account for increasingly complex sample matrices.

  2. Influences of Zr, Ce and Ba fission products on the surface properties of UO2: Atomistic simulations

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Tian, Xiaofeng; Chen, Hongsheng

    2016-07-01

    Molecular dynamics (MD) simulations with a shell-core model have been carried out to investigate the influences of Zr, Ce and Ba fission products on the surface properties of UO2. Simulation results indicate that (i) the presence of these fission products will change the surface energy of three low-index surfaces in UO2; (ii) the individual addition of Ce has no significant effect on the surface energy, while the individual addition of Ba will dramatically decrease the surface energy of UO2 by approximately 18% on (100) surface, 7% on (110) surface and 9% on (111) surface with the Ba contents ranging from 0 to 12.5 mol% at 300 K, which is obviously contrary to the Zr; (iii) the combined additions of Zr, Ce and Ba fission products will continuously increase the surface energy of UO2 (100), (110) and (111) surfaces; (iv) the structures of the three low-index surfaces in pure UO2 as well as U0.8(Zr, Ce, Ba)0.2O2 are dramatically disturbed after the free relaxation; (v) The nearest O atoms move towards the Zr and Ce atoms center by about 0.21 Å and 0.12 Å but move away from the Ba atom center by about 0.27 Å.

  3. A new evaluation of fission product yields and the production of a new library (UKFY2) of independent and cumulative yields. Pt. 2

    International Nuclear Information System (INIS)

    A new evaluation has been prepared of the independent and cumulative yields of the products of fission induced by thermal, fast, and 14 MeV neutrons in nuclides important for reactor design and operation and for fuel and waste management. Three spontaneously fissioning nuclides were also considered. The evaluation used a database that is considered to be complete up to early 1989. Careful study was made of experimental uncertainties and discrepancies, emphasising the need for further measurements. Gaps in the data were filled by interpolation and extrapolation, using fits to empirical models. The yields were adjusted to fit physical constraints of the fissioning process. The present report contains Tables of cumulative and chain yields, and of fractional independent yields. Each set of Tables gives all the relevant measurements in the database, with uncertainties. Recommended weighted averages are included with standard deviations, discrepant sets of measurements are clearly indicated and references to all the entries in the database are listed. (author)

  4. Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions

    International Nuclear Information System (INIS)

    The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO2 and highly enriched (HEU) TRISO UC2 particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO2 particles and 23.5 and 74% FIMA for UC2 particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium

  5. Intermediate energy nuclear fission

    International Nuclear Information System (INIS)

    Nuclear fission has been investigated with the double-kinetic-energy method using silicon surface barrier detectors. Fragment energy correlation measurements have been made for U, Th and Bi with bremsstrahlung of 600 MeV maximum energy. Distributions of kinetic energy as a function of fragment mass are presented. The results are compared with earlier photofission data and in the case of bismuth, with calculations based on the liquid drop model. The binary fission process in U, Yb, Tb, Ce, La, Sb, Ag and Y induced by 600 MeV protons has been investigated yielding fission cross sections, fragment kinetic energies, angular correlations and mass distributions. Fission-spallation competition calculations are used to deduce values of macroscopic fission barrier heights and nuclear level density parameter values at deformations corresponding to the saddle point shapes. We find macroscopic fission barriers lower than those predicted by macroscopic theories. No indication is found of the Businaro Gallone limit expected to occur somewhere in the mass range A = 100 to A = 140. For Ce and La asymmetric mass distributions similar to those in the actinide region are found. A method is described for the analysis of angular correlations between complementary fission products. The description is mainly concerned with fission induced by medium-energy protons but is applicable also to other projectiles and energies. It is shown that the momentum and excitation energy distributions of cascade residuals leading to fission can be extracted. (Author)

  6. Development of fission Mo-99 production technology - A nuclear feasibility study on UN target for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun; Kim, Woo Sik [Kyunghee University, Seoul (Korea)

    2000-03-01

    Nuclear target design satisfying all the constraints for fission moly production in HANARO was proposed in this project. The 'MCNP-ORIGEN' code system which was previously proposed for a design tool, was evaluated by the comparison with through the 'MCNP-Analytic Eq.' system. A characteristics of each chemical processing step were analysed and material balance was set up to evaluate the overall yield ratio of Mo-99 recovery. A parametric study was done for the optimum HEU target design. Tested parameters were target thickness, recoil-loss rate to the fuel thickness, target radius, cladding materials, thickness of irradiation guide tube, and barrier materials. Optimized HEU target design was proposed which satisfying the constraints and having high production yield. For a LEU target design using 19.7 w/o UN powder fuel, a parametric study was also done for the optimization of fuel thickness, powder packing density, mixture material volume ratio. 24 refs., 35 figs., 57 tabs. (Author)

  7. Fission product release and microstructure changes of irradiated MOX fuel at high temperatures

    Science.gov (United States)

    Colle, J.-Y.; Hiernaut, J.-P.; Wiss, T.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    present in the vapour between 83 and 300 a.m.u. were measured during the heating. Additionally, the 85Kr isotope was analysed in a cold trap by β and γ counting. The long-lived fission gas isotopes correspond to masses 131, 132, 134 and 136 for Xe and 83, 84, 85 and 86 for Kr. The absolute quantities of gas released from specimens of sample types A and B were also determined using the in-house built Q-GAMES (Quantitative gas measurement system), described in detail in [15].For each of the samples, fragments were also annealed and measured in the KEMS up to specific temperatures corresponding to different stages of the FGs or He release. These fragments were subsequently analysed by Scanning Electron Microscopy (SEM, Philips XL40) [16] in order to investigate the relationship between structural changes, burn-up, irradiation temperature and fission products release. SEM observations were also done on the samples before the KEMS experiments and the fracture surface appearance of the samples is shown in Fig. 3, revealing the presence of the high burnup structure (HBS) in the Pu-rich agglomerates.A summary of the 12 samples analysed by KEMS, SEM and Q-GAMES is given in Table 1. At 1300 K no clear change potentially related to gas release appears in the UM and PA. At 1450 K a beginning of grain boundaries opening can be observed as well as rounding of the grains attributed to thermal etching. At 1600 K a densification is observed in the PA, smalls grains seem to agglomerate. At 1800 K grain coalescence has occurred in the PA together with formation of large pores. In the UM one observes the formation of a network of intergranular channels. Finally, at 2100 K re-sintering proceeds further and large intra-granular bubbles and five metal precipitates becomes visible. The micrographs of sample type B at 1700 K in Fig. 10, show the formation of small intergranular channel not observed on the image of the sample type A at 1600 K. At 2200 K the intragranular bubbles and

  8. HTR-2014 Paper Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2001-10-01

    The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of fission products silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of fission products from fuel compacts and fuel particles, and retention of fission products in the compacts outside of the SiC layer. PARFUME-predicted fractional release of these fission products was determined and compared to the PIE measurements. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed silicon carbide (SiC) layers, the over-prediction is by a factor of about two, corresponding to an over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of about 100. For intact particles, whose release is much lower, the over-prediction is by an average of about an order of magnitude, which could additionally be attributed to an over-estimated diffusivity in SiC by about 30%. The release of strontium from intact particles is also over-estimated by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Furthermore, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the

  9. Production Cycle for Large Scale Fission Mo-99 Separation by the Processing of Irradiated LEU Uranium Silicide Fuel Element Targets

    Directory of Open Access Journals (Sweden)

    Abdel-Hadi Ali Sameh

    2013-01-01

    Full Text Available Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2 compound. Excellent results are achieved by the oxidizing dissolution of the fuel meat in hydrofluoric acid at room temperature. The resulting solution is directly behind added to an over stoichiometric amount of potassium hydroxide solution. Uranium and the bulk of fission products are precipitated together with the transuranium compounds. The filtrate contains the molybdenum and the soluble fission product species. It is further treated similar to the in-full scale proven process. The generated off gas stream is handled also as experienced before after passing through KOH washing solution. The generated alkaline fluoride containing waste solution is noncorrosive. Nevertheless fluoride can be selectively bonded as in soluble CaF2 by addition of a mixture of solid calcium hydroxide calcium carbonate to the sand cement mixture used for waste solidification. The generated elevated amounts of LEU remnants can be recycled and retargeted. The related technology permits the minimization of the generated fuel waste, saving environment, and improving processing economy.

  10. High density electronic excitation effects on microstructural evolution in CeO2 under irradiations with high energy fission products

    International Nuclear Information System (INIS)

    For progressing high burnup extension of LWR fuels, formation and growth mechanism of a crystallographic re-structuring in the periphery region of high burnup fuel pellets, as named 'rim structure', should be clarified. The structure shall be formed by the accumulation and mutual interactions of radiation damages, fission products (FPs) and electronic excitations deposited partially by nuclear fissions. In order to clarify electronic excitation effects on the microstructural evolution in CeO2, 70-210 MeV FP ions (Xe, I, Zr) irradiation examinations on CeO2 have been done at JAERI-Tandem facility. These experiments clarify that 1) the effective area of electronic excitation by high energy fission products might be around 5-7 mmφ, and the square of ion track diameter tends to be proportional to the electronic stopping power (Se), and 2) overlapping of ion tracks, under 210 MeV Xe irradiation to a fluence of 1x1015 ions/cm2, makes the surface to be rough, whose size of the roughness is around 1 μm. (author)

  11. In-pile release behavior of metallic fission products in graphite materials of an HTGR fuel assembly

    International Nuclear Information System (INIS)

    Distribution of metallic fission products in the graphite sleeve and block of the fifth OGL-1 fuel assembly was measured by gamma spectrometry with lathe sectioning. Considerably large release fractions of long-lived fission products with smooth axial profiles were observed in the sleeve due to a large failure fraction of coated fuel particles accompanied with failed silicon carbide layers. Nevertheless, a key nuclide 110mAg, whose large release is suspected at increased burnups for low-enriched uranium fuels, was effectively retained within the graphite sleeve. The retention was also observed for 125Sb, 154Eu and 155Eu up to a burnup of 3.2% fission per initial metal atom, but was limited for 134Cs and 137Cs at high sleeve-temperatures above 9000C. In-pile diffusion coefficients in IG-110 graphite have been evaluated for Cs, Ag and Sb; those for Cs are in reasonable agreement with available in-pile data. (orig.)

  12. ASTEC V2 severe accident integral code: Fission product modelling and validation

    Energy Technology Data Exchange (ETDEWEB)

    Cantrel, L., E-mail: laurent.cantrel@irsn.fr; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-06-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix.

  13. NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

    Energy Technology Data Exchange (ETDEWEB)

    OH,S.Y.; CHANG,J.; MUGHABGHAB,S.

    2000-05-11

    Neutron cross section evaluations of the fission-product isotopes, {sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, {sup 141}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd, and {sup 157}Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of {sup 155}Gd and {sup 157}Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations.

  14. Final report on ARPA fission yield project work at Battelle-Northwest, April 1970--April 1973

    International Nuclear Information System (INIS)

    The overall objective has been to measure the independent and cumulative fission yields of selected halogen and rare gas nuclides for application to characterization of underground nuclear detonations. The studies have included fission yield measurements for thermal, fission spectrum, and 15 MeV neutron-induced fission events. Target materials included 235U, 238U and 239Pu. The research effort was divided into two basic parts. In one part, the nuclides of interest were separated radiochemically and determined by gamma-ray spectrometry. This approach provides information on the independent and cumulative yields of nuclides with half-lives of a few seconds or greater. The second part of our effort involved the use of on-line mass separation techniques. This approach yields information on independent fission yields of nuclides with half-lives ranging down to fractions of a second and provides data on all significant isotopes of a given fission product element in one set of measurements. The main effort in the radiochemistry program was centered on measurements of the cumulative fission yield of 89Kr. Cumulative fission yields of 89Kr were measured for thermal-neutron fission of 239Pu and for fission-spectrum and 15-MeV neutron fission of 235U, 238U and 239Pu. In addition, cumulative fission yields of the other rare gas radionuclides, /sup 85m/Kr, 87Kr, 88Kr, 137Xe, 138Xe, were measured for the same fission type events. Fractional independent yields of 89Rb and 138Cs were also measured for a limited number of fission systems. On-line mass spectrometer facilities were established at a Van de Graaff accelerator and at a nuclear reactor. Measurements were made of relative independent fission yields of rubidium isotopes of masses 89 through 97 and of cesium isotopes of masses 139 through 145.(U.S.)

  15. Actinide, Activation Product and Fission Product Decay Data for Reactor-based Applications

    International Nuclear Information System (INIS)

    The UK Activation Product Decay Data Library was first released in September 1977 as UK-PADD1, to be followed by regular improvements on an almost yearly basis up to the assembly of UKPADD6.12 in March 2013. Similarly, the UK Heavy Element and Actinide Decay Data Library followed in December 1981 as UKHEDD1, with the implementation of various modifications leading to UKHEDD2.6, February 2008. Both the data content and evaluation procedures are defined, and the most recent evaluations are described in terms of specific radionuclides and the resulting consistency of their recommended decay-data files. New versions of the UKPADD and UKHEDD libraries are regularly submitted to the NEA Data Bank for possible inclusion in the JEFF library

  16. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    Science.gov (United States)

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-01

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still

  17. Effects of radiation and fission product incorporation in a yttria-stabilized zirconia based inert matrix fuel

    Science.gov (United States)

    Zhu, Sha

    This work has investigated the irradiation and incorporation effects of fission products in a yttria-stabilized zirconia (YSZ) based inert matrix fuel (IMF). The concept of inert matrix fuel is based on a new strategy for disposition of plutonium generated from the reprocessing of commercial nuclear fuel and the dismantling of nuclear weapons, i.e. using uranium-free oxides to "burn" plutonium and other actinides (Np, Cm, and Am) in reactors. This approach allows direct disposal, without reprocessing, after once-through burn-up. YSZ and MgAl2O4-YSZ composites are among the potential ceramics for IMF due to their high chemical durability and radiation resistance. The research involved investigating the production, nature, and accumulation of irradiation-induced defects, the behavior of the fission products in the ceramics, the structural stability and amorphization resistance of the YSZ during implantation. Ion implantations were conducted with 200--400 keV Cs+, Sr+, I+, Xe+ and Ti+ up to fluences of 1 x 1017/cm 2 at both room temperature and temperatures of 600--700°C. Thermal annealing was subsequently completed after room temperature ion implantations. In situ and ex situ transmission electron microscopy (TEM), optical absorption spectroscopy, photo-luminescence spectroscopy, and electron paramagnetic resonance (EPR) spectroscopy were employed to characterize the irradiation induced defect evolution and analyze the defect structures. Various irradiation effects were observed and determined in the experiments, such as point defects (F type and V type color centers), defect clusters (dislocation loops), cavities (voids and bubbles), the crystalline-to-amorphous transition, and the phase transformation from fluorite to pyrochlore structure. The ion irradiation-induced amorphization mechanism, the retention ability of the fission products, and structural stability of YSZ are discussed in terms of ion incorporation effects, implanted ion radii, and the solubility

  18. Neutron capture cross-section of fission products in the European activation file EAF-3. Presented at the NEA Specialists` Meeting on Fission Product Nuclear Data, JAERI, Japan, 25-27 May 1992

    Energy Technology Data Exchange (ETDEWEB)

    Kopecky, J.; Delfini, M.G.; Kamp, H.A.J. van der; Gruppelaar, H.; Nierop, D. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1992-05-01

    This paper contains a description of the work performed to extend and revise the neutron capture data in the European Activation File (EAF-3) with emphasis on nuclides in the fission-product mass range. The starter was the EAF-1 data file from 1989. The present version, EAF/NG-3, contains (n,{gamma}) excitation functions for all nuclides (729 targets) with half-lives exceeding 1/2 day in the mass range from H-1 to Cm-248. The data file is equipped with a preliminary uncertainty file, that will be improved in the near future. (author). 19 refs.; 5 figs.; 3 tabs.

  19. Proceedings of the Eleventh Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation

    International Nuclear Information System (INIS)

    Partitioning and transmutation (P and T) is one of the key technologies for reducing the radiotoxicity and volume of radioactive waste arisings. Recent developments indicate the need for embedding P and T strategies in advanced fuel cycles considering both waste management and economic issues. In order to provide experts a forum to present and discuss state-of-the-art developments in the P and T field, the OECD/NEA has been organising biennial information exchange meetings on actinide and fission product partitioning and transmutation since 1990. The previous meetings were held in Mito (Japan) in 1990, at Argonne (United States) in 1992, in Cadarache (France) in 1994, in Mito (Japan) in 1996, in Mol (Belgium) in 1998, in Madrid (Spain) in 2000, in Jeju (Korea) in 2002, in Las Vegas (United States) in 2004, in Nimes (France) in 2006 and in Mito (Japan) in 2008. They have often been co-sponsored by the European Commission (EC) and the International Atomic Energy Agency (IAEA). The 11. Information Exchange Meeting was held in San Francisco, California, United States on 1-4 November 2010, comprising a plenary session on national P and T programmes and six technical sessions covering various fields of P and T. The meeting was hosted by the Idaho National Laboratory (INL), United States. The information exchange meetings on P and T form an integral part of NEA activities on advanced nuclear fuel cycles. The meeting covered scientific as well as strategic/policy developments in the field of P and T, such as: fuel cycle strategies and transition scenarios; radioactive waste forms; the impact of P and T on geological disposal; radioactive waste management strategies (including secondary wastes); transmutation fuels and targets; pyro and aqueous separation processes; materials, spallation targets and coolants; transmutation physics, experiments and nuclear data; transmutation systems (design, performance and safety); handling and transportation of transmutation fuels; and

  20. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    Science.gov (United States)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  1. Decay Heat Analyses after Thermal-Neutron Fission of {sup 235}U and {sup 239}Pu by SCALE-6.1.3 with Recently Available Fission Product Yield Data

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The heat reaches about 1.5% after one hour and falls to 0.4% after a day. After a week it will be about 0.2%. The reactor, however, still requires further cooling for several years to keep the fuel rods safe. In general, the decay heat in the reactors can be calculated using a summation calculation method, which is simply the sum of the activities of the fission products produced during the fission process and after the reactor shutdown weighted by the mean decay energies. Consequently, the method is strongly dependent on the available nuclear structure data. Nowadays, the method has been implemented in various burnup and depletion programs such as ORIGEN and CINDER. In this study, the decay heat measurements after thermal-neutron fission of {sup 235}U and {sup 239}Pu have been evaluated by the ORIGEN-S with the decay data and fission product yield libraries included in the SCALE-6.1.3 software package. The new libraries were applied to the decay heat calculations, and the results were compared with those by the ORIGEN reference calculation. The decay heat measurements for very short cooling times after thermal-neutron fission of {sup 235}U and {sup 239}Pu have been evaluated by the ORIGEN-S summation calculation. The reference calculation results by the latest ORIGEN data libraries of the SCALE-6.1.3 have been validated with the measurements by ORNL and Studsvik. In addition, the generation of the new ORIGEN yield libraries has been completed based on the ENDF/B-VII.1, JEFF-3.1.1, JENDL/FPY-2011, and JENDL-4.0. The new libraries have been successfully applied to the decay heat calculations and comparative analyses have been devoted to verifying the importance of the fission product yield data when estimating the decay heat values for each isotope in a very short time. The decay data library occupies an important position in the ORIGEN summation calculation along with the fission product yield library.

  2. Studies on short-lived fission products at the Mainz TRIGA reactor

    International Nuclear Information System (INIS)

    Neutron-rich nuclei of medium mass number are produced by thermal-neutron-induced fission of heavy elements, e.g., 235U, 239Pu, and 249Cf. Pulse irradiations lead to an enhancement of the ratio of short-lived activities to the accompanying longer-lived components. One approach for investigating the properties of short-lived nuclei consists in a combination of rapid chemical separations with higher-resolution gamma spectroscopy. This is demonstrated by the isolation of neutron-rich isotopes of niobium by sorption on glass and of ruthenium by solvent extraction. Other rapid separation procedures from aqueous solutions are briefly summarized and a few examples for their application in nuclear fission- and delayed neutron studies are given. Some experiments with an on-line mass separator of the ISOLDE-type, using chemical targets, are described. (U.S.)

  3. Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

    2010-09-23

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

  4. Arrival time and magnitude of airborne fission products from the Fukushima, Japan, reactor incident as measured in Seattle, WA, USA

    CERN Document Server

    Leon, J Diaz; Knecht, A; Miller, M L; Robertson, R G H; Schubert, A G

    2011-01-01

    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.

  5. Analytic benchmark solution to the linear, time-dependent multiregion isothermal slug flow fission-product plateout problem

    Energy Technology Data Exchange (ETDEWEB)

    Durkee, J.W. Jr.; Lee, C.E. (Texas A and M Univ., College Station (USA). Dept. of Nuclear Engineering)

    1984-01-01

    The analytic solution to the time-dependent, linear, convective-diffusion equation with radioactive decay is derived for axisymmetric slug flow through multiple materials in a cylindrical pipe under isothermal conditions. The Davidson variable metric minimization algorithm is used to determine the coupling coefficients: These solutions, which describe the transport of fission products in a flowing stream, are then used to determine the concentration of radioactive material deposited on a conduit wall and on the adsorber materials using a standard mass-transfer model.

  6. Heated uranium tetrafluoride target system to release non-rare gas fission products for the TRISTAN isotope separator

    International Nuclear Information System (INIS)

    Off-line experiments indicated that fluorides of As, Se, Br, Kr, Zr, Nb, Mo, Tc, Ru, Sb, Te, I and Xe could be volatilized, but except for Br, Kr, I and Xe, none of these elements were observed after mass separation in the on-line experiments. The results of the on-line experiments indicated a very low level of hydride contamination at ambient temperature and consequently, uranium tetrafluoride replaced uranyl stearate as the primary gaseous fission product target. Possible reasons for the failure of the heated target system to yield non-rare gas activities are discussed and suggestions for designing a new heated target system are presented

  7. The methodical substantiation of measures to improve the use of basic production assets of the enterprise

    Directory of Open Access Journals (Sweden)

    Korol Svetlana Anatolevna

    2013-07-01

    Full Text Available The method of calculating the performance measures to improve the use of basic production assets of the enterprise: production cost, production volume, number of additional workers, depreciation and amortization.

  8. Comparison of predicted and measured fission product behavior in the Fort St. Vrain HTGR during the first three cycles of operation

    International Nuclear Information System (INIS)

    Fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors which is consistent with plateout probe measurements

  9. Spontaneous fission

    International Nuclear Information System (INIS)

    Recent experimental results for spontaneous fission half-lives and fission fragment mass and kinetic-energy distributions and other properties of the fragments are reviewed and compared with recent theoretical models. The experimental data lend support to the existence of the predicted deformed shells near Z = 108 and N = 162. Prospects for extending detailed studies of spontaneous fission properties to elements beyond hahnium (element 105) are considered. (orig.)

  10. Summary Report: Glass-Ceramic Waste Forms for Combined Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Riley, Brian J.; Turo, Laura A.; Tang, Ming; Kossoy, Anna

    2011-09-23

    Glass-ceramic waste form development began in FY 2010 examining two combined waste stream options: (1) alkaline earth (CS) + lanthanide (Ln), and (2) + transition metal (TM) fission-product waste streams generated by the uranium extraction (UREX+) separations process. Glass-ceramics were successfully developed for both options however; Option 2 was selected over Option 1, at the conclusion of 2010, because Option 2 immobilized all three waste streams with only a minimal decrease in waste loading. During the first year, a series of three glass (Option 2) were fabricated that varied waste loading-WL (42, 45, and 50 mass%) at fixed molar ratios of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali both at 1.75. These glass-ceramics were slow cooled and characterized in terms of phase assemblage and preliminary irradiation stability. This fiscal year, further characterization was performed on the FY 2010 Option 2 glass-ceramics in terms of: static leach testing, phase analysis by transmission electron microscopy (TEM), and irradiation stability (electron and ion). Also, a new series of glass-ceramics were developed for Option 2 that varied the additives: Al{sub 2}O{sub 3} (0-6 mass%), molar ratio of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali (1.75 to 2.25) and waste loading (50, 55, and 60 mass%). Lastly, phase pure powellite and oxyapatite were synthesized for irradiation studies. Results of this fiscal year studies showed compositional flexibility, chemical stability, and radiation stability in the current glass-ceramic system. First, the phase assemblages and microstructure of all of the FY 2010 and 2011 glass-ceramics are very similar once subjected to the slow cool heat treatment. The phases identified in these glass-ceramics were oxyapatite, powellite, cerianite, and ln-borosilicate. This shows that variations in waste loading or additives can be accommodated without drastically changing the phase assemblage of the waste form, thus making the processing and performance

  11. Properties of the platinoid fission products during vitrification of high-level radioactive waste

    Science.gov (United States)

    Gong, W.; Lutze, W.; Perez-Cardenas, F.; Matlack, K. S.; Pegg, I. L.

    2006-05-01

    Platinoid fission products present in high-level nuclear wastes present particular challenges to their treatment by vitrification. The platinoid metals Ru, Rh, Pd, and their compounds are sparingly soluble in borosilicate glass melts. During glass melting under oxidizing conditions, the platinoids form small crystals of highly dense solid intermetallic phases and oxides. Under reducing conditions, the platinoids form only intermetallic phases. A fraction of these crystals settles to the bottom of the melting furnace, forming an immobile sludge. The fraction settling reported in the literature is highly variable. In the present work, the fraction settling was found to be >90% under reducing conditions but only 10 to 20% under oxidizing conditions. The thickness of the sludge layer depends on the volume fraction of platinoid crystals in the sludge, which is poorly known (typically ~0.06 under oxidizing conditions). Since the electrical conductivity of the sludge can be >10X that of the melt, in joule-heated melters the presence of such a layer can lead to diversion of the electric current, thereby compromising melter operability. The time to failure by this mechanism is clearly of practical importance. A variety of data are required in order to estimate the time to failure due to this mechanism and such data must be obtained under conditions representative of those in a full-size melting furnace. We have acquired such data using a melting furnace installed in our laboratory. This furnace is a one-third scale prototype of the system to be used for the vitrification of defense HLW at Hanford, WA. In the present work, simulated Hanford HLW material was combined with glass formers to produce a melter feed slurry that was then spiked with the platinoids. Over one thousand chemical and optical analyses were performed on hundreds of samples taken from the feed, various locations inside the furnace, the glass melt during pouring, the solid glass, and various locations along

  12. Status of fission yield evaluations

    International Nuclear Information System (INIS)

    Very few yield compilations are also evaluations, and very few contain an extensive global library of measured data and extensive models for unmeasured data. The earlier U.K. evaluations and US evaluations were comparable up to the retirements of the primary evaluators. Only the effort in the US has been continued and expanded. The previous U.K. evaluations have been published. In this paper we summarize the current status of the US evaluation, philosophy, and various integral yield tests for 34 fissioning nuclides at one or more neutron incident energies and/or for spontaneous fission. Currently there are 50 yield sets and for each we have independent and cumulative yields and uncertainties for approximately 1100 fission products. When finalized, the recommended data will become part of the next version of the US Evaluated Nuclear Data File (ENDF/B-VI). The complete set of data, including the basic input of measured yields, will be issued as a sequel to the General Electric evaluation reports (better known by the authors' names: Rider - or earlier - Meek and Rider). 16 references

  13. Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor.

    Science.gov (United States)

    Košťál, Michal; Švadlenková, Marie; Koleška, Michal; Rypar, Vojtěch; Milčák, Ján

    2015-11-01

    Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins.

  14. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Science.gov (United States)

    Chadwick, M. B.; Herman, M.; Obložinský, P.; Dunn, M. E.; Danon, Y.; Kahler, A. C.; Smith, D. L.; Pritychenko, B.; Arbanas, G.; Arcilla, R.; Brewer, R.; Brown, D. A.; Capote, R.; Carlson, A. D.; Cho, Y. S.; Derrien, H.; Guber, K.; Hale, G. M.; Hoblit, S.; Holloway, S.; Johnson, T. D.; Kawano, T.; Kiedrowski, B. C.; Kim, H.; Kunieda, S.; Larson, N. M.; Leal, L.; Lestone, J. P.; Little, R. C.; McCutchan, E. A.; MacFarlane, R. E.; MacInnes, M.; Mattoon, C. M.; McKnight, R. D.; Mughabghab, S. F.; Nobre, G. P. A.; Palmiotti, G.; Palumbo, A.; Pigni, M. T.; Pronyaev, V. G.; Sayer, R. O.; Sonzogni, A. A.; Summers, N. C.; Talou, P.; Thompson, I. J.; Trkov, A.; Vogt, R. L.; van der Marck, S. C.; Wallner, A.; White, M. C.; Wiarda, D.; Young, P. G.

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range

  15. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M.B.; Herman, M.; Author(s): Chadwick,M.B.; Herman,M.; Oblozinsky,P.; Dunn,M.E.; Danon,Y.; Kahler,A.C.; Smith,D.L.; Pritychenko,B.; Arbanas,G.; Arcilla,R.; Brewer,R.; Brown,D.A.; Capote,R.; Carlson,A.D.; Cho,Y.S.; Derrien,H.; Guber,K.; Hale,G.M.; Hoblit,S.; Holloway,S.: Johnson,T.D.; Kawano,T.; Kiedrowski,B.C.; Kim,H.; Kunieda,S.; Larson,N.M.; Leal,L.; Lestone,J.P.; Little,R.C.; McCutchan,E.A.; MacFarlane,R.E.; MacInnes,M.; Mattoon,C.M.; McKnight,R.D.; Mughabghab,S.F.; Nobre,G.P.A.; Palmiotti,G.; Palumbo,A.; Pigni,M.T.; Pronyaev,V.G.; Sayer,R.O.; Sonzogni,A.A.; Summers,N.C.; Talou,P.; Thompson,I.J.; Trkov,A.; Vogt,R.L.; van der Marck,S.C.; Wallner,A.; White,M.C.; Wiarda,D.; Young,P.G.

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides {sup 235,238}U and {sup 239}Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on {sup 239}Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0

  16. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M. B. [Los Alamos National Laboratory (LANL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Oblozinsky, Pavel [Brookhaven National Laboratory (BNL); Dunn, Michael E [ORNL; Danon, Y. [Rensselaer Polytechnic Institute (RPI); Kahler, A. [Los Alamos National Laboratory (LANL); Smith, Donald L. [Argonne National Laboratory (ANL); Pritychenko, B [Brookhaven National Laboratory (BNL); Arbanas, Goran [ORNL; Arcilla, r [Brookhaven National Laboratory (BNL); Brewer, R [Los Alamos National Laboratory (LANL); Brown, D A [Brookhaven National Laboratory (BNL); Capote, R. [International Atomic Energy Agency (IAEA); Carlson, A. D. [National Institute of Standards and Technology (NIST); Cho, Y S [Korea Atomic Energy Research Institute; Derrien, Herve [ORNL; Guber, Klaus H [ORNL; Hale, G. M. [Los Alamos National Laboratory (LANL); Hoblit, S [Brookhaven National Laboratory (BNL); Holloway, Shannon T. [Los Alamos National Laboratory (LANL); Johnson, T D [Brookhaven National Laboratory (BNL); Kawano, T. [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Kim, H [Korea Atomic Energy Research Institute; Kunieda, S [Los Alamos National Laboratory (LANL); Larson, Nancy M [ORNL; Leal, Luiz C [ORNL; Lestone, J P [Los Alamos National Laboratory (LANL); Little, R C [Los Alamos National Laboratory (LANL); Mccutchan, E A [Brookhaven National Laboratory (BNL); Macfarlane, R E [Los Alamos National Laboratory (LANL); MacInnes, M [Los Alamos National Laboratory (LANL); Matton, C M [Lawrence Livermore National Laboratory (LLNL); Mcknight, R D [Argonne National Laboratory (ANL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Nobre, G P [Brookhaven National Laboratory (BNL); Palmiotti, G [Idaho National Laboratory (INL); Palumbo, A [Brookhaven National Laboratory (BNL); Pigni, Marco T [ORNL; Pronyaev, V. G. [Institute of Physics and Power Engineering (IPPE), Obninsk, Russia; Sayer, Royce O [ORNL; Sonzogni, A A [Brookhaven National Laboratory (BNL); Summers, N C [Lawrence Livermore National Laboratory (LLNL); Talou, P [Los Alamos National Laboratory (LANL); Thompson, I J [Lawrence Livermore National Laboratory (LLNL); Trkov, A. [Jozef Stefan Institute, Slovenia; Vogt, R L [Lawrence Livermore National Laboratory (LLNL); Van der Marck, S S [Nucl Res & Consultancy Grp, Petten, Netherlands; Wallner, A [University of Vienna, Austria; White, M C [Los Alamos National Laboratory (LANL); Wiarda, Dorothea [ORNL; Young, P C [Los Alamos National Laboratory (LANL)

    2011-01-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He; Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl; K; Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides (235,238)U and (239)Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es; Fm; and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on (239)Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide

  17. Practical limitations for the release of fission products during the operation of a research reactor: a case study of BR2

    International Nuclear Information System (INIS)

    Failure of the cladding of a fuel element is an event occurring from time to time while operating a research reactor. As a consequence, fission products are released in the primary circuit of the reactor. This contamination means no direct hazard for the workers or for the environment in case the reactor has a closed primary circuit. The operator can decide to continue the irradiation to finish a scientific experiment or a commercial isotope production program. However, the operator cannot prolong the cycle regardless the concentration fission products in the primary loop. Beside the limitations imposed by the regulatory authorities, ALARA considerations should be taken into account. An untimely stop of the reactor can have serious financial consequences and prolonged operation causes higher radiation doses. This paper gives an overview of decision process applied in case of detection of fission products in the primary circuit of BR2. (author)

  18. Practical limitations for the release of fission products during the operation of a research reactor: a case study of BR2

    Energy Technology Data Exchange (ETDEWEB)

    Joppen, F. [Health Physics and Safety Department, SCK-CEN, B-2400 Mol (Belgium)

    1998-07-01

    Failure of the cladding of a fuel element is an event occurring from time to time while operating a research reactor. As a consequence, fission products are released in the primary circuit of the reactor. This contamination means no direct hazard for the workers or for the environment in case the reactor has a closed primary circuit. The operator can decide to continue the irradiation to finish a scientific experiment or a commercial isotope production program. However, the operator cannot prolong the cycle regardless the concentration fission products in the primary loop. Beside the limitations imposed by the regulatory authorities, ALARA considerations should be taken into account. An untimely stop of the reactor can have serious financial consequences and prolonged operation causes higher radiation doses. This paper gives an overview of decision process applied in case of detection of fission products in the primary circuit of BR2. (author)

  19. Investigations on the gettering of metallic fission products in the primary circuit of a high temperature reactor

    International Nuclear Information System (INIS)

    A new concept of gettering Ag-110 m and Cs-134 137 in the primary circuit of a High Temperature Reactor (HTR) is presented. It is based upon the known fact that the vapor pressure of metallic fission products in solid or liquid solutions is lower compared with that of the pure fission products. Although metallic additives were found not to influence the silver release from oxide fuel kernels, the effective diffusion coefficient of Ag-110 m in graphite matrix is reduced by about two orders of magnitude by small amounts of the metallic Cu, Ge, Sn or Au additions. However, these reduced silver diffusion coefficients are not sufficiently low in order to retain Ag-110 m in the fuel-free zone of spherical HTR fuel elements. On the other hand, metallic additives were found to be very efficient in gettering Ag-110 m from the gaseous pahse: During a contact time of only 0.15 seconds at 950sup(o)C more than 80%, at 850sup(o)C even more than 99% of the Ag-110 m could be absorbed from the streaming gas by using a metal-containing graphite filter. The best results were obtained by using Sn or Au additives. By optimizing the filter geometry further increase of the efficiency should be possible. (orig./HP)

  20. Cost analysis for application of solidified waste fission product canisters in U.S. Army steam plants

    International Nuclear Information System (INIS)

    The main objectives of the present study are to design steam plants using projected waste fission product canister characteristics, to analyze the overall impact and cost/benefit to the nuclear fuel cycle associated with these plants, and to develop plans for this application if the cost analysis so warrants it. The construction and operation of a steam plant fueled with waste fission product canisters would require the involvement and cooperation of various government agencies and private industry; thus the philosophies of these groups were studied. These philosophies are discussed, followed by a forecast of canister supply, canister characteristics, and strategies for Army canister use. Another section describes the safety and licensing of these steam plants since this affects design and capital costs. The discussion of steam plant design includes boiler concepts, boiler heat transfer, canister temperature distributions, steam plant size, and steam plant operation. Also, canister transportation is discussed since this influences operating costs. Details of economics of Army steam plants are provided including steam plant capital costs, operating costs, fuel reprocessor savings due to Army canister storage, and overall economics. Recommendations are made in the final section

  1. Separation of actinides and long-lived fission products from high-level radioactive wastes (a review)

    International Nuclear Information System (INIS)

    The management of high-level radioactive wastes is facilitated, if long-lived and radiotoxic actinides and fission products are separated before the final disposal. Especially important is the separation of americium, curium, plutonium, neptunium, strontium, cesium and technetium. The separated nuclides can be deposited separately from the bulk of the high-level waste, but their transmutation to short-lived nuclides is a muchmore favourable option. This report reviews the chemistry of the separation of actinides and fission products from radioactive wastes. The composition, nature and conditioning of the wastes are described. The main attention is paid to the solvent extraction chemistry of the elements and to the application of solvent extraction in unit operations of potential partitioning processes. Also reviewed is the behaviour of the elements in the ion exchange chromatography, precipitation, electrolysis from aqueous solutions and melts, and the distribution between molten salts and metals. Flowsheets of selected partitioning processes are shown and general aspects of the waste partitioning are shortly discussed. (orig.)

  2. Analytic benchmark solution to the linear, time-dependent isothermal laminar flow fission-product plateout problem

    Energy Technology Data Exchange (ETDEWEB)

    Durkee, J.W. Jr.; Lee, C.E. (Texas A and M Univ., College Station (USA). Dept. of Nuclear Engineering)

    1984-01-01

    The time-dependent convective-diffusion equation with radioactive decay is solved analytically in axisymmetric cylindrical geometry for laminar and slug velocity profiles under isothermal conditions. Concentration-dependent diffusion is neglected. The laminar flow solution is derived using the method of separation of variables and Frobenius' technique for constructing a series expansion about a regular singular point. These solutions, which describe the transport of fission products in a flowing stream, are then used to determine the pointwise and integrated concentrations of radioactive material deposited on a conduit wall using a standard mass-transfer model. Extensive parametric investigations have been conducted by varying the wall mass-transfer coefficient, diffusion coefficient, flow velocity, pipe radius and species half-life in the deposition models. The computational results indicate that the plateout estimates for the slug flow model are typically 5-100% greater than for the laminar model. The effect of axial diffusion is necessarily negligible for Peclet numbers greater than 100. Little increased plateout is observed for Peclet numbers less than 100; an additional 8% is predicted for a Peclet number of 20 if axial diffusion is included. Characteristic stream, wall and integrated deposition profiles are shown. Representative results from the analysis of fission-product deposition measurements for diffusion tubes in the Fort St. Vrain high-temperature gas-cooled reactor plateout probe are presented.

  3. Release behavior of fission products from coated fuel particles during post-irradiation heating at abnormally high temperatures

    International Nuclear Information System (INIS)

    The present report describes experimental results on release behavior of metal fission-products during isochronal and isothermal heating of TRISO-and BISO-coated fuel particles at temperatures from 1400 to 2200degC. The particles were irradiated in the eighth OGL-1 fuel assembly at 1300 - 1400degC for 53.8 days up to burnups around 1% fima. Fuel particles without failed or defective coating layer in X-ray radiography were heated up one by one in a graphite electrode, whose temperature was measured by a pyrometer. The fractional releases of 137Cs from the TRISO-coated fuel particles were only around one percent, even after isothermal heating at temperatures from 1600 to 2000degC for several hundred hours. Effective retention of 137Cs by the silicon-carbide layer was observed below 2000degC through comparison of its fractional releases from the TRISO-and BISO-coated fuel particles. Above 2050degC, however, fractional releases of metal fission-products from the TRISO-coated particles increased rapidly, due to thermal degradation of the SiC layer. (author)

  4. Distribution of fission products in graphite sleeves and blocks of the ninth and tenth OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Distribution of fission products in graphite sleeves and blocks of the ninth and tenth OGL-1 fuel assemblies was measured by gamma spectrometry with lathe sectioning. The assemblies were loaded with HTGR fuel compacts, which had been produced by a scaled-up facility for the High Temperature Engineering Test Reactor (HTTR) being developed by JAERI; and they were irradiated in an in-pile gas loop, OGL-1. Fission products detected in the sleeves were 137Cs, 134Cs, 155Eu, 154Eu, 144Ce, 125Sb and 110mAg. The last nuclide, however, may have been produced by activation of a stable isotope, 109Ag, contained as impurity. Effective retention capability of the sleeve was observed for 155Eu, 154Eu, 144Ce and 125Sb; while, not for 137Cs and 134Cs. Concentration of 137Cs in the graphite blocks was markedly higher at the downstream side than at the upstream side of the coolant. This was ascribed to migration of the nuclide with the coolant flow and its subsequent sorption on the surface of the block. (author)

  5. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  6. Measurement of fission products β decay properties using a total absorption spectrometer

    Science.gov (United States)

    Zakari-Issoufou, A.-A.; Porta, A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Agramunt, J.; Äystö, J.; Bowry, M.; Bui, V. M.; Caballero-Folch, R.; Cano-Ott, D.; Eloma, V.; Estévez, E.; Farrelly, G. F.; Garcia, A.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez, A.; Podolyak, Zs.; Penttilä, H.; Regan, P. H.; Rissanen, J.; Rubio, B.; Weber, C.

    2013-12-01

    In a nuclear reactor, the β decay of fission fragments is at the origin of decay heat and antineutrino flux. These quantities are not well known while they are very important for reactor safety and for our understanding of neutrino physics. One reason for the discrepancies observed in the estimation of the decay heat and antineutrinos flux coming from reactors could be linked with the Pandemonium effect. New measurements have been performed at the JYFL facility of Jyväskylä with a Total Absorption Spectrometer (TAS) in order to circumvent this effect. An overview of the TAS technique and first results from the 2009 measurement campaign will be presented.

  7. Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

    Directory of Open Access Journals (Sweden)

    Jaroszewicz Janusz

    2014-07-01

    Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

  8. Exact Solution of Fractional Diffusion Model with Source Term used in Study of Concentration of Fission Product in Uranium Dioxide Particle*

    Institute of Scientific and Technical Information of China (English)

    FANG Chao; CAO Jian-Zhu; SUN Li-Feng

    2011-01-01

    The exact solution of fractional diffusion model with a location-independent source term used in the study of the concentration of fission product in spherical uranium dioxide (U02) particle is built. The adsorption effect of the fission product on the surface of the U02 particle and the delayed decay effect are also considered. The solution is given in terms of Mittag-Leffler function with finite Hankel integral transformation and Laplace transformation. At last, the reduced forms of the solution under some special physical conditions, which is used in nuclear engineering, are obtained and corresponding remarks are given to provide significant exact results to the concentration analysis of nuclear fission products in nuclear reactor.

  9. Measurements of extinct fission products in nuclear bomb debris: Determination of the yield of the Trinity nuclear test 70 y later.

    Science.gov (United States)

    Hanson, Susan K; Pollington, Anthony D; Waidmann, Christopher R; Kinman, William S; Wende, Allison M; Miller, Jeffrey L; Berger, Jennifer A; Oldham, Warren J; Selby, Hugh D

    2016-07-19

    This paper describes an approach to measuring extinct fission products that would allow for the characterization of a nuclear test at any time. The isotopic composition of molybdenum in five samples of glassy debris from the 1945 Trinity nuclear test has been measured. Nonnatural molybdenum isotopic compositions were observed, reflecting an input from the decay of the short-lived fission products (95)Zr and (97)Zr. By measuring both the perturbation of the (95)Mo/(96)Mo and (97)Mo/(96)Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the (95)Zr and (97)Zr isotopes formed in the nuclear detonation. Together with a determination of the amount of plutonium in the debris, these measurements of extinct fission products allow for new estimates of the efficiency and yield of the historic Trinity test.

  10. Model uncertainty in deterministic safety analysis: the influence of horizontal feeders nodalization on fission products masses, deposited in CANDU primary circuit

    International Nuclear Information System (INIS)

    The paper presents the influence of horizontal feeders' nodalization on fission products masses deposited in CANDU primary circuit, during a postulated severe accident induced by Loss of Coolant Accident (LOCA) followed by Loss of Emergency Core Cooling (LECC). The feeders length has been divided in 5, 4, 3 and 2 nodes; for each situation, the deposited masses in CANDU primary circuit for the most important fission products (for source term) were calculated by using SOPHAEROS computer code and the transferred masses to the containment, as well. Based on these results, the relations which represent the nodes numbers influence on fission products masses deposited into CANDU primary circuit, have been established. The influence of the nodalization is analyzed in detail and recommendations for users were formulated. (authors)

  11. Experimental determination of the antineutrino spectrum of the fission products of U238

    International Nuclear Information System (INIS)

    Accurate predictions of the antineutrino spectrum emitted by a nuclear reactor are of paramount importance for current and future reactor neutrino experiments. The antineutrinos are produced in the β - decays of the fission daughters of the four main fuel isotopes 235U, 238U, 239Pu, and 241Pu. One way to calculate the total anti νe - spectrum emitted by a fuel assembly is to experimentally determine the cumulative β-spectra emitted after fission of these four main fuel isotopes and to convert these into the corresponding anti νe-spectra. Three of the four spectra could already be determined in the 1980's, but only recently an experiment at the scientific neutron source FRM II in Garching could be performed to measure the anti νe-spectrum of 238U which contributes 10 % to the total antineutrino output of a standard PWR. With this spectrum, it is now possible to predict the antineutrino output of a reactor without the use of theoretical calculations for the contributing spectra. This talk describes the results of the experiment and discusses the impact on the current analysis of reactor neutrino experiments and the reactor antineutrino anomaly, which may give a hint on the possible existence of light sterile neutrinos.

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

    International Nuclear Information System (INIS)

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

  13. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Susmikanti, Mike, E-mail: mike@batan.go.id [Center for Development of Nuclear Informatics, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Dewayatna, Winter, E-mail: winter@batan.go.id [Center for Nuclear Fuel Technology, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Sulistyo, Yos, E-mail: soj@batan.go.id [Center for Nuclear Equipment and Engineering, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia)

    2014-09-30

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo{sup 99} used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 10{sup 6} cm{sup −1}) in a tube, their delta

  14. Standard test method for gamma energy emission from fission products in uranium hexafluoride and uranyl nitrate solution

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This test method covers the measurement of gamma energy emitted from fission products in uranium hexafluoride (UF6) and uranyl nitrate solution. It is intended to provide a method for demonstrating compliance with UF6 specifications C 787 and C 996 and uranyl nitrate specification C 788. 1.2 The lower limit of detection is 5000 MeV Bq/kg (MeV/kg per second) of uranium and is the square root of the sum of the squares of the individual reporting limits of the nuclides to be measured. The limit of detection was determined on a pure, aged natural uranium (ANU) solution. The value is dependent upon detector efficiency and background. 1.3 The nuclides to be measured are106Ru/ 106Rh, 103Ru,137Cs, 144Ce, 144Pr, 141Ce, 95Zr, 95Nb, and 125Sb. Other gamma energy-emitting fission nuclides present in the spectrum at detectable levels should be identified and quantified as required by the data quality objectives. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its us...

  15. TRANCS, a computer code for calculating fission product release from high temperature gas-cooled reactor fuel, (1)

    International Nuclear Information System (INIS)

    The computer program, TRANCS, has been developed for evaluating the fractional release of long-lived fission products from coated fuel particles. This code numerically gives the non-stationary solution of the diffusion equation with birth and decay terms. The birth term deals with the fissile material in the fuel kernel, the contamination in the coating layers and the fission-recoil transfer from the kernel into the buffer layer; and the decay term deals with effective decay not only due to beta decay but also due to neutron capture, if appropriate input data are given. The code calculates the concentration profile, the release to birth rates (R/B), and the release and residual fractions in the coated fuel particle. Results obtained numerically have been in good agreement with the corresponding analytical solutions after the Booth model. Thus, the validity of the present code was confirmed, and further undate of the code has been discussed for extention of its computation scopes and models. (author)

  16. A new evaluation of fission product yields and the production of a new library (UKFY2) of independent and cumulative yields. Pt. 3

    International Nuclear Information System (INIS)

    A new evaluation has been prepared of the independent and cumulative yields of the products of fission induced by thermal, fast, and 14 MeV neutrons in nuclides important for reactor design and operation and for fuel and waste management. Three spontaneously fissioning nuclides were also considered. The evaluation used a database that is considered to be complete up to early 1989. Careful study was made of experimental uncertainties and discrepancies, emphasising the need for further measurements. Gaps in the data were filled by interpolation and extrapolation, using fits to empirical models. The yields were adjusted to fit physical constraints of the fissioning process. The present report contains lists of chain and independent yields for which the experimental data may be regarded as inadequate. An entry is made if (i) there are no measurements, (ii) there is only one measurement, or (iii) there are several measurements but they are discrepant in the sense that the probability of the calculated value of χ2 arising by chance is less than 10%. So that the source of each measurement may be identified, the database reference list is included as an Appendix. (author)

  17. Investigation of fission product retention by a HTR containment building designed according to the vented confinement concept. Technical report 1.7

    International Nuclear Information System (INIS)

    The investigations cover the following work: For a defined core heat-up accident in the THTR-300, up-dated design-basis data describing the accident scenario and the release of fission products from the primary loop are chosen to establish a model of the thermohydraulics of the cooling gas escape, and to simulate the cooling gas diffusion into the building and the environment. The retention of fission product aerosols and gas by the containment building is calculated. The retaining capacity for aerosols is calculated to be 75-96%, and that for iodine 95-99%. (orig.)

  18. UO2 oxidation in air or steam-release or retention of the fission products Ru, Ba, Ce, Eu, Sb and Nb

    International Nuclear Information System (INIS)

    The release of fission products from irradiated UO2 has been studied under both inert (argon) and oxidizing (air and steam) conditions at temperatures up to about 1600 degrees Celsius. The microstructures resulting from post-irradiation annealing and oxidation are presented and the release and retention of some of the metallic fission products are documented. Ruthenium and antimony are the main metallic elements which are released and ruthenium is only released in the presence of oxygen. Cerium, because of its low energy gamma-ray provides an excellent indicator of density changes in the fuel during oxidation

  19. Fission yield measurements at IGISOL

    Science.gov (United States)

    Lantz, M.; Al-Adili, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Mattera, A.; Moore, I.; Penttilä, H.; Pomp, S.; Prokofiev, A. V.; Rakopoulos, V.; Rinta-Antila, S.; Simutkin, V.; Solders, A.

    2016-06-01

    The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL) technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f) and Th(p,f) have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn) reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  20. Fission yield measurements at IGISOL

    Directory of Open Access Journals (Sweden)

    Lantz M.

    2016-01-01

    Full Text Available The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f and Th(p,f have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  1. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    Energy Technology Data Exchange (ETDEWEB)

    Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2004-10-18

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

  2. Measurement of fission products β decay properties using a total absorption spectrometer

    Directory of Open Access Journals (Sweden)

    Zakari-Issoufou A.-A.

    2013-12-01

    Full Text Available In a nuclear reactor, the β decay of fission fragments is at the origin of decay heat and antineutrino flux. These quantities are not well known while they are very important for reactor safety and for our understanding of neutrino physics. One reason for the discrepancies observed in the estimation of the decay heat and antineutrinos flux coming from reactors could be linked with the Pandemonium effect. New measurements have been performed at the JYFL facility of Jyväskylä with a Total Absorption Spectrometer (TAS in order to circumvent this effect. An overview of the TAS technique and first results from the 2009 measurement campaign will be presented.

  3. Delayed neutron spectra and their uncertainties in fission product summation calculations

    Energy Technology Data Exchange (ETDEWEB)

    Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)

    1997-03-01

    Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)

  4. Prompt Neutrons from Fission

    International Nuclear Information System (INIS)

    -number fragments. On the basis of a simple fragment-deformation theory, the deformation parameter is calculated directly from experimental data, and is seen to have a very similar dependence on mass for four types of fission. These ideas seem likely to lead to a more basic understanding of the fission process, including the mass yields and energies of the fragments. (author)

  5. Synthesis report on the relevant diffusion coefficients of fission products and helium in spent nuclear fuels; Rapport de synthese sur les coefficients de diffusion des produits de fission et de l'helium dans le combustible irradie

    Energy Technology Data Exchange (ETDEWEB)

    Lovera, P.; Ferry, C.; Poinssot, Ch. [CEA Saclay, Dept. de Physico-Chimie (DPC), 91 - Gif sur Yvette (France); Johnson, L. [Nagra, Baden (Switzerland)

    2003-07-01

    This document corresponds to the deliverable D2 of the Work Package 1 of the 'Spent Fuel Stability under repository conditions' (SFS) European project. It constitutes a synthesis report on relevant diffusion coefficients of fission products and helium in spent nuclear fuels at high and low temperatures. Coefficients corresponding to thermally activated diffusion were reviewed from literature data for O, U (self-diffusion coefficients), fission gases and other fission products. Data showed that thermal diffusion was irrelevant at temperatures expected in repository conditions. The occurrence of diffusion enhanced by alpha self-irradiation was studied through different theoretical approaches. A 'best estimate' value of the alpha self-irradiation diffusion coefficient, D (m{sup 2}.s{sup -1}), is proposed. It is extrapolated from enhanced diffusion under irradiation observed in reactor and would be proportional to the volume alpha activity in the spent nuclear fuel, A{sub v} (Bq.m{sup -3}) as: D/A{sub v} {approx_equal} 2.10{sup -41} (m{sup 5})The migration of stable Pb in Oklo's uraninites was studied in order to validate the proposed diffusion coefficient. The obtained value is one order of magnitude higher than the theoretical proposed value. As for He behaviour in spent nuclear fuel, a few data are today available in open literature. The document will be completed as soon as new experimental results are available. (authors)

  6. The maximum entropy production principle: two basic questions

    OpenAIRE

    Martyushev, Leonid M.

    2010-01-01

    The overwhelming majority of maximum entropy production applications to ecological and environmental systems are based on thermodynamics and statistical physics. Here, we discuss briefly maximum entropy production principle and raises two questions: (i) can this principle be used as the basis for non-equilibrium thermodynamics and statistical mechanics and (ii) is it possible to ‘prove’ the principle? We adduce one more proof which is most concise today.

  7. Thermal transport in UO2 with defects and fission products by molecular dynamics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cooper, Michael William Donald [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lashley, Jason Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Byler, Darrin David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-14

    The importance of the thermal transport in nuclear fuel has motivated a wide range of experimental and modelling studies. In this report, the reduction of thermal transport in UO2 due to defects and fission products has been investigated using non-equilibrium MD simulations, with two sets of empirical potentials for studying the degregation of UO2 thermal conductivity including a Buckingham type interatomic potential and a recently developed EAM type interatomic potential. Additional parameters for U5+ and Zr4+ in UO2 have been developed for the EAM potential. The thermal conductivity results from MD simulations are then corrected for the spin-phonon scattering through Callaway model formulations. To validate the modelling results, comparison was made with experimental measurements on single crystal hyper-stoichiometric UO2+x samples.

  8. Determination of burnup balance for nuclear reactor fuel on the basis of γ-spectrometric determination of fission products

    International Nuclear Information System (INIS)

    Results are given of experimental investigations in one of the versions of the method for determination of the balance of nuclear fuel burnup process by means of the γ-spectrometry of fission products. In the version being considered a balance of the burnup process was determined on the base of 106Ru, 134Cs.Activity was measured by means of a γ-spectrometer with Ge counter. Investigations were done on the natural uranium metal fuel from the heavy-water moderated reactor of the first Czechoslovakian nuclear power plant A1 in Yaslovske Bohunice. Possibility was checked of determination of the fuel burnup depth as well as of the isotope ratio and content of plutonium. Results were compared with the control data which had been obtained on the base of the mass-spectrometry of U, Pu and Nd. The reasors for deviations were estimated in the cases when they were greater tan error in the control data

  9. Automated analysis for large amount gaseous fission product gamma-scanning spectra from nuclear power plant and its data mining

    International Nuclear Information System (INIS)

    Based on the Linssi database and UniSampo/Shaman software, an automated analysis platform has been setup for the analysis of large amounts of gamma-spectra from the primary coolant monitoring systems of a CANDU reactor. Thus, a database inventory of gaseous and volatile fission products in the primary coolant of a CANDU reactor has been established. This database is comprised of 15,000 spectra of radioisotope analysis records. Records from the database inventory were retrieved by a specifically designed data-mining module and subjected to further analysis. Results from the analysis were subsequently used to identify the reactor coolant half-life of 135Xe and 133Xe, as well as the correlations of 135Xe and 88Kr activities. (author)

  10. Comparison of Computational Estimations of Reactivity Margin From Fission Products and Minor Actinides in PWR Burnup Credit

    International Nuclear Information System (INIS)

    This paper has presented the results of a computational benchmark and independent calculations to verify the benchmark calculations for the estimation of the additional reactivity margin available from fission products and minor actinides in a PWR burnup credit storage/transport environment. The calculations were based on a generic 32 PWR-assembly cask. The differences between the independent calculations and the benchmark lie within 1% for the uniform axial burnup distribution, which is acceptable. The Δk for KENO - MCNP results are generally lower than the other Δk values, due to the fact that HELIOS performed the depletion part of the calculation for both the KENO and MCNP results. The differences between the independent calculations and the benchmark for the non-uniform axial burnup distribution were within 1.1%

  11. Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products.

    Science.gov (United States)

    Košťál, Michal; Švadlenková, Marie; Milčák, Ján; Rypar, Vojtěch; Koleška, Michal

    2014-07-01

    The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well.

  12. A calculation on the diffusion and release behavior of fission products in the first and second OGL-1 fuel sleeves

    International Nuclear Information System (INIS)

    Based on the Fick's law the computer program FPDR has been developed to calculate the one-dimensional diffusion and release behavior of fission products in the graphite sleeves of the first and second OGL-1 fuel assembly. Through the comparison between the measured and calculated penetration profiles, the diffusion coefficient of 90Sr in the first fuel sleeve has been estimated to be (2 -- 5) x 10-13 m2/s; those of 137Cs and 90Sr in the second fuel sleeve around or larger than 1 x 10-12 m2/s, and --10-14 m2/s, respectively. The release of 90Sr from the second fuel sleeve is negligible; that of 137Cs depends linearly on its diffusion coefficient if the coefficient is larger than 10-12 m2/s, but practically does not depend on its evaporation parameter. (author)

  13. The effect of rare-earth fission products on the rate of U3O8 formation on UO2

    International Nuclear Information System (INIS)

    The rate of U3O8 formation on the surface of flat neodymium-doped UO2 disks was measured by X-ray diffraction and the kinetic data were fitted to a two-dimensional nucleation-and-growth model. The results indicate that neodymium doping in the UO2 tends to inhibit U3O8 formation. A quantitative relationship between the activation energy for U3O8 formation and the neodymium content of the UO2 has been derived from the kinetic data. Our data are consistent with recent results obtained for the oxidation of used LWR fuel, which suggests that fission products in solid solution are likely the cause of U3O8 inhibition observed for used fuel. (orig.)

  14. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  15. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  16. First burnup credit application including actinides and fission products for transport and storage cask by using French experiments

    International Nuclear Information System (INIS)

    The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C-E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA-AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TN's application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies. (author)

  17. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    International Nuclear Information System (INIS)

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses

  18. Low-energy ternary fission

    International Nuclear Information System (INIS)

    With the detector system DIOGENES thermal neutron induced and spontaneous α particle associated fission and spontaneous nuclear tripartition into three fragments of similar masses has been investigated. DIOGENES is a concentric arrangement of toroidal angular position sensitive ionization chambers and proportional counters to measure the kinetic energies and relative angular distributions of the three reaction products of ternary fission. For α-particle accompanied fission some of the many possible α particle fission-fragment parameter correlations will be discussed. For nearly symmetric low-energy nuclear tripartition new upper limits are presented. Former experimental results which pretended evidence for so called true ternary fission could be explained by charged-particle associated fission with a light particle in the mass range of 13 < A < 23

  19. Fission waves can oscillate

    CERN Document Server

    Osborne, Andrew G

    2016-01-01

    Under the right conditions, self sustaining fission waves can form in fertile nuclear materials. These waves result from the transport and absorption of neutrons and the resulting production of fissile isotopes. When these fission, additional neutrons are produced and the chain reaction propagates until it is poisoned by the buildup of fission products. It is typically assumed that fission waves are soliton-like and self stabilizing. However, we show that in uranium, coupling of the neutron field to the 239U->239Np->239Pu decay chain can lead to a Hopf bifurcation. The fission reaction then ramps up and down, along with the wave velocity. The critical driver for the instability is a delay, caused by the half-life of 239U, between the time evolution of the neutron field and the production of 239Pu. This allows the 239Pu to accumulate and burn out in a self limiting oscillation that is characteristic of a Hopf bifurcation. Time dependent results are obtained using a numerical implementation of a reduced order r...

  20. Microscopic dynamical description of proton-induced fission with the constrained molecular dynamics model

    Science.gov (United States)

    Vonta, N.; Souliotis, G. A.; Veselsky, M.; Bonasera, A.

    2015-08-01

    The microscopic description of nuclear fission still remains a topic of intense basic research. Understanding nuclear fission, apart from a theoretical point of view, is of practical importance for energy production and the transmutation of nuclear waste. In nuclear astrophysics, fission sets the upper limit to the nucleosynthesis of heavy elements via the r process. In this work we initiated a systematic study of intermediate-energy proton-induced fission using the constrained molecular dynamics (CoMD) code. The CoMD code implements an effective interaction with a nuclear matter compressibility of K =200 (soft equation of state) with several forms of the density dependence of the nucleon-nucleon symmetry potential. Moreover, a constraint is imposed in the phase-space occupation for each nucleon restoring the Pauli principle at each time step of the collision. A proper choice of the surface parameter of the effective interaction has been made to describe fission. In this work, we present results of fission calculations for proton-induced reactions on: (a) 232Th at 27 and 63 MeV; (b) 235U at 10, 30, 60, and 100 MeV; and (c) 238U at 100 and 660 MeV. The calculated observables include fission-fragment mass distributions, total fission energies, neutron multiplicities, and fission times. These observables are compared to available experimental data. We show that the microscopic CoMD code is able to describe the complicated many-body dynamics of the fission process at intermediate and high energy and give a reasonable estimate of the fission time scale. Sensitivity of the results to the density dependence of the nucleon symmetry potential (and, thus, the nuclear symmetry energy) is found. Further improvements of the code are necessary to achieve a satisfactory description of low-energy fission in which shell effects play a dominant role.