WorldWideScience

Sample records for basic fission products

  1. Basic actinide and fission products chemistry in the CEC-coordinated project: Migration of radionuclides in the geosphere (MIRAGE)

    International Nuclear Information System (INIS)

    The paper reviews the research works performed on the basic actinide (Am, Pu, Np) and fission product (Tc, Sr) chemistry by four CEC member laboratories under the project named 'MIRAGE' for the years 1983 and 1984. Research subjects dealt with are solubility, carbonate complexation, hydrolysis reaction, colloid generation, speciation methods and sorption phenomena. Important achievements are summarized and discussed for each subject separately. (orig.)

  2. 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)

  3. Fission product detection

    International Nuclear Information System (INIS)

    The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells

  4. 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)

  5. 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

  6. Fission product revaporization

    International Nuclear Information System (INIS)

    One of the major developmental advances in severe accident analysis since the Reactor Safety Study relates to the accounting for radionuclide retention in the reactor coolant system (RCS). The retention is predicted to occur as materials released during core heatup and degradation are transported through the RCS to the break (broken pipe, relief valve, etc.). For accidents involving relatively long RCS-transit times (e.g., station blackout in PWRs), the fraction of released material predicted to remain in the RCS can be large. For example, calculations for the Surry station blackout sequence showed retention of approximately 80% of the cesium and iodine species. Factors affecting fission product revaporization are post-vessel-failure thermal hydraulics, heat loss through vessel and pipe walls, and revaporization chemistry. The accident conditions relevant to this issue range from those present immediately after vessel failure to those present after containment failure. The factors that affect fission product revaporization are discussed

  7. Fission products in glasses. Pt. 2

    International Nuclear Information System (INIS)

    Glass ceramics of different composition with high leach and impact resistance can be produced for fission product solidification. In contrast to commercial glass products, they consist of a number of crystalline phases and a residual glass phase. The major crystalline phase allows a classification into celsian, diopside, encryptite, and perovskite ceramics. They all are of special importance as host phases for long-lived fission products. The paper reports on relations between product composition and melting properties, viscosity, crystallization properties, and fixation capability for fission products. Further investigations deal with dimensional stability, impact resistance, thermal expansion, and thermal conductivity. The properties of the ceramics are compared with those of the basic products. The problems still to be solved with regard to further improvement and application of these products are discussed. (RB)

  8. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    This is the tenth issue of a report series on Fission Product Data, which informs us about all the activities in this field, which are planned, ongoing, or have recently been completed. The types of activities included 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, lumped fission product data (decay heat, absorption, etc.). There is also a section with recent references relative to fission product nuclear data

  9. Fission product solvent extraction

    International Nuclear Information System (INIS)

    Two main objectives concerning removal of fission products from high-level tank wastes will be accomplished in this project. The first objective entails the development of an acid-side Cs solvent-extraction (SX) process applicable to remediation of the sodium-bearing waste (SBW) and dissolved calcine waste (DCW) at INEEL. The second objective is to develop alkaline-side SX processes for the combined removal of Tc, Cs, and possibly Sr and for individual separation of Tc (alone or together with Sr) and Cs. These alkaline-side processes apply to tank wastes stored at Hanford, Savannah River, and Oak Ridge. This work exploits the useful properties of crown ethers and calixarenes and has shown that such compounds may be economically adapted to practical processing conditions. Potential benefits for both acid- and alkaline-side processing include order-of-magnitude concentration factors, high rejection of bulk sodium and potassium salts, and stripping with dilute (typically 10 mM) nitric acid. These benefits minimize the subsequent burden on the very expensive vitrification and storage of the high-activity waste. In the case of the SRTALK process for Tc extraction as pertechnetate anion from alkaline waste, such benefits have now been proven at the scale of a 12-stage flowsheet tested in 2-cm centrifugal contactors with a Hanford supernatant waste simulant. SRTALK employs a crown ether in a TBP-modified aliphatic kerosene diluent, is economically competitive with other applicable separation processes being considered, and has been successfully tested in batch extraction of actual Hanford double-shell slurry feed (DSSF)

  10. 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)

  11. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    This is the 12th issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the 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 type of activities included 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 first part of the report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The second part contains recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences

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

    International Nuclear Information System (INIS)

    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 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission products from spontaneous fission of 252Cf. 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 252Cf spontaneous fission products are reported from an E–v measurement

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. Fission product transport in the atmosphere

    International Nuclear Information System (INIS)

    In the present thesis,a theoretical treatment has been developed for investigating the fission product transport in the atmosphere, and expressing its underlying dynamics. Some basic material has been critically reviewed for the purpose of establishing the mathematical equations that govern the motion of a general pollutant - whether it is radioactive or not - in the atmosphere. Such review included the study of particulate matter motion in the atmosphere and radioactive cloud motion, together with the analysis of available mathematical models for atmospheric dispersion

  19. 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

  20. 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)

  1. 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

  2. Downstream behavior of fission products

    International Nuclear Information System (INIS)

    The downstream behavior of fission products has been investigated by injecting mixtures of CsOH, CsI, and Te into a flowing steam/hydrogen stream and determining the physical and chemical changes that took place as the gaseous mixture flowed down a reaction duct on which a temperature gradient (10000 to 2000C) had been imposed. Deposition on the wall of the duct occurred by vapor condensation in the higher temperature regions and by aerosol deposition in the remainder of the duct. Reactions in the gas stream between CsOH and CsI and between CsOH and Te had an effect on the vapor condensation. The aerosol was characterized by the use of impingement tabs placed in the gas stream

  3. 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

  4. Calculation code of the fission products activity

    International Nuclear Information System (INIS)

    The document describes the two codes for the calculation of the fission products activity. The ''Pepin le bref'' code gives the exact value of the beta and gamma activities of completely known fission products. The code ''Plus Pepin'' introduces the beta and gamma activities whose properties are partially known. (A.L.B.)

  5. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    This is the seventh 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 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 sixth issue of this series has been published in June 1980 as INDC(NDS)-113/G+P. The present issue includes contributions which were received by NDS between 1 August 1980 and 25 May 1981

  6. 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

  7. Properties and detection of ionizing radiation resulting from instantaneous fission and fission product mixture

    International Nuclear Information System (INIS)

    The different types of ionizing radiation accompanying fission and mixtures of fission products, their activity, the determination of the age of fission products and the biological hazard of radiation caused by instantaneous fission are described. The possibility is described of detection, and of the dosimetry of ionizing radiation resulting from instantaneous fission and emitted by a mixture of fission products, the determination of the dose of neutron radiation, surface contamination, internal contamination and the contamination of water and foods. (J.P.)

  8. 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.)

  9. Actinide and fission product separation and transmutation

    International Nuclear Information System (INIS)

    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.)

  10. Applications of nuclear data on short-lived fission products

    International Nuclear Information System (INIS)

    The study of short-lived fission products gives information about the nuclear structure on the neutron-rich side of stability. The data are also of interest for various applications both to basic science and to nuclear technology. Some of these applications, taken up by the OSIRIS group at Studsvik, are described in the present contribution. (orig.)

  11. Recent progress in analysis for fission products

    International Nuclear Information System (INIS)

    A great deal of progress has been achieved in analysis of fission products during the 1980s. In situ analysis of fission products and direct assay of radiowaste packages have been developed to meet the needs of radiowaste treatment and disposal. Activation analysis and non-radiometric method have been used to measure long-lived fission product nuclides. Their sensitivity is superior to that of traditional radiochemical analysis. Some new work on the Cherenkov counting technique and rapid radiochemical analysis has been published. The progress is reviewed from the point of view of methodology

  12. Thermodynamic analysis of volatile organometallic fission products

    International Nuclear Information System (INIS)

    The ability to perform rapid separations in a post nuclear weapon detonation scenario is an important aspect of national security. In the past, separations of fission products have been performed using solvent extraction, precipitation, etc. The focus of this work is to explore the feasibility of using thermochromatography, a technique largely employed in superheavy element chemistry, to expedite the separation of fission products from fuel components. A series of fission product complexes were synthesized and the thermodynamic parameters were measured using TGA/DSC methods. Once measured, these parameters were used to predict their retention times using thermochromatography. (author)

  13. 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

  14. 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.

  15. 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)

  16. Adsorption of fission products on mediterranean mud

    International Nuclear Information System (INIS)

    Partition coefficients of some fission products have been measured in sea water on mud taken from the bottom of the Mediterranean sea. A discussion follows on the behaviour of these radioisotopes. (author)

  17. TMI-2 fission product inventory estimates (draft)

    International Nuclear Information System (INIS)

    This report presents the results of analyses performed to estimate the inventory and distribution of selected radioisotopes within the TMI-2 reactor system. The intent of the report is to document the method used in estimating the fission product inventory and associated uncertainties. The values presented should be viewed as preliminary. Selected radioisotopes for which best-estimate inventories and uncertainties are presented include: Krypton (Kr-85), Cesium (Cs-137), Iodine (I-129), Antimony (Sb-125), Ruthenium (Ru-106), Strontium (Sr-90), Cerium (Ce-144), and Europium (Eu-154). The TMI-2 inventory data will provide a basis for relating the fission product behavior during a large-scale severe accident to smaller-scale experimental data and fission product behavior modeling work. This is an important link in addressing the many technical questions that relate to core damage progression and fission product behavior during severe accidents. 11 refs., 7 figs., 15 tabs

  18. Systematics of Fission-Product Yields

    International Nuclear Information System (INIS)

    Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number ZF = 90 thru 98, mass number AF = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru ∼200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from ∼ 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron (∼ fission spectrum) induced fission reactions

  19. Modeling Fission Product Sorption in Graphite Structures

    Energy Technology Data Exchange (ETDEWEB)

    Szlufarska, Izabela [University of Wisconsin, Madison, WI (United States); Morgan, Dane [University of Wisconsin, Madison, WI (United States); Allen, Todd [University of Wisconsin, Madison, WI (United States)

    2013-04-08

    The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission

  20. 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.

  1. 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.)

  2. JNDC nuclear data library of fission products

    International Nuclear Information System (INIS)

    The JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library for 1172 fission products is described in this report. The gross theory of beta decay has been used extensively for estimating unknown decay data and also some of known decay data with poor accuracy. The calculated decay powers of fission products using the present library show excellent agreement with the latest measurements at ORNL (Oak Ridge National Laboratory), LANL (Los Alamos National Laboratory) and UTT (University of Tokyo, Tokai) for cooling times shorter than 103 s after irradiation. The calculated decay powers by the existing libraries showed systematic deviations at short cooling times; the calculated beta and gamma decay powers after burst fission were smaller than the experimental results for cooling times shorter than 10 s, and in the cooling time range 10 to 103 s the beta-decay power was larger than the measured values and the gamma decay power smaller than the measured results. The present JNDC FP nuclear data library resolved these discrepancies in the short cooling time ranges. The decay power of fission products has been calculated for ten fission types and the results have been fitted by an analytical function with 31 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)

  3. 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

  4. Basic physics of the fission process. Chapter 2

    International Nuclear Information System (INIS)

    A general description of the fission process is given with special emphasis on those aspects which are necessary for the understanding of the measurements and calculations of neutron-induced fission cross-sections. Having considered the various phases of the process, some typical properties of the low-energy fission of actinide nuclei are presented and the more specific features of neutron induced fission are examined. (U.K.)

  5. Correlation of recent fission product release data

    International Nuclear Information System (INIS)

    For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab

  6. A device for trapping fission products

    International Nuclear Information System (INIS)

    Description is given of a device for trapping the solid fission products carried by the coolant of a high temperature nuclear reactor, driven through the core, then through the reactor reflector through channels. This device is characterized in that it comprises stacks of balls or cylinders of an adsorbent substances, mounted in housings provided in the reflector. This device can adsorb 99% of the fission products carried by the coolant, without running the risk of re-cycling these products should be a depressurization occur

  7. 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

  8. Reactor with very low fission product inventory

    International Nuclear Information System (INIS)

    A fast converter with one zone and an internal breeding ratio of 1.00, with liquid fuel in the form of molten plutonium- uranium- and sodium chloride, with a thermal power of 3 GW (th) allows continuous extraction of the volatile fission products (Br, I, Kr, Xe, Te) by means of helium purging in the core. The non-volatile fission products e.g. Sr and Cs can continuously be extracted in a chemical reprocessing plant at the reactor site. The impact on an accidental release of fission products is rather significant; the amounts released are 50-100 times smaller than those in a reference reactor (LWR with oxide fuel). Because the heat sink is relatively large and after heat reduced, the temperature of the fuel does not exceed 5000C after an accident, which greatly reduces the consequences of an accident. (Auth.)

  9. Chemistry of fission products for accident analysis

    International Nuclear Information System (INIS)

    Current knowledge concerning the chemical state of the fission product elements during the development of accidents in water reactor systems is reviewed in this paper. The fission product elements which have been considered are Cs, I, Te, Sr and Ba but aspects of the behavior of Mo, Ru and the lanthanides are also discussed. Some features of the reactions of the various species of these elements with other components of the reactor systems are described. The importance of having an adequate knowledge of thermodynamic data and phase equilibria of relatively simple systems in order to interpret experimental observations on complex multi-component systems is stressed

  10. Ex-vessel fission product release modeling

    International Nuclear Information System (INIS)

    Release of fission products from core debris after reactor vessel failure is of interest in current severe accident source term research. This paper focuses on significant physical phenomena, requirements, and feedbacks in the context of integrated accident analysis for modeling these releases after initial corium distribution outside the vessel. There are many assumptions made in integrated accident analyses to which ex-vessel fission product release is sensitive. One assumption internal to the release model, the allowed species list, has been demonstrated to significantly affect release of strontium and lanthanum

  11. 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.)

  12. Fission product release mechanisms and groupings

    International Nuclear Information System (INIS)

    During CANDU postulated accidents the reactor fuel is estimated to be exposed to a variety of conditions. These conditions are dynamic and, during the course of an accident, the fuel may experience a wide range of temperatures and conditions from highly oxidizing to mildly reducing environments. The exposure of the reactor fuel to these environments and temperatures may affect its stoichiometry and release performance. In this paper a review of the important fission product release mechanisms is presented, the results of three out-of-pile experimental programs are summarized, and fission product release groups, for both oxidizing and reducing conditions are proposed. (author)

  13. 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

  14. 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.)

  15. 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.

  16. Fission product yields from 22 MeV neutron-induced fission of 235U

    International Nuclear Information System (INIS)

    The chain yields of 28 product nuclides were determined for the fission of 235U induced by 22 MeV neutrons for the first time. Absolute fission rate was monitored with a double-fission chamber. Fission product activities were measured by HPGe γ-ray spectrometry. Time of flight technique was used to measure the neutron spectrum in order to estimate fission events induced by break-up neutrons and scattering neutrons. A mass distribution curve was obtained and the dependence of fission yield on neutron energy is discussed

  17. 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

  18. 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.)

  19. 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.)

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Fission product yields from 19.1 MeV neutron induced fission of 238U

    International Nuclear Information System (INIS)

    36 chain yields were determined for the fission of 238U induced by 19.1 MeV neutrons for the first time. Absolute fission rate was monitored with a double-fission chamber. Fission product activities were measured by HPGe γ-ray spectrometry. Threshold detector method was used to measure the neutron spectrum in order to estimate the fission events induced by break-up neutrons and scattering neutrons. A mass distribution curve was obtained and the dependence of fission yield on neutron energy was discussed

  5. Delayed Neutrons and Photoneutrons from Fission Products

    International Nuclear Information System (INIS)

    Delayed neutrons: Most studies of the delayed neutrons from fission have involved analysis of the kinetic behaviour of fusion chain- reacting systems, analysis of the gross neutron decay (resolved into six groups with approximate half-lives of 0.2, 0.5, 2, 6, 22 and 55 s) and some measurements of the neutron spectra (the energies extendfrom 0.1 to 1.2 MeV, peaking in the range 0.2 to 0.5 MeV). Rapid separations of fission-produced halogens have indicated seven isotopes (Br87,88,89,90 and I137,138,139). and rare gas analysis has indicated 1.5-s Kr and 6-s Rb as definite delayed neutron precursors. These identified precursors account for some 80% of the total delayed neutron yields. Theoretical predictions of possible precursors point to a few tens of such nuclides to be found mainly in regions just above closed neutron shells. Total neutron yields are observed to increase with mass number and decrease with atomic number of the fissioning nuclide. Yields are nearly independent of the energy of the incident fissioning neutron at energies up to several MeV. In this range observed group yields,-especially of the long-lived precursors, ate in fairly good agreement with fission mass and charge distributions, and calculated neutron emission probabilities. . Further detailed studies of delayed neutron precursors (particularly in the difficult short half-life region) require development of ultra-fast radiochemical separation procedures (or on-line isotope separation) and fast neutron spectroscopy of high resolution and efficiency. Photoneutrons; A knowledge of the intensities and gamma-ray spectra of fission products is of practical importance in reactor technology particularly with respect to gamma heating, shielding and radiation effects. Gamma-rays of energies greater than 2.23 and 1.67 MeV cause emission of photoneutrons from deuterium and beryllium respectively, and are important in the kinetics of heavy water and beryllium-moderated reactors. The rate of photoneutron

  6. Fission Product Inventory in CANDU Fuel

    International Nuclear Information System (INIS)

    When the reactor is operated at power, fuel composition changes continuously. The fission reaction produces a large variety of fission fragments which are radioactive and decay into other isotopic species. For different accident analyses or operational events, detailed calculations of the fuel radioactive inventory (fission products and actinides) are needed. The present paper reviews two types of radioactive inventory calculations performed at Cernavoda NPP: one for determining the whole core inventory and one for determining the evolution of the inventory within fuel bundles stored in the Spent Fuel Bay. Two computer codes are currently used for radioactive inventory calculations: ORIGEN-S and ELESTRES-IST. The whole core inventory calculation was performed with both codes, the comparison showing that ELESTRES-IST gives a more conservative result. One of the challenges met during the analysis was to set a credible, yet conservative “image” of the in core fuel power/burnup distribution. Consequently, a statistical analysis was performed to find the best estimate plus uncertainties map for the power/burnup distribution of all in core fuel elements. For each power/burnup in the map, the fission product inventory was computed using a scaled irradiation history based on the Limiting Overpower Envelope. After the Fukushima accident, the problem of assessing the consequences of a loss of cooling event at the Spent Fuel Bay was raised. In order to estimate its impact, a calculation for determining the fission products inventory and decay heat evolution within the spent fuel bundles stored in the bay was performed. The calculation was done for a bay filled with fuel bundles up to its maximum capacity. The results obtained have provided a conservative estimation of the decay heat released and the expected evolution of the water temperature in the bay. This provided a technical basis for selecting the emergency actions required to cope with such events. (author)

  7. 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

  8. Behaviour and retention of fission products in the containment

    International Nuclear Information System (INIS)

    Fission product retention in the containment is a key issue in estimating radiological source terms from accidents in LWRs. Since the publication of the Reactor Safety Study considerable progress has been made in many countries in the development of models and codes to describe fission product behaviour. The codes have been compared. Experimental investigations, both basic research and large-scale integral simulations, were done and are being continued. The current status is that codes are believed to be almost in their final stage for application to source term assessment. The questions that remain concern coupling to thermal-hydraulics and the influence of spatial inhomogeneities in complicated containment geometries. Such influences, although believed to be of lesser importance, remain a possibility to alter source terms; their magnitude has still to be quantified. The results of calculations with the new codes show a reduction of the source term by orders of magnitude compared with those from the early risk studies. (author)

  9. 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

  10. Basic semantics of product sounds

    OpenAIRE

    Özcan Vieira, E.; Egmond , R. van

    2012-01-01

    Product experience is a result of sensory and semantic experiences with product properties. In this paper, we focus on the semantic attributes of product sounds and explore the basic components for product sound related semantics using a semantic differential paradigmand factor analysis. With two experiments, we determined eight factors that underlie the semantic associations of product sounds (attention, roughness, smoothness, temporal constancy, (un)familiarity, unpleasantness, machinery, a...

  11. Two-lump fission product model for fast reactor analysis

    International Nuclear Information System (INIS)

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

  12. Behaviour of Fission Products in Liquid Sodium

    International Nuclear Information System (INIS)

    Out-of-pile experiments were performed to study the behaviour of fission products released in sodium during the melting of a specimen of irradiated uranium. With the experimental rig employed it was possible to heat 250 litres of sodium to a temperature of 550°C and to melt a fuel sample in it containing about 200 mCi of fresh fission products. Samples were taken from the crucible, the sodium and the cover argon to determine the various diffusion coefficients for the fuel in the sodium and the sodium in the argon. The behaviour and the efficiency of various filters were measured with an argon sampling circuit fitted with a coarse filter for the sodium vapour, a magnetic filter, various iodine traps and a rare-gas trap. With this experimental rig it is possible to determine the wall contamination rate and to check the efficiency of various decontamination methods. (author)

  13. Preparative electrophoresis of industrial fission product solutions

    International Nuclear Information System (INIS)

    The aim of this work is to contribute to the development of the continuous electrophoresis technique while studying its application in the preparative electrophoresis of industrial fission product solutions. The apparatus described is original. It was built for the purposes of the investigation and proved very reliable in operation. The experimental conditions necessary to maintain and supervise the apparatus in a state of equilibrium are examined in detail; their stability is an important factor, indispensable to the correct performance of an experiment. By subjecting an industrial solution of fission products to preparative electrophoresis it is possible, according to the experimental conditions, to prepare carrier-free radioelements of radiochemical purity (from 5 to 7 radioelements): 137Cs, 90Sr, 141+144Ce, 91Y, 95Nb, 95Zr, 103+106Ru. (author)

  14. HAMCIND, Cell Burnup with Fission Products Poisoning

    International Nuclear Information System (INIS)

    1 - Description of program or function: HAMCIND is a cell burnup code based in a coupling between HAMMER-TECHNION and CINDER. The fission product poisoning is taken into account in an explicit fashion. 2 - Method of solution: The nonlinear coupled set of equations for the neutron transport and nuclide transmutation equations and nuclide transmutation equations in a unit cell is solved by HAMCIND in a quasi-static approach. The spectral transport equation is solved by HAMMER-TECHNION at the beginning of each time-step while the nuclide transmutation equations are solved by CINDER for every time-step. The HAMMER-TECHNION spectral calculations are performed taking into account the fission product contribution to the macroscopic cross sections (fast and thermal), in the inelastic scattering matrix and even in the thermal scattering matrices. 3 - Restrictions on the complexity of the problem: Restrictions and/or limitations for HAMCIND depend upon the local operating system

  15. 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.)

  16. 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.)

  17. Model Calculation of Fission Product Yields Data using GEF Code

    International Nuclear Information System (INIS)

    Fission yields data are classified with spontaneous fission data and neutron induced fission data. The fission product yields data at several energy points for the limited actinides are included in nuclear data libraries such as ENDF/B, JEFF and JENDL because production of those is based mainly on experimental results and it is very difficult to conduct experiments for all actinides and continuous energies. Therefore, in order to obtain fission yields data without experimental data, a theoretical fission model should be introduced to produce the yields data. GEneral Fission model (GEF) is developed to predict the properties for fissioning systems that have not been measured and that are not accessible to experiment. In this study, the fission yields data generated from GEF code are compared with the measured data and the recently available nuclear data libraries. The GEF code is very powerful tool to generate fission yields without measurements. Also, it can produce the distribution of fission product yields for continuous neutron energy while measured data are given only at several energies. The fission yields data of 235U have been tentatively generated with GEF code in this work. Comparing GEF results with measurements and recently released evaluated fission yields data, it is confirmed that GEF code can successfully predict the fission yields data. With its sophisticated model, GEF code is playing a significant role in nuclear industry

  18. Fission product measurement methods. Present state of knowledge

    International Nuclear Information System (INIS)

    Latest state of development of nuclear charge and mass distributions in fission products is presented. A global view (still incomplete) is given using distribution variations in function of number of mass, atomic number and excitation energy of the fissioning nucleus

  19. Fission product release mechanisms and pathways

    International Nuclear Information System (INIS)

    It is axiomatic that the severity of a nuclear reactor accident is determined by the extent of radioactivity escape which results. The main focus of site safety analyses is thus on fission product release and transport. Of all the processes involved, fission product escape from the fuel-cladding region into the primary coolant circuit is perhaps the most simple to describe; even so, it is an extremely complex function of the time/temperature history of the fuel-cladding system during an accident, since many mechanisms for release are involved. Depending upon the particular fission product species, these release mechanisms range from simple gaseous expansion processes at low temperatures to evaporation-condensation processes (aerosol formation) over molten fuel. Because of these complexities, it is convenient to subdivide the time/temperature sequence of an accident into more or less discrete phases over which specific release mechanisms dominate. Four such phases are the periods of (1) gap release, (2) meltdown release, (3) vaporization, and (4) oxidation release. This approach simplifies the problem considerably, although some loss of uniformity results. The methodology applies to BWR and PWR reactors with appropriate adaptations

  20. 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

  1. Progress in fission product nuclear data. No. 13

    International Nuclear Information System (INIS)

    This is the 13th issue of a report series published by the Nuclear Data Section of the IAEA. The types of activities included 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 bumped fission product data (decay heat, absorption, etc.). The first part of the report consists of unaltered original data 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. Part 3 contains requirements for further measurements

  2. 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

  3. 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

  4. Energy Dependence of Plutonium Fission-Product Yields

    International Nuclear Information System (INIS)

    A method is developed for interpolating between and/or extrapolating from two pre-neutron-emission first-chance mass-asymmetric fission-product yield curves. Measured 240Pu spontaneous fission and thermal-neutron-induced fission of 239Pu fission-product yields (FPY) are extrapolated to give predictions for the energy dependence of the n + 239Pu FPY for incident neutron energies from 0 to 16 MeV. After the inclusion of corrections associated with mass-symmetric fission, prompt-neutron emission, and multi-chance fission, model calculated FPY are compared to data and the ENDF/B-VII.1 evaluation. The ability of the model to reproduce the energy dependence of the ENDF/B-VII.1 evaluation suggests that plutonium fission mass distributions are not locked in near the fission barrier region, but are instead determined by the temperature and nuclear potential-energy surface at larger deformation.

  5. Energy Dependence of Plutonium Fission-Product Yields

    Science.gov (United States)

    Lestone, J. P.

    2011-12-01

    A method is developed for interpolating between and/or extrapolating from two pre-neutron-emission first-chance mass-asymmetric fission-product yield curves. Measured 240Pu spontaneous fission and thermal-neutron-induced fission of 239Pu fission-product yields (FPY) are extrapolated to give predictions for the energy dependence of the n + 239Pu FPY for incident neutron energies from 0 to 16 MeV. After the inclusion of corrections associated with mass-symmetric fission, prompt-neutron emission, and multi-chance fission, model calculated FPY are compared to data and the ENDF/B-VII.1 evaluation. The ability of the model to reproduce the energy dependence of the ENDF/B-VII.1 evaluation suggests that plutonium fission mass distributions are not locked in near the fission barrier region, but are instead determined by the temperature and nuclear potential-energy surface at larger deformation.

  6. 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.

  7. Recent progress in fission product separation

    International Nuclear Information System (INIS)

    Successful experiments have been done on the method described at Geneva in 1958. The process has been considerably improved: 1 - Initially, the caesium phospho tungstate precipitate was leached barium hydroxide in the centrifuge and this was followed by a distillation of ammonia in a concentrator. The barium hydroxide was then eliminated by carbonate precipitation and centrifugation. It has been proved that the ammonia distillation could be replaced by its evaporation during centrifugation, thus eliminating the need of a concentrator. It was then possible to carry out the carbonation on the solide-liquid mixture produced by the baryte water leaching. 2 - In applying the above process to the treatment of solutions derived from uranium molybdenum fuels, concentrating is to be recommended in order to hold the molybdenum in solution by complexing it with phosphoric acid. This complexing process provides a suspension of zirconium phosphate and ammonium phospho tungstate. These are separated by passing into a basic medium which precipitates the zirconium oxide, then turning back to an acid medium; the end of the treatment remains unchanged. 3 - Studies carried out in several countries on the exchange properties of hetero-polyacid salts have always met with difficulties as a result of the poor mechanical properties of these substances. This difficulty has been overcome by wrapping the ammonium phospho tungstate in a zirconium phosphate matrix. The exchanger obtained possesses: satisfactory mechanical properties, - a capacity of 0.1 milli equivalent per gram in concentrated nitric acid solution. It can be eluted and regenerated by a solution of an ammonium salt. The procedure for recovery of these various fission products is briefly the following: extraction of rare earths by di-2-ethyl hexyl phosphoric acid into dodecane at pH 2, the chemical impurities being complexed by citric acid, extraction of most of the magnesium at pH 4 by the same solvents the solvent being

  8. 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.)

  9. The Technology and Applications of Large Fission Product Beta Sources

    International Nuclear Information System (INIS)

    Beta emitters have not received consideration as large sources of radiation power because in the past, the radiation processes of interest have been based on particles with high penetration power; hence the great emphasis on gammas and artificially accelerated electrons. About four years ago, it became apparent that a broad field of potential applications involving surface radiation treatment was developing, e. g. surface modification of formed plastics by graft copolymerization and surface pasteurization of food. For these applications, penetration in depth is wasteful and potentially harmful. Also there are two other areas for which machine electrons were not well suited: radiation-induced chemical syntheses in pressure vessels, and certain types of free radical chain reactions for which the production rate per kilowatt decreases with the square root of the dose rate. Broad area beta sources showed obvious potential advantages in all these categories and, since they are available in good yield from the fission process, merited a careful re-appraisal. On the basics of these considerations an AEC sponsored study of the applications and technology of fission product beta sources was performed. The results indicate the following: 1. There are promising areas for commercial application of fission product beta emitters in the radiation processing field, particularly in the graft copolymerization modification of formed plastic surfaces and textiles. 2. Massive, rugged, inert, safe, inexpensive beta sources may be fabricated by suitable extensions of existing techniques. Source-bearing glass formulations show particular promise. 3. Beta absorption calculations indicate that extended sources can be designed with power utilization efficiencies as high as 20 per cent. Equations and curves describing dosage and beta utilization efficiency as a function of the geometry and composition of various source-target systems were developed. An experimental program is in progress to

  10. HTR fuel: prediction of fission product release in accidents

    International Nuclear Information System (INIS)

    The basic fuel unit of the HTR is the coated particle of about 1 mm diameter. An oxidic fuel kernel is surrounded by a low density buffer layer and a silicon carbide coating sandwiched between high density pyrocarbon coatings. The total release of fission products during accidents is determined not only by the transient-induced and the irradiation-induced failure of the coatings, but also by the levels of manufacturing defects and the level of heavy metal contamination in the fuel matrix material. Modern coated fuel particles are designed so that the fission gas pressure-induced stress in the SiC coating remains small relative to the strength of the SiC even under full design burnup conditions. Therefore the pressure vessel failure of the particles is insignificant both in normal operations and in accidents. Silicon carbide thermal decomposition becomes the dominant failure mode as temperatures exceed 2000 deg. C. Interaction of fission products with silicon carbide leading to SiC corrosion is the dominant failure mechanism below 2000 deg. C. Laboratory simulations of HTR transients have usually measured the release of Cs 137 and Kr 85 as indicators of the coating failure. Once the silicon carbide fails by corrosion or decomposition, Cs 137 is released and is taken as the direct indicator of SiC failure in fuel performance modeling studies. In the case of Kr, an additional delay beyond the Cs release is found due to the time required for Kr to diffuse through the remaining outer pyrocarbon coating. The delay between the SiC failure and gas release is analyzed to yield data on the diffusion coefficient of Kr in pyrocarbon. The present data suggest that, in terms of expected values, the fission product release during a modular reactor system transient to 1600 deg. C is dominated by the manufacturing defects and heavy metal contamination rather than irradiation-induced or transient-induced coating failure. (author)

  11. The role of fission products in whole core accidents

    International Nuclear Information System (INIS)

    The review of the role of fission products in whole-core accidents falls into two parts. Firstly, there is a discussion of the hypothetical accidents usually considered in the UK and how they are dealt with. Secondly, there is a discussion of individual topics where fission products are known to be important or might be so. There is a brief discussion of the UK work on the establishment of an equation of state for unirradiated fuel and how this might be extended to incorporate fission product effects. The main issue is the contribution of fission products to the effective vapour pressure and the experimental programme on the pulsed reactor VIPER investigates this. Fission products may influence the probability of occurrence and the severity of MFCIs. Finally, the fission product effects in the pre-disassembly, disassembly and recriticality stages of an accident are discussed. (author)

  12. Energy production using fission fragment rockets

    Science.gov (United States)

    Chapline, G.; Matsuda, Y.

    1991-08-01

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: approximately twice the efficiency if the fission fragment energy can be directly converted into electricity; reduction of the buildup of a fission fragment inventory in the reactor could avoid a Chernobyl type disaster; and collection of the fission fragments outside the reactor could simplify the waste disposal problem.

  13. Energy production using fission fragment rockets

    International Nuclear Information System (INIS)

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: Approximately twice as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs

  14. 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)

  15. 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)

  16. Impact of fuel chemistry on fission product behaviour

    International Nuclear Information System (INIS)

    The report contains a series of papers presented at SCK-CEN's workshop on the impact of fuel chemistry on fission product behaviour. Contributing authors discuss different processes affecting the behaviour of fission products in different types of spent nuclear fuel. In addition, a number of papers discusses the behaviour of actinides and fission products released from spent fuel and vitrified high-level waste in geological disposal conditions

  17. Migration of fission products in UO2. Final report

    International Nuclear Information System (INIS)

    Results of an experimental and calculational effort to examine the fundamental mechanisms of fission product migration in and release from polycrystalline uranium dioxide are reported. The experiments were designed to provide diffusion parameters for the representative fission products tellurium, iodine, xenon, molybdenum and ruthenium under both reducing and oxidizing conditions. The calculational effort applied a new model of fission product release from reactor fuel that incorporates grain growth as well as grain boundary and lattice diffusion

  18. 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

  19. 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

  20. 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

  1. Extraction process of fission products from spent nuclear fuel elements

    International Nuclear Information System (INIS)

    Process for extracting fission products contained in irradiated nuclear fuel elements consisting in bringing these elements into contact with water after having treated them mechanically to remove their cladding and/or cut them up, then separate these treated elements from the aqueous solution and recuperating at least one of the fission products concerned from this by concentrating it by distillation so as to obtain a concentrate containing these fission products and then processing this concentrate in order to ensure a long term storage of these fission products

  2. 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

  3. 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)

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. ENDF/B fission product decay data

    International Nuclear Information System (INIS)

    The fission product data have been organized by A-chains in order of ascending A from A = 72 to A = 167. The heading page is followed by more detailed information on the individual members of the chain in order of increasing Z and decreasing metastable state. The detailed information for each member includes the ENDF/B-IV File 1 comments and references if available and applicable to the decay data. Following the comments is a decay scheme of the nuclide tabulating the quantities T/sub 1/2/, Q, branching ratio (BR), (E/sub γ/), (E/sub β/), and (E/sub α/). Uncertainties are given if available in the file. Independent fission yields are given, as well as thermal cross sections and resonance integrals as obtained from ENDF/B-IV. All energies listed in this publication are in keV, and all branching ratios sum to unity. If there are spectra in the decay data file, the decay scheme is followed by tables of photon, particle, and characteristic radiation. For cases in which the multipolarities could be obtained from the file the tables also contain information on x rays, conversion electrons, and Auger electrons. Associated with the photon and particle radiation tables are the appropriate average energies per decay for each type of radiation, including neutrino radiation

  7. 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

  8. Library of data for fission products

    International Nuclear Information System (INIS)

    This is the fourth version of the CEA fission products nuclear data library. The third one has been previously published in CEA-N--1526. Data for 635 nuclides ranging from mass A=71 up to A=170 are arranged in increasing order of atomic number. Data are presented in two tables: the first one gives for each nuclide, the half-life, the Q-values and branching ratios for the various decay modes, the energies and intensities of the β-, β+ and isomeric transitions and of gamma rays; the second one gives an ordered list of all gamma ray energies, with associated nuclide, half-life and intensity. Bibliographic references and, for most of the data, uncertainties are provided

  9. Core degradation and fission product release

    International Nuclear Information System (INIS)

    Experiments on core degradation and melt progression in severe LWR accidents have provided reasonable understanding of the principal processes involved in the early phase of melt progression that extends through core degradation and metallic material melting and relocation. A general but not a quantitative understanding of late phase melt progression that involves ceramic material melting and relocation has also been obtained, primarily from the TMI-2 core examination. A summary is given of the current state of knowledge on core degradation and melt progression obtained from these integral experiments and of the principal remaining significant uncertainties. A summary is also given of the principal results on in-vessel fission product release obtained from these experiments. (author). 8 refs, 5 figs, 3 tabs

  10. Fission product source term research at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    The purpose of this work is to describe some of the research being performed at ORNL in support of the effort to describe, as realistically as possible, fission product source terms for nuclear reactor accidents. In order to make this presentation manageable, only those studies directly concerned with fission product behavior, as opposed to thermal hydraulics, accident sequence progression, etc., will be discussed

  11. Immobilization of fission products in phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    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

  12. Some 235U reference fission product yield data evaluation

    International Nuclear Information System (INIS)

    To satisfy the requirement of application for reference fission product yield data, the data have been and will be continuously evaluated. Present work, in which the reference data for 20 product nuclides from 235U fission were evaluated, is a part of the whole work

  13. Solidification of residual fission-product solutions; laboratory studies

    International Nuclear Information System (INIS)

    This paper describes the results obtained, at laboratory scale, during the study of the incorporation of fission products into glasses and synthetic micas. The rate of leaching of fission products from the glass and their volatility during firing were measured. A hot cell was built to complete these results. (author)

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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)

  19. Fission-product burn-up in fast reactors

    International Nuclear Information System (INIS)

    In fast reactors where breeding is emphasized the burn-up of fission products can be of considerable importance. Statistical estimates of fission-product cross-sections are combined with recent yield data for the various fissionable species to estimate the gross fission-product cross-section as a function of irradiation time in a number of fast reactor spectra with various fuels. Because of gaps in yield data for some of the fuel species, it is necessary to interpolate on the yield curves in some cases. The chain yield for a given mass is then apportioned among the chain members through use of the equal charge displacement recipe. The cross-sections estimated for U235 fission products by previous authors are supplemented by estimates for fission products important for other fuels. A range of such spectra is considered. These spectra are characterized by the index (average (Ε-1/2)) in the spectra. The sensitivity of the gross poisoning and its burn-up with respect to spectrum variations are considered. The results are also expressed in terms of a few pseudo-fission products, so that changes in effective cross-section of fission products with irradiation can be taken into account in a simple computational fashion. (author)

  20. Treatment of solutions of fission products - Separation of caesium-137

    International Nuclear Information System (INIS)

    For the industrial recovery of caesium-137 from solutions of fission products, the authors utilized the analytical method for determination of caesium by dipicrylamine, adapting it to use on an industrial scale and to the high level of the activities encountered. The process recommended makes it possible both to isolate caesium as a chloride and to recover the precipitation reagent, in one and the same operation. A basic method is suggested. The authors studied the effect of radiation on dipicrylamine and its compounds, this effect proving to be practically nil for solid compounds and negligible for their solutions. The entrainment of caesium by ammonia ion was also studied. The advantages of the proposed process are : high decontamination of the caesium, simple operation and free recycling fo the reagent, high yield for caesium recovery and for dipicrylamine, considerable concentration of caesium activity, operation at room temperature and possibility of continuous operation. By this process caesium can be recovered before certain fission products are eliminated. (author)

  1. 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

  2. Separation of actinides and fission products from carbonate containing streams

    International Nuclear Information System (INIS)

    The capacities of the anion exchange resins AG 1-X8, AG 2-X8 and Bio-Rex 5 were determined for the carbonato complexes of UO22+, NpO22+, PuO22+, Pu4+, AmO22+ and Am3+ in batch and dynamic experiments. The Bio-Rex 5 resin, used for the first time in such experiments, shows a clear superiority over the strong basic resins which have been used in the treatment of uranium ores. The influence of the ratio U : CO32-, the pH-value, the temperature, the equilibration of the resin, the contact time and the concentration of uranium to the column parameters distribution coefficient, hold back- and break through capacities have been investigated for batch and dynamic experiments. The best results were obtained for a medium with pH 6-8 and low concentrations of actinides and carbonate ions, 0.04 M and 0.12 M respectively. In order to obtain informaiton on the behaviour of the fission products occuring in the recovery of the organic phase of the Purex-process, these expected fission products were added to the uranium solution, fixed and eluted together with the uranium and Bio-Rex 5. (orig./HK)

  3. Development of glass ceramics for the incorporation of fission products

    International Nuclear Information System (INIS)

    Spontaneous devitrification of fission-product-containing borosilicate glasses can be avoided by controlled crystallization after melting. Glass ceramics have been developed from a vitrified simulated waste and further improvement of product properties was achieved. In particular perovskite, h-celsian, diopside and eucryptite glass ceramics were prepared. These contained leach resistant host phases which exhibited considerable enrichment of long-lived fission products. All products showed increased impact resistance, but the thermal expansion was only slightly improved

  4. Fission product revaporization in the reactor cooling system

    International Nuclear Information System (INIS)

    The reactor cooling system (RCS) of an LWR can act as an efficient scrubber of volatile fission products released during a meltdown accident before vessel melt-through. This assertion is based on calculations that consider transport of the volatile fission products as vapours or condensed on particles. Retention in the primary system occurs by condensation or reaction with structural surfaces or by fallout of particles containing fission products. It is shown that this picture is perturbed by inclusion of decay heating in the thermal-hydraulic calculations. To do so we make use of the TRAP-MELT3 code which integrates the MERGE and TRAP-MELT2 codes and thus permits simultaneous calculation of thermal-hydraulics and fission product transport in the RCS during the meltdown phase of a severe LWR accident. Calculations on the Surry TMLB' sequence show that while structure temperatures can rise as much as 100 K with inclusion of decay heat, little additional fission product release from the RCS results before melt-through of the reactor vessel. After melt-through, structural temperatures are likely to continue to rise and fission products migrate along the RCS by revolatilizing in the hotter regions and condensing in the cooler regions. The potential for a significant source term of volatile fission products to the containment after melt-through thus exists. For these materials, therefore, the RCS may act more as a retardant than a retainer. Quantification of this conjecture will require further analyses. (author)

  5. 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

  6. 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

  7. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  8. 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.

  9. Determination of 140La fission product interference factor for INAA

    International Nuclear Information System (INIS)

    Instrumental Neutron Activation Analysis (INAA) is a technique widely used to determine the concentration of several elements in several kinds of matrices. However if the sample of interest has higher relative uranium concentration the obtained results can be interfered by the uranium fission products. One of these cases that is affected by interference due to U fission is the 140La, because this radioisotope used in INAA for the determination of concentration the La is also produced by the −β of 140Ba, an uranium fission product. The 140La interference factor was studied in this work and a factor to describe its time dependence was obtained

  10. 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

  11. 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)

  12. Delayed-neutron branching ratios of fission products

    International Nuclear Information System (INIS)

    Delayed-neutron branching ratios have been reviewed for 86 nuclides, including a few isomers, among the fission products. The list comprises values reported before the end of December, 1987. (authors) (33 refs.)

  13. Nuclear Qsub(β)-values for fission products

    International Nuclear Information System (INIS)

    Experimental Qsub(β)-values for fission products are presented. The sources were produced as mass separated fission products at the OSIRIS on-line isotope separator. Recently determined Qsub(β)-values for 79,81Ga, 79,81,82Ge, 89,90Br, 116,121Ag, 119,121Cd and 139I are, together with 40 earlier measured values, compared with mass formula predictions. (orig.)

  14. 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)

  15. Application of mercury cathode electrolysis to fission-product separation

    International Nuclear Information System (INIS)

    A method involving controlled potential mercury cathode electrolysis has been developed to separate fission products. It allows the radiochemical determination of Ag, Cd, Pd, Rh, Ru, Sn, Te, Sb and Mo from solutions of fission products highly concentrated in mineral salts. The general procedure consists in three main steps: electrolytic amalgam generation, destruction of amalgams and ultimate purification of elements by other means. Electrolytic operations last about five hours. Chemical yields lie between 10 per cent and 70 per cent. (authors)

  16. Atomic masses of fission product nuclei far from stability

    International Nuclear Information System (INIS)

    The techniques for measuring fission product masses far from stability are discussed and recent progress in experimental measurements is reviewed. A comparison of new mass values with predictions of 10 mass equations suggests that most theories predict far-from-stability fission product nuclei to be more bound than is found experimentally. A closer look at several isotopic chains is used to identify regions of structural change where mass equations encounter difficulty. 31 references

  17. SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

    International Nuclear Information System (INIS)

    1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

  18. 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.

  19. 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.

  20. 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

  1. Fission product release from fuel of water-cooled reactors

    International Nuclear Information System (INIS)

    The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

  2. 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

  3. Time Dependent Radio-toxicology of Fission Products

    International Nuclear Information System (INIS)

    Ionizing radiation, emitted by radiological materials, is known to cause damage to biological tissue. Prolonged exposure to radiation may cause a vast array of harmful medical effects, from enhancing future probability of cancer, up to Acute Radiation Syndrome resulting in multi-system failure. In complex radiologic release events involving fission products (nuclear fallout, reactor failures), the products' physical decay chains dictate a time dependent product inventory. As the ratios between different products vary, so does the toxicology of the radioactive inventory as a whole. The temporally varying toxicological factors should be taken into account when producing radiological risk assessments for populations. In this paper we study the time varying toxicology of fission products, using a specialized model named Koala, developed in Soreq NRC. A significant and monotonous rise in the aggregate toxicity of ingested fission products was noted. This result carries important implications for risk assessment, as it partially cancels out the fission product physical decay. A similar, albeit less pronounced rise was found for external exposure. Factoring activity and toxicity together allows computation of effective source terms for simple events involving fission products. We demonstrate one such source term, based on fallout from a nuclear explosion. This source term may be easily introduced into suitable atmospheric dispersion models

  4. 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

  5. Interpretation of fission product release from overheated fuel

    International Nuclear Information System (INIS)

    Recent laboratory data on the high temperature release of fission products from uranium dioxide can be described by intragranular diffusion. A first class of volatile fission products is characterized by interstitial transport with relatively small activation energy and diffusion entropy, the latter being determined by atomic sizes; chemically these products are unbound. A second class of non-volatile products is characterized by substitutional transport with relatively high activation energy and diffusion entropy; these products are bound to either oxygen or uranium. Releases predicted by this model for a certain temperature excursion of reactor fuel are compared with measurements taken during the severe fuel damage test and with computed source terms. It is concluded that only few volatile fission products will be released by the fuel and that most of them will be held back, even in the event of extreme accidents. (author)

  6. 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

  7. 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

  8. PEPIN, Methodology for Computing Concentrations, Activities, Gamma-Ray Spectra, and Residual Heat from Fission Products

    International Nuclear Information System (INIS)

    1 - Description of program or function: The concentrations, activities, gamma-ray spectra and residual heat from fission products can be calculated as a function of time for instantaneous fission or for one or more irradiation steps. 2 - Methods: Using the basic data in the libraries, the PEPIN code solves the differential equations satisfied by the fission product concentrations. Data Libraries: Independent Yields Library: 8 independent yields for 235U, 238U, 239U, 232U, 233U. Chain library: Precursor chain file for 635 nuclides. Gamma-Ray Energies Library: Average beta and gamma-ray energies in increasing order. 3 - Restrictions on the complexity of the problem: The number of nuclides must not exceed 650. The number of Gamma-Rays must not exceed 8500. The number of decay times must not exceed 59. The number of irradiation steps must not exceed 40

  9. Fission Product Yields from Fission Spectrum n+239Pu for ENDF/B-VII.1

    International Nuclear Information System (INIS)

    We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small - especially for 99Mo - we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (4%-relative) and 147Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on the

  10. 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

  11. 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.

  12. Kinetics of fission product release prior to fuel slumping

    International Nuclear Information System (INIS)

    This paper describes the primary physical/chemical models recently incorporated into a mechanistic code (FASTGRASS) for the estimation of fission product release from fuel, and compares predicted results with test data. The theory of noble gas behavior is discussed in relation to its effect on the release behavior of I, Cs, Te, Ba, and Sr. The behavior of these fission products in the presence of fuel liquefaction/dissolution and oxidation grain-growth phenomena is presented, as is the chemistry of Sr, Ba, I, and Cs. Comparison of code predictions with data indicates the following trends. Fission product release behavior from solid strongly depends on fuel microstructure, irradiation history, time at temperature, and internal fuel rod chemistry. Fuel liquefaction/dissolution, fracturing, and oxidation also exert a pronounced effect on release during fuel rod degradation. For very low burnup fuel appreciable fission product retention in previously liquefied fuel can occur due to the low concentration of fission products, and the limited growth of bubbles in the liquefied material. 24 refs., 13 figs., 9 tabs

  13. Integral test of fission-product cross sections

    International Nuclear Information System (INIS)

    A test of more than 50 nuclides of the fission-product file of the JEF-1 data library has been performed, using integral data measured in Dutch, French and US facilities. Some results are given for the capture cross sections of the 40 most important fission products in a fast reactor. The inelastic scattering cross sections of many even-mass nuclides are systematically too low due to neglect of direct-collective effects. In lumped fission-product cross sections the uncertainties due to the release of gaseous products have been reduced by means of a new burn-up model with parameters tuned to leakage data of irradiated PHENIX fuel pins

  14. 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

  15. 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.

  16. Characterization of wastes from fission 99 Mo production

    International Nuclear Information System (INIS)

    This work is a preliminary study on waste-streams generated in a fission 99 Mo production plant, their characterization and quantification. The study is based on a plant whose 99 Mo production process is the alkaline dissolution of U-target. The target is made of 1 g of enriched 235 U, therefore most of radionuclides present in the waste-streams are fission products. All the radionuclides inventories were estimated based on ORIGEN-2 Code. The characterization was done as a primary stage for the establishment of waste management plan, which should be subject for further study. (author)

  17. Integral measurement of fission products capture in fast breeder reactors

    International Nuclear Information System (INIS)

    For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set

  18. Study of the production of fission fragments from neutron induced fission on uranium 238

    International Nuclear Information System (INIS)

    This work is devoted to the study of the mass and charge distributions of fission fragments from the fission induced on uranium 238 by neutrons from 1 to 150 MeV. An experimental program allowed us to gather and analyse new data. The obtained results were interpreted by an original model, based on a microscopic description of the reaction. Data were taken at the LANSCE laboratory of Los Alamos (USA) where we used the neutron source WNR and the germanium array GEANIE. The aim was to measure secondary fission fragment production yields from a spectroscopic analysis of the prompt gamma and x-rays. A device with photovoltaic cells used as fission fragment detectors was developed. The trigger created with this device allowed us to reduce the background from the other neutron induced reactions. Close to one hundred fragments were identified and excitation functions were extracted for about thirty of them. Mass and charge distributions at different incident energies were extracted from these measurements. These results were then compared to evaluated reference data (Wahl systematics). It showed that the calculations are consistent with the measurements at low energies (below 20 MeV) but partially fail to reproduce the data at higher energy. To go into more detail about the obtained results, the reaction was studied using an original model. It provided a dynamic and totally microscopic description of the fission from constrained self-consistent Hartree-Fock-Bogoliubov calculations. This work was completed in two parts. First, a potential energy surface of the fissioning system was calculated in a deformation plane defined by the elongation and asymmetry variables. The second part was to use the resolution of the dynamic Schroedinger equation on this surface giving us a fragment mass distribution which we then compared to the low energy data. (author)

  19. 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)

  20. 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

  1. Fission product removal from molten salt using zeolite

    International Nuclear Information System (INIS)

    Spent nuclear fuel (SNF) can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. The fuel is dissolved in molten chloride salt. Non-transuranic fission products in the molten salt are ion-exchanged into zeolite A, which is subsequently mixed with glass and consolidated. Zeolite was found to be effective in removing fission product cations from the molten salt. Breakthrough of cesium and the alkaline earths occurred more rapidly than was observed for the rare earths. The effluent composition as a function of time is presented, as well as results for the distribution of fission products along the length of the column. Effects of temperature and salt flow rate are also discussed

  2. 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.

  3. The chemistry of fission products for accident analysis

    International Nuclear Information System (INIS)

    Current knowledge concerning the chemical state of the fission product elements during the development of accidents in water reactor systems is reviewed in this paper. The fission products elements which have been considered are Cs, I, Te, Sr and Ba but aspects of the behaviour of Mo, Ru and the lanthanides are also discussed. Some features of the reactions of the various species of these elements with other components of the reactor systems are described. The importance of having an adequate knowledge of thermodynamic data and phase equilibria of relatively simple systems in order to interpret experimental observations on complex multi-component systems is stressed

  4. 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)

  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. Gaseous fission products monitoring in an irradiation circuit

    International Nuclear Information System (INIS)

    A study on the behaviour of the gaseous fission products as a linear source of radiation. The study is intended to support the specification of an appropriate radiation detection device for the gas and/or pressurized water systems of an irradiation circuit. Based upon the generation of the gaseous fission products, the more appropriate isotopes were selected, regarding aspects like concentrations, half-life and gamma emission. The isotopes to be monitored were then chosen to be: 135 Xe, 85 K and 131 I. (author)

  7. Fission product release from irradiated LWR fuel under accident conditions

    International Nuclear Information System (INIS)

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 20000C are presented in this paper

  8. Cerenkov Detectors for Fission Product Monitoring in Reactor Coolant Water

    International Nuclear Information System (INIS)

    The expected properties of Cerenkov detectors when used for fission product monitoring in water cooled reactors and test loops are discussed from the point of view of the knowledge of the sensitivity of these detectors to some beta emitting isotopes. The basic theory for calculation of the detector response is presented, taking the optical transmission in the sample container and the properties of the photomultiplier tube into account. Special attention is paid to the energy resolution of this type of Cerenkov detector. For the design of practical detectors the results from several investigations of various window and reflector materials are given, and the selection of photomultiplier tubes is briefly discussed. In the case of optical reflectors and photomultiplier tubes reference is made to two previous reports by the author. The influence of the size and geometry of the sample container on the energy resolution follows from a separate investigation, as well as the relative merits of sample containers with transparent inner walls. Provided that the energy resolution of the Cerenkov detector is sufficiently high, there are several reasons for using this detector type for failed-fuel-element detection. It seems possible to attain the desired energy resolution by careful detector design

  9. Experiments With Mass-Spectroscopically Separated Fission Product Ions

    International Nuclear Information System (INIS)

    After a short description of the fission product mass spectrograph installed at the reactor at Garching (Munich) a survey is given of the experiments done with it. Some of these concern the interaction of the swift separated fission product ions with matter. Exact distributions of ion charge numbers were obtained for particles having almost their initial kinetic energies and for those slowed down more or less in thin foils. Their multiple scattering in such foils was also measured. It is hoped that in addition their single scattering on gas atoms can be examined. The energy losses in foils will be studied exactly. The pulse-height defects observed in counting fission particles with surface barrier counters were measured extensively. Other experiments are concerned with features of the fission process and decays of the fission products. In nuclear emulsions the number of β-particle tracks emerging from the ends of the tracks of fission particles of definite mass numbers were counted. From the distributions of these numbers conclusions can be drawn on the distributions of the initial nuclear charges in the decay chains. In addition to this the β- decays of the particles of definite mass numbers are examined directly with a 4π counter. The distributions of the initial kinetic energies of the particles of different mass numbers are also studied. The possibility is discussed of measuring in coincidence with the help of two such spectrographs, both particles of the fission processes travelling in opposite directions. It might also be of advantage to use a mass spectrograph in connection with a large accelerator to study the products of spallation and fragmentation processes. For experiments on the interaction of fission particles with matter the intensity of our spectrograph is sufficient in most cases. In studies of the fission process and the β decays often a-much higher rate of separated particles would be desired. Therefore plans have been made for a new

  10. 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)

  11. Progress in Establishment of Fission Mo Production Technology in Korea

    International Nuclear Information System (INIS)

    Research activities have been made in both the development of the fission Mo production process and the designing of the production facility that will be established at Kijang, Korea including a new research reactor in 2017. Progress in the process development for target preparation, target dissolution, Mo extraction, and purification has been made. It is also a great concern to minimize the radioactive wastes or at least to generate the wastes in readily treatable forms in the project. After series of cold experiments, the target dissolution and solution formulation for a column operation are optimized. Progress in the design of the production facility has been made. Two trains of hot cells including the waste storages have been proposed for the alternative operation of the facility. A radioisotope production facility is designed to locate next to the fission Mo production building to provide a simpler and easier handling pathway of the products

  12. ENDF/B-6 fission-product yield sublibraries

    International Nuclear Information System (INIS)

    The contents and the documentation of the ENDF/B-6 fission-product yield sublibraries which were released in 1991 and updated in 1993, are summarized. Copies of the data libraries are available on magnetic tape of PC diskettes from the IAEA Nuclear Data Section, costfree upon request. (author). 1 tab

  13. 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)

  14. Products of fission, fusion and deep inelastic reactions

    International Nuclear Information System (INIS)

    Factors which influence the yields in heavy ion reactions such as fusion, fission fragment production and deep inelastic reactions are considered in the context of the design of spectroscopic experiments. Factors examined include the suitability of a reaction for a particular application, the expected yield of the required nucleus, and parameters responsible for uncertainties in predicted yields. (U.K.)

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

    International Nuclear Information System (INIS)

    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 ∼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

  16. 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)

  17. Fission products analysis. Strontium 89 and strontium 90 radiometric determination

    International Nuclear Information System (INIS)

    Determination of strontium 89 et 90 in nitric solutions of fission products, suitable for strontium content giving a nuclear activity of at least 10-5 microcurie/ml. Calcium, barium, yttrium and rare earths are eliminated before beta counting with and without threshold

  18. 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

  19. Mo-99 production by fission and future projections

    International Nuclear Information System (INIS)

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

  20. 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

  1. 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)

  2. Simultaneous estimation of Pu and fission products by gamma spectrometry

    International Nuclear Information System (INIS)

    A gamma spectrometric method is described for the simultaneous estimation of Pu and fission products using a 62cc intrinsic germanium detector coupled to a 4K MCA. The 120 KeV peak of 239Pu was employed for the assay of plutoniu m. The low energy 51 KeV photopeak of 239Pu was not employed due to the interfer ence of the germanium escape peak from 241Am gamma. A nonlinear exponential parame terised function was employed to relate the concentration of 239Pu and the counts obtained from 129 KeV photopeak after subtracting the compton and background. Standard solutions were used for computing the fitting parameters. For fission product analysis, an efficiency versus energy plot was generated using a fission product solution of known individual activities. This was then fitted in a quadratic equation and the fitting parameters were obtained. The inbuilt programme in 4K MCA was used to calculate the 239Pu and the fission products concentration from the respective fitting factors. The isotopic composition was fed externally to obtain the total plutonium concentration. This method was used for the analysis of Pu in the range of 0.5 to 5 g/1. The values when compared with those obtained with coulometric meth od show an agreement within ± 2.5 per cent in the above range. (author). 4 refs., 4 figs., 2 tables

  3. Calculation of vapor pressure of fission product fluorides and oxyfluorides

    International Nuclear Information System (INIS)

    The equilibrium diagrams of the condensed phases - solid and liquid - and vapor phase are collected for the principal fluorides and oxyfluorides of fission product elements (atomic number from 30 to 66). These diagrams are used more particularly in fuel reprocessing by fluoride volatility process. Calculations and curves (vapor pressure in function of temperature) are processed using a computer program given in this report

  4. 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

  5. Reducing uncertainties for short lived cumulative fission product yields

    International Nuclear Information System (INIS)

    Uncertainties associated with short lived (halflives less than 1 day) fission product yields listed in databases such as the National Nuclear Data Center's ENDF/B-VII are large enough for certain isotopes to provide an opportunity for new precision measurements to offer significant uncertainty reductions. A series of experiments has begun where small samples of 235U are irradiated with a pulsed, fission neutron spectrum at the Nevada National Security Site and placed between two broad-energy germanium detectors. The amount of various isotopes present immediately following the irradiation can be determined given the total counts and the calibrated properties of the detector system. The uncertainty on the fission yields for multiple isotopes has been reduced by nearly an order of magnitude. (author)

  6. Fission Product Yields from Fission Spectrum n+ 239Pu for ENDF/B-VII.1

    Science.gov (United States)

    Chadwick, M. B.; Kawano, T.; Barr, D. W.; Mac Innes, M. R.; Kahler, A. C.; Graves, T.; Selby, H.; Burns, C. J.; Inkret, W. C.; Keksis, A. L.; Lestone, J. P.; Sierk, A. J.; Talou, P.

    2010-12-01

    We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small — especially for 99Mo — we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (4%-relative) and 147Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on

  7. Evaluation of fission product yields from fission spectrum n+239Pu using a meta analysis of benchmark data

    Science.gov (United States)

    Chadwick, Mark B.

    2009-10-01

    Los Alamos conducted a dual fission-chamber experiment in the 1970s in the Bigten critical assembly to determine fission product data in a fast (fission neutron spectrum) environment, and this defined the Laboratory's fission basis today. We describe how the data from this experiment are consistent with other benchmark fission product yield measurements for 95,97Zr, 140Ba, 143,144Ce, 137Cs from the NIST-led ILRR fission chamber experiments, and from Maeck's mass-spectrometry data. We perform a new evaluation of the fission product yields that is planned for ENDF/B-VII.1. Because the measurement database for some of the FPs is small—especially for 147Nd and 99Mo—we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data. The %-relative changes compared to ENDF/B-VI are small for some FPs (less than 1% for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (3%) and 147Nd (5%). We suggest an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average energies.

  8. 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)

  9. Simulation of fission products behavior in severe accidents for advanced passive PWR

    International Nuclear Information System (INIS)

    Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

  10. 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

  11. Measurement of fission product gases in the atmosphere

    Science.gov (United States)

    Schell, W. R.; Tobin, M. J.; Marsan, D. J.; Schell, C. W.; Vives-Batlle, J.; Yoon, S. R.

    1997-01-01

    The ability to quickly detect and assess the magnitude of releases of fission-produced radioactive material is of significant importance for ongoing operations of any conventional nuclear power plant or other activities with a potential for fission product release. In most instances, the control limits for the release of airborne radioactivity are low enough to preclude direct air sampling as a means of detection, especially for fission gases that decay by beta or electron emission. It is, therefore, customary to concentrate the major gaseous fission products (krypton, xenon and iodine) by cryogenic adsorption for subsequent separation and measurement. This study summarizes our initial efforts to develop an automated portable system for on-line separation and concentration with the potential for measuring environmental levels of radioactive gases, including 85Kr, 131,133,135Xe, 14C, 3H, 35S, 125,131I, etc., without using cryogenic fluids. Bench top and prototype models were constructed using the principle of heatless fractionation of the gases in a pressure swing system. This method removes the requirement for cryogenic fluids to concentrate gases and, with suitable electron and gamma ray detectors, provides for remote use under automatic computer control. Early results using 133Xe tracer show that kinetic chromatography, i.e., high pressure adsorption of xenon and low pressure desorption of air, using specific types of molecular sieves, permits the separation and quantification of xenon isotopes from large volume air samples. We are now developing the ability to measure the presence and amounts of fission-produced xenon isotopes that decay by internal conversion electrons and beta radiation with short half-lives, namely 131mXe, 11.8 d, 133mXe, 2.2 d, 133Xe, 5.2 d and 135Xe, 9.1 h. The ratio of the isotopic concentrations measured can be used to determine unequivocally the amount of fission gas and time of release of an air parcel many kilometers downwind from a

  12. 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

  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. 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)

  15. Rapid quantitation of uranium from mixed fission product samples

    International Nuclear Information System (INIS)

    Chemical similarities between uranium(VI) and molybdenum(VI) create challenges for separation and quantitation of uranium isotopes from a mixed fission product sample. The purpose of this work was to demonstrate an improved chemical separation for the detection of 235U using gamma spectroscopy. The optimized method, which included extraction and anion exchange chromatography, demonstrated a consistent chemical yield of 74 ± 3 % for uranium. Using a fresh fission product sample the minimum detectable activity for 235U, 237U, and 238U was reduced by a factor of two. The chemical isolation of uranium was achieved in less than 4 h, with a separation factor of 1.41 9 x 105 from molybdenum. (author)

  16. 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.

  17. Fission-product release from irradiated LWR fuel

    International Nuclear Information System (INIS)

    An experimental investigation of fission product release from commercial LWR fuel under accident conditions is being conducted at Oak Ridge National Laboratory (ORNL). This work, which is sponsored by the US Nuclear Regulatory Commission (NRC), is an extension of earlier experiments up to 16000C and is designed to obtain the experimental data needed to reliably assess the consequences of accidents for fuel temperatures up to melting. The objectives of this program are (1) to determine fission product release rates from fully-irradiated commercial LWR fuel in high-temperature steam; (2) to collect and characterize the aerosol released; (3) to identify the chemical forms of the released material; (4) to correlate the results with related experimental data and develop a consistent source term model; and (5) to aid in the interpretation of tests using simulated LWR fuel

  18. Pipe contamination by fission products in PWRs. Profip code

    International Nuclear Information System (INIS)

    The estimate of fission products activities in a PWR primary circuit, in steady state and accidental conditions is necessary for protection and safety analysis. In the other hand the knowledge of these activities allow, if the release mechanisms are well described, to determine clad failures characteristics and to localize the failures in the reactor core. For this purpose, the computer code PROFIP has been developed which predict fission product activities in a PWR primary circuit dispending of fuel failures characteristics. In this paper the description of the PROFIP code is presented as well its application to fuel clad failures characterization in the Tihange 1 reactor. A method is then described, which allows to localize a failed rod on a of the core, using ratios of cesiums activities in the primary coolant

  19. 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)

  20. Recoil release of fission products from nuclear fuel

    Science.gov (United States)

    Wise, C.

    1985-10-01

    An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO 2. The calculations presented here are one way of allowing for this, other methods are suggested.

  1. Most probable charge of fission products in 24 MeV proton induced fission of 238U

    International Nuclear Information System (INIS)

    The charge distributions of fission products in 24 MeV proton-induced fission of 238U were measured by the use of an ion-guide isotope separator on line. The most probable charge (Zp) of the charge distribution was discussed in view of the charge polarization in the fission process. It was found that Zp mainly lies on the proton-rich side in the light mass region and on the proton-deficient side in the heavy mass region compared with the postulate of the unchanged charge distribution. The charge polarization was examined with respect to production Q values. copyright 1998 The American Physical Society

  2. 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)

  3. 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.)

  4. 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)

  5. 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)

  6. Data summary report for fission product release test VI-6

    International Nuclear Information System (INIS)

    Test VI-6 was the sixth test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium. The fuel had experienced a burnup of ∼42 MWd/kg, with inert gas release during irradiation of ∼2%. The fuel specimen was heated in an induction furnace at 2300 K for 60 min, initially in hydrogen, then in a steam atmosphere. The released fission products were collected in three sequentially operated collection trains designed to facilitate sampling and analysis. The fission product inventories in the fuel were measured directly by gamma-ray spectrometry, where possible, and were calculated by ORIGEN2. Integral releases were 75% for 85Kr, 67% for 129I, 64% for 125Sb, 80% for both 134Cs and 137Cs, 14% for 154Eu, 63% for Te, 32% for Ba, 13% for Mo, and 5.8% for Sr. Of the totals released from the fuel, 43% of the Cs, 32% of the Sb, and 98% of the Eu were deposited in the outlet end of the furnace. During the heatup in hydrogen, the Zircaloy cladding melted, ran down, and reacted with some of the UO2 and fission products, especially Te and Sb. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.57 g, almost equally divided between thermal gradient tubes and filters. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL Diffusion Model

  7. Chemical factors affecting fission product transport in BWR severe accidents

    International Nuclear Information System (INIS)

    Chemical changes may significantly alter physical properties of fission product materials, and hence their state and transport rate. Thus, it is possible that an appropriate accounting of chemical change could have a large impact on transport model results. This paper will describe how the chemical reactions of Cs, I, and Te are being implemented in the transport model that is used in the Severe Accident Sequence Analysis (SASA) Program at Oak Ridge National Laboratory (ORNL)

  8. Data summary report for fission product release test VI-6

    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)

    1994-03-01

    Test VI-6 was the sixth test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium. The fuel had experienced a burnup of {approximately}42 MWd/kg, with inert gas release during irradiation of {approximately}2%. The fuel specimen was heated in an induction furnace at 2300 K for 60 min, initially in hydrogen, then in a steam atmosphere. The released fission products were collected in three sequentially operated collection trains designed to facilitate sampling and analysis. The fission product inventories in the fuel were measured directly by gamma-ray spectrometry, where possible, and were calculated by ORIGEN2. Integral releases were 75% for {sup 85}Kr, 67% for {sup 129}I, 64% for {sup 125}Sb, 80% for both {sup 134}Cs and {sup 137}Cs, 14% for {sup 154}Eu, 63% for Te, 32% for Ba, 13% for Mo, and 5.8% for Sr. Of the totals released from the fuel, 43% of the Cs, 32% of the Sb, and 98% of the Eu were deposited in the outlet end of the furnace. During the heatup in hydrogen, the Zircaloy cladding melted, ran down, and reacted with some of the UO{sub 2} and fission products, especially Te and Sb. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.57 g, almost equally divided between thermal gradient tubes and filters. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL Diffusion Model.

  9. Determination of 90Sr in uranium fission products

    International Nuclear Information System (INIS)

    A previously published radiochemical procedure for the determination of 90Sr in grass and soil has been successfully employed - with minor modifications - for the determination of this nuclide in a solution of uranium fission products. It is suitable for the determination of 90Sr in environmental materials following a nuclear accident. The procedure is based on tributylphosphate extraction of 90Y, precipitation of Y-oxalate, and counting in a proportional counter. (author) figs., tabs., 10 refs

  10. 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)

  11. 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

  12. Simulations of the fission-product stopping efficiency in IGISOL

    International Nuclear Information System (INIS)

    At the Jyvaeskylae Ion Guide Isotope Separator On-Line (IGISOL) facility, independent fission yields are measured employing the Penning-trap technique. Fission products are produced, e.g. by impinging protons on a uranium target, and are stopped in a gas-filled chamber. The products are collected by a flow of He gas and guided through a mass separator to a Penning trap, where their masses are identified. This work investigates how fission-product properties, such as mass and energy, affect the ion stopping efficiency in the gas cell. The study was performed using the Geant4 toolkit and the SRIM code. The main results show a nearly mass-independent ion stopping with regard to the wide spread of ion masses and energies, with a proper choice of uranium target thickness. Although small variations were observed, in the order of 5%, the results are within the systematic uncertainties of the simulations. To optimize the stopping efficiency while reducing the systematic errors, different experimental parameters were varied; for instance material thicknesses and He gas pressure. Different parameters influence the mass dependence and could alter the mass dependencies in the ion stopping efficiency. (orig.)

  13. Fission product release from fuel under LWR accident conditions

    International Nuclear Information System (INIS)

    Three tests have provided additional data on fission product release under LWR accident conditions in a temperature range (1400 to 20000C). In the release rate data are compared with curves from a recent NRC-sponsored review of available fission product release data. Although the iodine release in test HI-3 was inexplicably low, the other data points for Kr, I, and Cs fall reasonably close to the corresponding curve, thereby tending to verify the NRC review. The limited data for antimony and silver release fall below the curves. Results of spark source mass spectrometric analyses were in agreement with the gamma spectrometric results. Nonradioactive fission products such as Rb and Br appeared to behave like their chemical analogs Cs and I. Results suggest that Te, Ag, Sn, and Sb are released from the fuel in elemental form. Analysis of the cesium and iodine profiles in the thermal gradient tube indicates that iodine was deposited as CsT along with some other less volatile cesium compound. The cesium profiles and chemical reactivity indicate the presence of more than one cesium species

  14. 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

  15. 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

  16. 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.

  17. Measurement of the Ratio of Fissions in U238 to Fissions in U233 Using 1.60 Mev Gamma Rays of the Fission Product La140

    International Nuclear Information System (INIS)

    This paper describes a method for measuring δ28, the ratios of fissions in U238 to fissions in U235. The method was developed as a part of the D2O lattice programme at the Massachusetts Institute of Technology (MIT) ; however, it can be used for measurements in any thermal reactor of natural or slightly enriched uranium. The fast fission factor in uranium cannot be measured directly. It is, however, related to δ28 which can be measured: ϵ =1 + Cδ28 , where C is a constant involving nuclear properties of U238 and U235: Previous methods of measuring δ28 utilize a comparison of fission-product gamma or beta activity in foils of differing U235 concentration irradiated within a fuel rod in the lattice. A double fission chamber is then used to relate the U238 and U235 fission product activity to the ratio of the corresponding fission rates. Most of the experimental uncertainty associated with the measurement of δ28 a is generally attributed to the fission chamber calibration. The method developed at MIT avoids the need for a fission chamber calibration and is accomplished directly with foils irradiated within a fuel rod in the lattice. Two foils of differing U235 concentration are irradiated and allowed to cool for at least a week. The relative activity of the 1.60 MeV gamma ray of the fission product La140 is determined for the two foils. This ratio, the foil weights and atomic densities, and the ratio of fission yields β25/β28 for La140 are then used to determine δ28. This value of δ28 is used to calibrate simpler measurements in which the relative gamma activity above 0.72 MeV is determined for sets of foils irradiated in fuel rods of the lattices of interest. The energy 0.72 MeV is a convenient discrimination level, as it is the maximum energy of Bremsstrahlung from 2.3-d Np239. This method appears to offer the advantages of direct measurement and increased accuracy (the major uncertainty being the ratio of β25/β28 La140). In addition, the results can be

  18. 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.

  19. 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.

  20. 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

  1. 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)

  2. Distributions of Fission Products on PCI In Spent PWR Fuel Using EPMA

    International Nuclear Information System (INIS)

    The fuel specimen with 62,000 MWd/tU and the spent failed fuel rod with 53,000 MWd/tU by commercials PWR fuel were examined to compare with oxygen rich and average region at fuel-clad gap. To observe chemical behaviors and distributions of fission products on fuel-clad gap region by EPMA (Electron probe Micro-Analyzer). The results of this study can be use also in the interim storage facilities for spent fuels which were used in the Korea nuclear power plant. In addition, for comparisons of each plant’s spent fuel characteristics this data will be use as a basic material. EPMA technique offers the possibility of identifying and analyzing such phases and segregations in spent PWR fuel, although the small amount expected to be present and the background radiation, present a significant analytical challenge. The detailed characterization of spent fuel fuel-clag gap region of fission products before and after its expose from neutron is an important part of longterm storage of spent fuel. This report presents the results of EPMA examination of a spent fuel specimen with 62,000 MWd/tU performed with the aim of EPMA technique to analyses of fission products on fuel-clad gap region. (author)

  3. 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.

  4. 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)

  5. Methods to Collect, Compile, and Analyze Observed Short-lived Fission Product Gamma Data

    Energy Technology Data Exchange (ETDEWEB)

    Finn, Erin C.; Metz, Lori A.; Payne, Rosara F.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Pierson, Bruce D.; Ellis, Tere A.

    2011-09-29

    A unique set of fission product gamma spectra was collected at short times (4 minutes to 1 week) on various fissionable materials. Gamma spectra were collected from the neutron-induced fission of uranium, neptunium, and plutonium isotopes at thermal, epithermal, fission spectrum, and 14-MeV neutron energies. This report describes the experimental methods used to produce and collect the gamma data, defines the experimental parameters for each method, and demonstrates the consistency of the measurements.

  6. Study of the short-lived fission products. Separation of iodine and xenon fission radionuclides

    International Nuclear Information System (INIS)

    The separation by distillation in a sulfuric acid or phosphoric acid-hydrogen peroxide medium of the iodine isotopes (8 day iodine-131, 2,3 hour iodine-132 21 hour iodine-133, 53 minute iodine-134 and 6,7 hour iodine-135) present in a uranium sample after different irradiation and cooling times is here described. It is also reported the use of active charcoal columns for the retention of xenon isotopes (5,27 days xenon-133 and 9,2 hours xenon-135) either released during the dissolution of the uranium irradiated samples or generated along the fission isobaric chains in the solutions of distillated iodine. In both cases the radiochemical purity of the separated products is established by gamma spectrometry. (Author) 15 refs

  7. Most probable charge of fission products in proton-induced fission of 238U and 232Th

    International Nuclear Information System (INIS)

    The charge distributions of fission products in proton-induced fission of 238U and 232Th 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)

  8. 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)

  9. Chromatographic separation of actinides and fission products from nuclear wastes

    International Nuclear Information System (INIS)

    Although a number of partitioning processes have been proposed and studied to separate the minor actinides (MA: Am, Cm, Np) and some fission product elements (FPs) from nuclear wastes, most of these processes essentially utilize liquid-liquid extraction technology by using a mixture of organic extractants hydrocarbon diluents. A large amount of the secondary waste, which is difficult for treatment and disposal will be generated by the extraction process. Compared to U and Pu, the MA and FPs are significantly abundant in the spent fuel, so that the scale of an efficient partitioning process for nuclear wastes reasonably small and result in less waste amount

  10. Crystallization study of a glass used for fission product storage

    International Nuclear Information System (INIS)

    The vitreous matrix used in France is a borosilicate glass of low melting point allowing introduction of volatil fission products and of good chemical stability. However, like any glass, if storage temperature is higher than transformation temperature a partial crystallization can occur. Before final storage, it is important to determine of leaching by water eventually occuring on the choosen site is modified by crystalline phases. The aim of this study is the determination of the leaching rate and the identification of crystalline phases formed during thermal treatment and evaluation of its volumic fraction

  11. Energy spectra of delayed neutrons from separated fission products. IV

    International Nuclear Information System (INIS)

    Energy spectra of delayed neutrons from the mass-separated fission products 8890Br, 138140I, 142(Xe,Cs) and 144Cs have been measured. Average level spacings, neutron envelopes and Psub(n) values were calculated and compared with the experimental data. The neutron envelopes are well reproduced for all precursors except 90Br and 140I. For the latter the neutron window predicted by various mass formulae is too wide and a considerable reduction was found necessary to bring calculated envelopes in agreement with the experimental distributions. (Auth.)

  12. Extraction chromatographic studies of actinides and fission products using CMPO

    International Nuclear Information System (INIS)

    The uptake behaviour of U(VI), Pu(IV), Am(III), Eu(III), Zr(IV), Fe(III), Ru(III) and Tc from nitric acid medium by octyl (phenyl)-N,N-diisobutyl carbomylmethylphosphine oxide (CMPO) adsorbed on chromosorb has been studied. Actinide metal ions along with rare earths are taken up to a greater extent as compared to the other fission products. The loading experiments have shown that at lower concentrations of the rare earths or U(VI), the uptake of Pu(IV), U(VI) and Am(III) are reasonably high. (author). 3 refs., 1 fig

  13. 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. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 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

  14. Transient fission product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Sweep gas experiments performed at CRL from 1979 to 1985 have been analysed to determine the fraction of the fission product gas inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the xenon release from companion fuel elements and from a well documented experimental fuel bundle irradiated in the NRU reactor. The measured gas release could be matched to within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. (author)

  15. Fission product source terms and engineered safety features

    International Nuclear Information System (INIS)

    New, technically defensible, methodologies to establish realistic source term values for nuclear reactor accidents are discussed. Although these methodologies will undoubtedly find widespread use in the development of emergency response procedures, that is, procedures to be implemented external to the plant, such as sheltering or evacuation of the surrounding population, it is less clear that the industry is preparing to employ the newer results to develop a more rational approach to the implementation of engineered safety features for the mitigation of fission product releases in the event of a nuclear reactor accident

  16. Behavior of Nb fission product during nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Investigations on niobium fission product behavior in nitric acid and tributyl phosphate media have been carried out in order to explain the difficulties encountered in separating this element from fissile materials during spent nuclear fuel reprocessing. The studies have shown that in nitric acid solution, pentavalent niobium has a colloidal hydroxide form. The so-obtained sols were characterized by light scattering, electronic microscopy, electrophoresis and ultracentrifugation methods. In heterogeneous extracting media containing tributyl phosphate and dibutyl phosphoric acid the niobium hydroxide sols could be flocculated by low dibutyl phosphoric acid concentration or extracted into the organic phase containing an excess of dibutyl phosphoric acid

  17. 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)

  18. Purification of uranium from fission products by ammonium uranyl carbonate precipitation

    International Nuclear Information System (INIS)

    Processing of the oxalate filtrate generated in plutonium reconversion laboratory involves recovery of plutonium by uranous oxalate carrier precipitation and uranium by ammonium diuranate precipitation. The ammonium di-uranate precipitate generally carries most of the fission products which are high energy gamma emitters. Purification of uranium from the fission products has been investigated employing ammonium carbonate which dissolves the slurry and re-precipitates uranium as ammonium uranyl carbonate. Fission product decontamination factor has been evaluated, which indicate the possibility of 99.6% recovery and purification of uranium from fission products. This method simplifies the purification process with less man-rem exposure and high quality end product. (author)

  19. Catalytic electrolytic extraction of long-lived fission products

    International Nuclear Information System (INIS)

    An electrolytic extraction method has been studied to separate fission products (Ru, Rh, Pd, Tc, Se, Te, etc) from the nuclear spent fuel. Yet they are rare metal fission products (RMFP), most are long-lived (LLFP; Pd, Tc, Se, Te). In the applied electrochemical separation process, Pd2+ cation itself would not only be easily deposited from various nitric acid solutions, but also enhances the other deposition of RuNO3+ and ReO4 by acting catalyst as Pdadatom. The same role also applies to the case of TcO4 deposition (i.e., CEE: Catalytic Electrolytic Extraction). One of the promising utilizations will be hydrogen production by alkaline or sea water electrolysis as FP-catalyst. The deposits of quaternary alloy consisting of Ru, Rh, Pd and Re show the highest catalytic reactivity, even superior to that of the smooth Pt electrode. Current interests are focused on the separability and catalytic reactivity of Re and Tc. (author)

  20. Fission product release out of the core of a pebble bed reactor in core heatup accidents

    International Nuclear Information System (INIS)

    This report presents the analysis of fission product release from the core of a pebble-bed high temperature reactor during hypothetical accidents. First the models describing fission product transport are discussed, and on the basis of these models a computer code is developped. This code includes the diffusion of fission products from particles and through the graphite, and the sorption of metallic fission product elements on graphite as well as the plateout of metallic fission product elements in the top- and bottom reflectors. In addition a review of the necessary empirical input data is given. Then the cesium release of a single fuel element at high temperatures is calculated, and the results are compared with experimental data. Furthermore calculations of the fission product release from the core of a 500 MW(th) high temperature reactor during core heatup accidents are made, and the influence of the most important parameters is described. (orig.)

  1. Reverse engineering of GETTER : a fission product release code for PBMR / Jeetesh Bhana Keshaw

    OpenAIRE

    Keshaw, Jeetesh Bhana

    2007-01-01

    Fission product release from spherical fuel spheres under different irradiation and heat-up conditions is one of the key criteria used in High Temperature Reactor (HTR) design. Accurate analyses of fuel performance and fission product behaviour is therefore essential in justifying the safe behaviour of the Pebble Bed Modular Reactor (PBMR). GETTER proved to be a very versatile tool for evaluating fission product transport problems; ranging from heating experiments (up to 1 800°C) to full core...

  2. Extraction process of fission products contained in irradiated nuclear fuel elements

    International Nuclear Information System (INIS)

    In the process described, the fission products contained in irradiated nuclear fuel elements are extracted before the fuel is dissolved by wet process. After the element have been mechanically removed from their cladding and/or sliced up, they are processed in water to cause the fission products to be dissolved in an aqueous solution, after which the processed elements are separated from the aqueous solution obtained and at least one of the fission products is retrieved from this aqueous solution

  3. FISPRO: a simplified computer program for general fission product formation and decay calculations

    International Nuclear Information System (INIS)

    This report describes a computer program that solves a general form of the fission product formation and decay equations over given time steps for arbitrary decay chains composed of up to three nuclides. All fission product data and operational history data are input through user-defined input files. The program is very useful in the calculation of fission product activities of specific nuclides for various reactor operational histories and accident consequence calculations

  4. 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)

  5. High flux transmutation of fission products and actinides

    International Nuclear Information System (INIS)

    Long-lived fission products and minor actinides accumulated in spent nuclear fuel of power reactors comprise the major part of high level radwaste. Their incineration is important from the point of view of radwaste management. Transmutation of these nuclides by means of neutron irradiation can be performed either in conventional nuclear reactors, or in specialized transmutation reactors, or in ADS facilities with subcritical reactor and neutron source with application of proton accelerator. Different types of transmutation nuclear facilities can be used in order to insure optimal incineration conditions for radwaste. The choice of facility type for optimal transmutation should be based on the fundamental data in the physics of nuclide transformations. Transmutation of minor actinides leads to the increase of radiotoxicity during irradiation. It takes significant time compared to the lifetime of reactor facility to achieve equilibrium without effective transmutation. High flux nuclear facilities allow to minimize these draw-backs of conventional facilities with both thermal and fast neutron spectrum. They provide fast approach to equilibrium and low level of equilibrium mass and radiotoxicity of transmuted actinides. High flux facilities are advantageous also for transmutation of long-lived fission products as they provide short incineration time

  6. Immobilization of fission products in phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    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 99Tc wastes from sorption processes

  7. Application of inorganic exchangers in fission product separation

    International Nuclear Information System (INIS)

    Synthetic ion exchangers ammonium phosphomolybdate/phosphotungstate (APW), polyantimonic acid (PA) and manganese dioxide have been investigated for separation of cesium, strontium and cerium respectively with a view to their use in fission product separation. Their breakthrough capacities and elution characteristics were determined using 137Cs, sup(85,89)Sr and 141Ce as tracers. Results indicate that : (1) Cs adsorbed on APW is easily eluted with 3M NH4NO3 at a temperature of 500C with an overall yield of 90% in about 10 column volumes, (2) strontium adsorbed on PA is completely eluted by 1M AgNo3 + 8M HNO3 at room temperature and (3) manganese sulphate (1 mg/ml) + 3M HNO3 elutes cerium adsorbed on manganese dioxide. Column characteristics (exchange capacity and flow rate) are not affected upto 6 cycles of sorption-elution. Based on these findings, a scheme of separation of fission products from waste solution is proposed. Pu uptake on PA is found to be governed by U/Pu ratio in the solution. The ratio > 104 inhibits the uptake. Pu on PA is eluted in 10 column volumes by 0.01M ascorbic acid +2M nitric acid. The exchange PA can be used over 20 cycles of sorption-elution. (M.G.B.)

  8. Fission products control by gamma spectrometry in purex process solutions

    International Nuclear Information System (INIS)

    This paper deals with a radiometric method for fission products analysisby gamma spectrometry. This method will be applied for fission productscontrol at the irradiated material processing facility, under construction inthe Instituto de Pesquisas Energeticas e Nucleares, SP, Brazil. Countinggeometry was defined taking into account the activities of process solutionsto be analysed, the remotely operated aliquotation device of analytical celland the available detection system. Natural and 19,91% enriched uraniumsamples were irradiated at IEAR-1 reactor in order to simulate thecomposition of Purex process solutions. After a short decay time, the sampleswere dissolved with HNO3 and then, conditioned in standard flasks withdefined geometry. The spectra were obtained by a Ge(Li) semiconductordetector and analysed by the GELIGAM software system, losing a floppy-diskconnected to a PDP-11/05 computer. Libraries were prepared and calibrationswere made with standard sources to fit the programs to the analysis offission products in irradiated uranium solutions. It was possible to choosethe best program to be used in routine analysis with the obtained data.(author)

  9. Airborne measurements of fission product fall-out

    International Nuclear Information System (INIS)

    During 1993 the Danish Emergency Management Agency will install an airborne γ-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)

  10. 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.

  11. Metabolism of fission products in man: Marshallese experience

    International Nuclear Information System (INIS)

    The medical study of the Marshallese accidentally exposed to local fall-out in 1954 is unique in that; along with he Japanese fishermen study, it provides the only data existing on the metabolism of mixed fission products in a human population. Early diagnosis of the internal radioactive contamination was made by radiochemical analysis of the excreta of the exposed people and by radiochemical analysis of the tissues of animals simultaneously exposed. Initially, Sr89, Ba140, I131 and its shorter-lived daughters and a number of rare-earth elements contributed the major portion of the internal radiation dose. After a year, the principal radioisotopes were Sr90, Cs137 and Zn65. Subsequently these radionuclides and more recently, Co60, have been measured periodically. Since 1958, the γ-spectra of a number of Marshallese have been obtained with a portable whole-body counter. The report discusses the findings of these studies for the past eight years. The results of an early attempt o alter the rate of removal of the mixed fission products in the Marshallese with calcium disodium EDTA are presented. The metabolism of die radionuclides and their relationship to levels present in the environment is also discussed. (author)

  12. Fission Product Fast Reactor Constants System of JNDC

    International Nuclear Information System (INIS)

    The Fission Product Fast Reactor Constants System of JNDC has been developed for providing the FP group constants set rather automatically from the Japanese Evaluated Nuclear Data Library (JENDL). In the present version, the evaluation by JNDC was adopted for the 28 important nuclides and the evaluation by Cook was supplementally used for the other nuclides to obtain the lumped group constants. The burn-up time dependence of the lumped constants were examined. The change of capture cross sections are about 5% between 60 days and 720 days of burn-up for any type of fast reactors. The 28 important nuclides take more than 80% of total capture by fission products and cover 40% of elastic scattering and 60% of inelastic scattering. The JNDC FP lumped constants were compared with those based on Cook's evaluation and on the ENDF/B-4. The discrepancies among the three are 15% for capture and 10% for both of elastic and inelastic scattering. A benchmark test was performed using the integral measurements made in RCN, Petten, the Netherlands, in order to check the reliability of the JNDC FP group constants. The JNDC constants give better agreements than the Cook and ENDF/B-4 constants with the experiments both for FP mixtures and for separated isotopes. (auth.)

  13. Basic requirements for a 1000-MW(electric) class tokamak fusion-fission hybrid reactor and its blanket concept

    International Nuclear Information System (INIS)

    Plasma size and other basic performance parameters for 1000-MW(electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is base don the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. THus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW(electric) hybrids. Typical plasma parameters for 1000-MW(electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of Ip = 15 MA, and toroidal field on the axis of Bo = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-producing hybrids. An optimization study is carried out for this blanket concept. The result shows that a compact, simple structure with a uniform fuel composition for the fissile region is sufficient to obtain optimal conditions for suppressing the thermal power increase caused by fuel burnup. The maximum increase in the thermal power is +3.2%. The M value estimated from the neutronics calculations is ∼7.0, which is confirmed to be compatible with the plasma requirement. These studies show that it is possible to use a tokamak fusion core with design requirements similar to those of ITER for a 1000-MW(electric) power reactor that uses existing thermal reactor technology for the blanket. 30 refs., 22 figs., 4 tabs

  14. 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)

  15. Advanced model for the prediction of the neutron-rich fission product yields

    OpenAIRE

    Rubchenya V.A.; Gorelov D.; Jokinen A.; Penttilä H.; Äystö J.

    2013-01-01

    The consistent models for the description of the independent fission product formation cross sections in the spontaneous fission and in the neutron and proton induced fission at the energies up to 100 MeV is developed. This model is a combination of new version of the two-component exciton model and a time-dependent statistical model for fusion-fission process with inclusion of dynamical effects for accurate calculations of nucleon composition and excitation energy of the fissioning nucleus a...

  16. Assessment of selected fission products in the Savannah River Site environment

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.

  18. 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.

  19. Uncertainty analysis on fission Mo production in HANARO

    International Nuclear Information System (INIS)

    Uncertainty analysis on fission-produced molybdenum production with low enriched uranium (LEU) and high enriched uranium (HEU) was performed using Crude Monte Carlo Method. The most important parameter affecting uncertainty of 99Mo yield and annul production amount was fuel thickness for LEU target. Therefore, it was important to minimize the fuel film fabrication tolerance for LEU target. Uncertainty of minimum required decontamination factor(MRDF) to satisfy U. S. P. (Unites States Pharmacopoeia) standard was very small for both target. Decontamination of Pu which is α-emitter was shown to be impossible using current. Cintichem porcess in LEU. However, it can be overcome by addition of one more purification step, because the uncertainty of MRDF was small within 3% for 10 confidence level

  20. 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

  1. Tables and figures from JNDC Nuclear Data Library of fission products, version 2

    International Nuclear Information System (INIS)

    The content of JNDC (Japanese Nuclear Data Committee) FP (Fission Product) Nuclear Data Library version 2 for 1227 fission products is presented in the form of tables and figures. The library is inclusive of evaluated decay data such as decay constant, Q-value, average energies of beta, gamma and internal conversion electron, spin-parity, branching ratio of each decay mode and fission yield. The neutron capture cross-sections are also contained for 166 nuclides. The mass number of the fission product nuclides ranges from A = 66 to A = 172. (author)

  2. 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

  3. 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

  4. 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.

  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. Behavior of fission products in sulfide reprocessing process

    International Nuclear Information System (INIS)

    For the recovery of nuclear materials from spent fuel with more effective and convenient methods comparing with conventional process, the sulfurization and dissolution behavior of fission products, such as rare-earths, alkali, alkalline-earth and platinum group elements were studied. The sulfurization experiment was carried out using tracer doped U3O8. The samples were reacted with CS2 at temperatures from 573 to 773 K for 1 hour followed by dissolution with 1M nitric acid solution for 1 hour at 323 K. The dissolution ratio for each element was obtained by α- and γ-ray spectrometry. The alkali and alkaline-earth elements show higher dissolution ratios as well as trivalent lanthanide elements. On the other hand, U, Zr, Ce, and Ru showed lower dissolution ratios. These results were in good agreement with those expected from the thermodynamic consideration. (author)

  7. Behaviour of fission-product iodine under severe accident conditions

    International Nuclear Information System (INIS)

    On account of the radiological properties of I-131 the behaviour of fission-product iodine is of great importance under severe reactor accident conditions. The chemical properties of iodine: Its easy conversion into several oxidation compounds, its capability of forming not only volatile (organo-iodide, elemental iodine), hardly volatile, readily soluble (cesium iodide/iodate) but also insoluble (silver iodide) compounds, and its susceptibility to ionizing radiation, are further aspects of significance. Intensive investigations on iodine behaviour under reactor accident conditions carried out worldwide over the last ten years have shown - even though a number of details have yet to be elucidated - that physicochemical processes form a natural, i.e. passive, barrier against the possible release of iodine. (orig.)

  8. Diffusion of Fission Product Elements in Compacted Bentonite

    International Nuclear Information System (INIS)

    Study on diffusion of fission product in compacted bentonite has been conducted. The information about mobilities of these elements have been obtained from the studies resulted in many countries. It is presented that the diffusion coefficient was varied by the function of solution phase condition as well as the nature of bentonite. It is also showed that the diffusion coefficient decreased by the increasing of density, as well as the increasing of montmorillonite content in bentonite. The ratio of bentonite/silica-sand used, was related to the increasing of elements mobility. In many case variation of diffusion coefficient was related to the variation of pH, redox condition, and the presence of complex ant in solution phase. The lower diffusion coefficient could give the higher retardation factor, which is a favorable factor to retard the radionuclides release from a disposal facility to geosphere. (author)

  9. Treatment of gaseous wastes in vitrification plants for fission products

    International Nuclear Information System (INIS)

    In order to solidify highly active fission product solutions from reprocessing of nuclear fuels, a discontinuous as well as continuous vitrification process has been developed and the appropriate plants been put into operation in Marcoule, France. The waste gases formed in the part processes of vitrification, evaporation, calcination and melting, are described according to their origin, chemical composition and their technical, chemical, radioactive or toxic effects and the effectivity of the equipment used in both methods to purify the waste gas are demonstrated. The behaviour and treatment of the volatile ruthenium, fluorine and mercury, as well as volatile components in the molten glass which are released when filling into storage containers (Cs 137, Ce 144, Ru 106) are particularly dealt with. In a bad case, decontamination factors of 1010 are reached. (RB)

  10. A review of libraries of fission product yields

    International Nuclear Information System (INIS)

    Several libraries of fission product yields are in use internationally. This paper summarizes and compares Chinese, French, UK and US libraries. These, being in the same format, can be quite readily compared. The different methods and philosophies of evaluation are reviewed, especially as they affect the recommended uncertainties. Detailed comparisons of the libraries are presented, and some of the larger differences studied in depth. The effects of any discrepancies on decay heat calculations are discussed. It is also noted that differences in uncertainties in yield data lead to some differences in uncertainties in summation calculations. There is great advantage in maintaining at least two independent yield libraries, and it is hoped that the libraries described will be continually improved and updated. Suggestions for improvements in evaluation methods, and for collaboration at various pre-evaluation stages are made

  11. 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)

  12. 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

  13. 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

  14. A revised ANS standard for decay heat from fission products

    International Nuclear Information System (INIS)

    The draft ANS 5.1 standard on decay heat was published in 1971 and given minor revision in 1973. Its basis was the best estimate working curve developed by K. Shure in 1961. Liberal uncertainties were assigned to the standard values because of lack of data for short cooling times and large discrepancies among experimental data. Research carried out over the past few years has greatly improved the knowledge of this phenomenon and a major revision of the standard has been completed. Very accurate determination of the decay heat is now possible, expecially within the first 104 seconds, where the influence of neutron capture in fission products may be treated as a small correction to the idealized zero capture case. The new standard accounts for differences among fuel nuclides. It covers cooling time to 109 seconds, but provides only an ''upper bound'' on the capture correction in the interval 104 9 seconds. (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. Basics of Fusion-Fission Research Facility (FFRF) as a Fusion Neutron Source

    International Nuclear Information System (INIS)

    FFRF, standing for the Fusion-Fission Research Facility represents an option for the next step project of ASIPP (Hefei, China) aiming to a first fusion-fission multifunctional device (1). FFRF strongly relies on new, Lithium Wall Fusion plasma regimes, the development of which has already started in the US and China. With R/a=4/1m/m, Ipl=5 MA, Btor=4-6 T, PDT=50- 100 MW, Pfission=80-4000MW, 1 m thick blanket, FFRF has a unique fusion mission of a stationary fusion neutron source. Its pioneering mission of merging fusion and fission consists in accumulation of design, experimental, and operational data for future hybrid applications.

  17. ZZ ORYX-E/38B, Group Constant Library from ENDF/B Fission Product Data for ORIGEN Calculation

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: Format: ORIGEN; Number of groups: 124 energy groups; Nuclides: H, He, Li, Be, B, C, N, O, F, Ne, Na, Mg, Al, Si, P, S, Cl, Ar, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Hf, Ta, W, Re, Os, Ir, Pt, Au, Hg, Tl, Pb, Bi, Po. Origin: ENDF/B-IV; Weighting spectrum: Maxwellian (1/E) fission spectrum with a one percent tolerance. ORYX-E increases the versatility of the program ORIGEN , the isotope generation and depletion code package by providing basic cross section and decay information for light element, fission-product, and actinide nuclides. This data library package results from data compiled for ORNL Chemical Technology Division's work with ORIGEN and from a 2-year effort of the cross section evaluation working group (CSEWG) fission product task force. 2 - Method of solution: The data is generated from ENDF/B-IV and is formatted for input to the ORIGEN code. Applications include calculations for waste projection, decay heat, nuclear safeguards, and fuel cycle economics. The data library is generated from the ENDF/B-IV fission product data. The capture cross section of all fission product nuclides for which capture cross section information is given (about 180 nuclides) were processed into 124 energy groups using MINX. Multigroup cross sections were generated at 0 degrees with infinite dilution and one broad thermal group. Fine group data was generated using a Maxwellian (1/E) fission spectrum with a one percent tolerance

  18. The behaviour of fission products in the HTGR fuel irradiated in the IVV-2M reactor

    International Nuclear Information System (INIS)

    The results of the post-irradiation investigations of fission products behaviour in HTGR fuel and its main elements such as kernels, protective coatings and matrix graphite are considered. The dominating role of SiC layer in the protective coating of coated particles in the retention of the volatile and solid fission products, being of great radiological importance, is noticed. (author)

  19. The Phebus Fission Product and Source Term International Programmes

    International Nuclear Information System (INIS)

    The international Phebus FP programme, initiated in 1988 is one of the major research programmes on light water reactors severe accidents. After a short description of the facility and of the test matrix, the main outcomes and results of the first four integral tests are provided and analysed. Several results were unexpected and some are of importance for safety analyses, particularly concerning fuel degradation, cladding oxidation, chemical form of some fission products, especially iodine, effect of control rod materials on degradation and chemistry, iodine behaviour in the containment. Prediction capabilities of calculation tools have largely been improved as a result of this research effort. However, significant uncertainties remain for a number of phenomena, requiring detailed physical analysis and implementation of improved models in codes, sustained by a number of separate-effect experiments. This is the subject of the new Source Term programme for a better understanding of the phenomenology on important safety issues, in accordance with priorities defined in the EURSAFE project of the 5th European framework programme aiming at reducing the uncertainties on Source Term analyses. It covers iodine chemistry, impact of boron carbide control rods degradation and oxidation, air ingress situations and fission product release from fuel. Regarding the interpretation of Phebus, an international co-operation has been established since over ten years, particularly helpful for the improvement and common understanding of severe accident phenomena. Few months ago, the Phebus community was happy to welcome representatives of a large number of organisations from the following new European countries: the Czech republic, Hungary, Lithuania, Slovakia, Slovenia and also from Bulgaria and Romania. (author)

  20. 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)

  1. 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)

  2. Fission product release analysis code during accident conditions of HTGR, RACPAC

    International Nuclear Information System (INIS)

    Fission product release analysis code, RACPAC (Fission Product Release Analysis Code from Fuel Particle in Accident Condition), was developed to calculate fractional release from the core during accident conditions of High Temperature Gas-cooled Reactor. RACPAC code has following features. (1) Fission product release fraction after the reactor scram is calculated based on the analytical solution with reduced diffusion coefficient. (2) The reduced diffusion coefficient for each nuclide is calculated from the (R/B) value, which is defined as release rate to birth rate of fission product. (3) The temperature transient after the accident can be taken into consideration in fractional release calculation with RACPAC. This paper describes calculation model of fission product release from fuel particle, calculation model of the reduced diffusion coefficient, users' manual and calculation examples. (author)

  3. 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

  4. Advanced model for the prediction of the neutron-rich fission product yields

    Directory of Open Access Journals (Sweden)

    Rubchenya V. A.

    2013-12-01

    Full Text Available The consistent models for the description of the independent fission product formation cross sections in the spontaneous fission and in the neutron and proton induced fission at the energies up to 100 MeV is developed. This model is a combination of new version of the two-component exciton model and a time-dependent statistical model for fusion-fission process with inclusion of dynamical effects for accurate calculations of nucleon composition and excitation energy of the fissioning nucleus at the scission point. For each member of the compound nucleus ensemble at the scission point, the primary fission fragment characteristics: kinetic and excitation energies and their yields are calculated using the scission-point fission model with inclusion of the nuclear shell and pairing effects, and multimodal approach. The charge distribution of the primary fragment isobaric chains was considered as a result of the frozen quantal fluctuations of the isovector nuclear matter density at the scission point with the finite neck radius. Model parameters were obtained from the comparison of the predicted independent product fission yields with the experimental results and with the neutron-rich fission product data measured with a Penning trap at the Accelerator Laboratory of the University of Jyväskylä (JYFLTRAP.

  5. Effect of fission product internal conversion on degree of their ionization

    International Nuclear Information System (INIS)

    Experimental study of distribution fine structure on ionization degree of a heavy group (A=131-145) separated by mass of 241Pu fission products at three values of kinetic energy, i.e. Ek=67, 70 and 73 MeV. Their structure smoothing and the separation of atomic and nuclear ionization effects of fission products was performed by their additional recharging in t -10 s after fission. It has been shown that the contribution of the nuclear effect of ionization depends on excitation energy of fission products and their nuclear structure attaining ∼ 50% for certain masses. Experimental proofs of the essential role of nuclear transition internal conversion of fission products in the formation of their electron shell have been obtained

  6. In the eighties French project for fission molybdenum-99 production

    International Nuclear Information System (INIS)

    In 1985, French Atomic Energy Commission decided to implant on the site of Saclay a production unit of fission Molybdenum-99 for supplying ORIS with this isotope for medical applications. A structure of project has been created where the Department of Radioactive Engineering was committed to define and entire new process of reprocessing of uranium in view of production of 2000 Ci 99Mo per week involving particularly severe work constraints. The production unit working in semi-automatic has been never permitted to operate in active. This article focuses mainly on the R ampersand D developed during this period to define and validate the overall process. The reprocessing diagram presents two originalities; the head-end a dissolution of the irradiated 235U/Al targets in sulfuric acid, warranting the keeping iodine in iodide and tellurium at elementary state permitting in this way their easy elimination as Te(0) and AgI, in respect to the safety constraints; the purification of molybdenum with the tributlyaceto-hydroxamine (TBAH), extractant of an exceptional affinity for Mo, used in chromatographic extraction allowing the direct extraction (with yields close to 1) of this element from the dissolution solution relieved of iodine and tellurium and without uranium extraction

  7. Price of fission product transmutation in power reactors

    International Nuclear Information System (INIS)

    The opportunity of Tc-99 and I-129 transmutation in Russian pressure water VVER-1000 reactor is discussed in this paper. Study of long-lived fission product transmutation shows that if Tc-99 or I-129 are located in VVER-type reactor for a total lifetime, then lifetime and burnup are reduced because of additional capture of neutrons. The reduction is proportional to the incinerated mass of nuclide. Transmutation of either 46.8 kg of Tc-99 or 45.8 kg of I-129 causes a reduction of burnup by 2.25 GW.d/ton that is 5.6 % with respect to the burnup without transmutation. This corresponds to a loss of electric power production of 49 GW.d. If both 42 kg Tc-99 and 43 kg I-129 are transmuted, then reduction of burnup is 4.51 GW.d/ton that is 11.3 % of the burnup without transmutation. Loss of electric power production is 99 GW.d. The result does not practically depend on way of transmuted nuclide placement. This loss of power is a price that should be paid for transmutation of nuclides without their removal during reactor operation. One would avoid the reduction of burnup and power loss if it would be possible to find such way of nuclide placement for irradiation in reactor, which would permit to extract nuclides from operating reactor a certain time before next fuel reloading

  8. 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.

  9. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

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

    International Nuclear Information System (INIS)

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

  11. 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

  12. Evaluation of Neutron Induced Reactions for 32 Fission Products

    International Nuclear Information System (INIS)

    Neutron cross sections for 32 fission products were evaluated in the neutron-incident energy range from 10-5 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, 95Mo, 101Ru, 103Rh, 105Pd, 109Ag, 131Xe, 133Cs, 141Pr, 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 sections for almost all reaction channels including photon

  13. Measurement of mass and isotopic fission yields for heavy fission products with the LOHENGRIN mass spectrometer

    International Nuclear Information System (INIS)

    In spite of the huge amount of fission yield data available in different libraries, more accurate values are still needed for nuclear energy applications and to improve our understanding of the fission process. Thus measurements of fission yields were performed at the mass spectrometer Lohengrin at the Institut Laue-Langevin in Grenoble, France. The mass separator Lohengrin is situated at the research reactor of the institute and permits the placement of an actinide layer in a high thermal neutron flux. It separates fragments according to their atomic mass, kinetic energy and ionic charge state by the action of magnetic and electric fields. Coupled to a high resolution ionization chamber the experiment was used to investigate the mass and isotopic yields of the light mass region. Almost all fission yields of isotopes from Th to Cf have been measured at Lohengrin with this method. To complete and improve the nuclear data libraries, these measurements have been extended in this work to the heavy mass region for the reactions 235U(nth,f), 239Pu(nth,f) and 241Pu(nth,f). For these higher masses an isotopic separation is no longer possible. So, a new method was undertaken with the reaction 239Pu(nth,f) to determine the isotopic yields by spectrometry. These experiments have allowed to reduce considerably the uncertainties. Moreover the ionic charge state and kinetic energy distributions were specifically studied and have shown, among others, nanosecond isomers for some masses. (author)

  14. Production and study of fission fragments, from Lohengrin to Alto

    International Nuclear Information System (INIS)

    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

  15. Isotope production in light charged particle induced fission

    International Nuclear Information System (INIS)

    In this thesis the production of neutron-rich isotopes in the mass region A=96-120 has been studied. Their yields have been extensively studied with the IGISOL isotope separator in the 238U (p,f) reaction. Deuteron and alpha particle induced reactions have also been investigated. In connection with this work several new isotopes have been identified for the first time. Specifically, the decays of 110Tc and 112Tc are discussed. Cumulative mass distributions as well as independent isotopic distributions have been constructed. In the method used here the total kinetic energy of the fragments is integrated so that there is no energy selection and the yields are post-neutron emission values. A theoretical model described in one of the joined papers has been used to extract the preneutron emission yields. These results can be used for estimating the possibilities of production of new neutron-rich isotopes, e.g. at the IGISOL facility. In addition, they are necessary to help refining the existing fission models. (orig.) (50 refs., 9 figs.)

  16. 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.

  17. 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

  18. Fission product yield distribution in the 12, 14, and 16 MeV bremsstrahlung-induced fission of 232Th

    International Nuclear Information System (INIS)

    The absolute cumulative yields of various fission products in the 12, 14, and 16 MeV bremsstrahlung-induced fission of 232Th were determined using a recoil catcher and an off-line γ -ray spectrometric technique using the ELBE electron linac of Helmholtz-Zentrum Dresden-Rossendorf in Dresden, Germany. The mass chain yields were obtained from the absolute cumulative yields by correcting the charge distribution. The peak-to-valley ratio, average light mass (left angle AL right angle) and heavy mass (left angle AH right angle) values, and average number of neutrons (left angle n right angle exp) in the bremsstrahlung-induced fission of 232Th at different excitation energies were obtained from the mass chain yield data. The present study and existing literature data for the 232Th(γ, f) reaction are compared with similar data for the 238U(γ, f) reaction at various excitation energies, and surprisingly different behavior was found in the two fissioning systems. (orig.)

  19. Fission-product energy release for times following thermal-neutron fission of 235U between 2 and 14000 seconds

    International Nuclear Information System (INIS)

    Fission-product decay energy-releases rates were measured for thermal-neutron fission of 235U. Samples of mass 1 to 10 μg were irradiated for 1 to 100 sec by use of the fast pneumatic-tube facility at the Oak Ridge Research Reactor. The resulting beta- and gamma-ray emissions were counted for times-after-fission between 2 and 14,000 seconds. The data were obtained for beta and gamma rays separately as spectral distributions, N(E/sub γ/) vs E/sub γ/ and N(E/sub beta/) vs E/sub β/. For the gamma-ray data the spectra were obtained by using a NaI detector, while for the beta-ray data the spectra were obtained by using an NE-110 detector with an anticoincidence mantle. The raw data were unfolded to provide spectral distributions of modest resolution. These were integrated over E/sub γ/ and E/sub β/ to provide total yield and energy integrals as a function of time after fission. Results are low compared to the present 1973 ANS Decay-heat standard. A complete description of the experimental apparatus and data-reduction techniques is presented. The final integral data are given in tabular and graphical form and are compared with published data. 41 figures, 13 tables

  20. Targets development of low enrichment for production of Mo99 for fission

    International Nuclear Information System (INIS)

    The purpose of this work is the replacement of the current targets of High Enrichment Uranium (HEU-90% p. 235U) by a uranium compound of low enrichment (LEU-20% p. 235U) for production of fission Mo99, using a dispersed uranium Aluminum phase. This work describes the manufacturing of UAl2 compound and the powder metallurgical compact process with an approximate density of 3 gU/cm3 in the targets. With this load of uranium reached in the targets, Uranium Aluminum can be replaced the current miniplate of High Enrichment Uranium (HEU), without any geometric modifications and maintaining the current process of basic dissolution in the radiochemical process. The Uranium Aluminide compound was selected due to its very convenient characteristic, specially in density it is 6,42 g/cm3. (author)

  1. 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

  2. Fuel-fission product interaction from oxide to metallic fuels - a paradigm shift

    International Nuclear Information System (INIS)

    Nuclear fuel is a complex material system undergoing massive changes during its lifetime. Currently thermal reactors are the main contributor to nuclear energy generation. In order to cater to the rapid growth in nuclear energy in the coming years Fast Breeder Reactors (FBR) have been inducted in our nuclear programme. Conventional fast reactors use ceramic fuels namely oxides, carbides, nitrides. Fuel behavior under normal operating conditions as well as off-normal conditions can be predicted from accurate measurements of thermophysical and thermodynamic data. Plethora of thermodynamic and thermophysical data has been generated in the triangle of fuel-fission product-cladding interactions. This data is useful not only for fuel designers but also at the back end of the fuel cycle, in the design of nuclear fuel processing methods and for the rationalization of fuel cladding interactions. Thus compounds that can form as a result of fission product-fuel, fission product-cladding and fission product-fission product interactions were investigated. The systems studied due to fission product fuel interactions are M-U-O, M-Th-O, (M= alkali or alkaline earths) U-Te-O, Th- Te-O, U-Ru, U-Rh, U-Pd and due to fission product-cladding interactions are: Fe-Te, Ni-Te, Cr-Te and between fission products themselves are the systems M-Zr-O and M-Mo-O as in a typical fuel element, the inner cladding scavenges excess oxygen by forming an oxide layer. In addition, a portion of the oxygen not bound to fission products remains in the fuel, thereby converting UO2 to UO2+x

  3. 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.

  4. 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.

  5. Identifying and quantifying short-lived fission products from thermal fission of HEU using portable HPGe detectors

    Energy Technology Data Exchange (ETDEWEB)

    Pierson, Bruce D.; Finn, Erin C.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Kephart, Rosara F.; Metz, Lori A.

    2013-03-01

    Due to the emerging potential for trafficking of special nuclear material, research programs are investigating current capabilities of commercially available portable gamma ray detection systems. Presented in this paper are the results of three different portable high-purity germanium (HPGe) detectors used to identify short-lived fission products generated from thermal neutron interrogation of small samples of highly enriched uranium. Samples were irradiated at the Washington State University (WSU) Nuclear Radiation Center’s 1MW TRIGA reactor. The three portable, HPGe detectors used were the ORTEC MicroDetective, the ORTEC Detective, and the Canberra Falcon. Canberra’s GENIE-2000 software was used to analyze the spectral data collected from each detector. Ultimately, these three portable detectors were able to identify a large range of fission products showing potential for material discrimination.

  6. Deep Atomic Binding (DAB) Approach in Interpretation of Fission Products Behavior in Terrestrial and Water Ecosystems

    International Nuclear Information System (INIS)

    A large number of studies and models were established to explain the fission products (FP) behavior within terrestrial and water ecosystems, but a number of behaviors were non understandable, which always attributed to unknown reasons. According to DAB hypothesis, almost all fission products behaviors in terrestrial and water ecosystems could be interpreted in a wide coincidence. The gab between former models predictions, and field behavior of fission products after accidents like Chernobyl have been explained. DAB represents a tool to reduce radio-phobia as well as radiation protection expenses. (author)

  7. Determination of isobar composition and yields of 239Pu fission-products by thermal neutrons

    International Nuclear Information System (INIS)

    On the research nuclear reactor WWR-SM of INP Uz AS by means of mass-spectrometer the heavy fission-products of 239Pu nuclei induced by thermal neutrons are measured in ranges of mass Ai = 125 -157, kinetic energies Ek = 45 - 87 MeV and effective ionic charges z* = 18 - 30. 102 isobar nuclei in composition of the measured fission-products, also the partial yields of the each element giving the contribution to formation of a total yield of heavy fission-product with mass Ai are defined. (authors)

  8. 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.

  9. Fission Product Radioactivity in the Air along the 80th Meridian — January—June 1957

    OpenAIRE

    Lockhart, L. B.; Baus, R. A.; Blifford Jr., I. H.

    2011-01-01

    Measurements of gross fission product radioactivity in the air at a number of sites along the 80th meridian (west) are reported for the period January—June 1957. The concentration of long-lived radioactive products (primarly fission products) in the air continues to remain considerably higher in the Northern than in the Southern Hemisphere. Among the more interesting developments, there has been obtained a definite inverse relation between the air concentration of radioactivity and rai...

  10. 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

  11. Chemical form of fission products in high burnup fuels

    International Nuclear Information System (INIS)

    In order to make a proper assessment of candidate materials for advanced high-burnup fuels, thermochemical studies of fuel materials have been performed. Using data from the ECN thermochemical database (TBASE), which has been updated and extended for the present work, the suitability of various advanced fuel materials and inert matrices is studied. Detailed thermodynamic equilibrium calculations are performed for Pu0.42U0.58O2 and Pu0.40U0.60N for values of the burnup up to 200 MWd/kgHM. The formation of metallic phases, the pressure buildup and the stability of nitride or oxide phases is studied for each fuel type. The results for the chemical form of the solid fission products are given. The chemical aspects of the use of the inert matrix spinel (MgAl2O4) in combination with oxide fuel will be discussed. Experimental research on the compatibility of various types of inert matrices (nitrides, spinel) is in progress at ECN. (author)

  12. Fission product release from core-concrete mixtures

    International Nuclear Information System (INIS)

    The objective of this research is to measure the amount of strontium, barium, and lanthanum that is vaporized from core-concrete mixtures. The measurements are being done using a transpiration method. Mixtures of limestone-aggregated concrete, urania doped with a small amount of La, Sr, Ba, and Zr oxides, and stainless steel were vaporized at 2150 K from a zirconia crucible into flowing He-6% H2-0.06% H2O (a partial molar free energy of oxygen of -420 kJ). The amounts that were vaporized was determined by weight change and by chemical analyses on condensates. The major phases present in the mixture were inferred from electron probe microanalysis (EPM). They were: (1) urania containing calcia and zirconia, (2) calcium zirconate, (3) a calcium magnesium silicate, and (4) magnesia. About 10% of the zirconia crucible was dissolved by the concrete-urania mixture during the experiment, which accounts for the presence of zirconia-containing major phases. To circumvent the problem of zirconia dissolution, we repeated the experiments using mixtures of the limestone-aggregate concrete and the doped urania in molybdenum crucibles. These studies show that thermodynamic calculations of the release of refractory fission products will yield release fractions that are a factor of sixteen too high if the effects of zirconate formation are ignored

  13. Data summary report for fission product release Test VI-7

    International Nuclear Information System (INIS)

    Test VI-7 was the final test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the Monticello boiling water reactor (BWR). The fuel had experienced a burnup of ∼-40 Mwd/kg U. It was heated in an induction furnace for successive 20-min periods at 2000 and 2300 K in a moist air-helium atmosphere. Integral releases were 69% for 85Kr, 52% for 125Sb, 71% for both 134Cs and 137Cs, and 0.04% for 154Eu. For the non-gamma-emitting species, release values for 42% for I, 4.1% for Ba, 5.3 % for Mo, and 1.2% for Sr were determined. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.89 g, with 37% being collected on the thermal gradient tubes and 63% downstream on filters. Posttest examination of the fuel specimen indicated that most of the cladding was completely oxidized to ZrO2, but that oxidation was not quite complete at the upper end. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL-Booth Model

  14. Fission-product tellurium and cesium telluride chemistry revisited

    International Nuclear Information System (INIS)

    The chemistry of fission-product tellurium is discussed with a focus on conditions in an operating CANDU reactor and in an accident scenario, i.e., a loss of coolant accident (LOCA). Cesium telluride, Cs2Te, is likely to be one of the most abundant tellurium species released to containment. Available thermodynamic data on gas phase Cs2Te is not complete; hence the volatility of cesium telluride was studied by Knudsen-cell mass spectrometry. Cesium telluride was found to vapourize incongruently, becoming more tellurium-rich in the condensed phase as vapourization progressed. Vapour-phase species that were observed were elemental cesium and tellurium, CsTe, Cs2Te, Cs2Te2 and Cs2Te3. Second-law enthalpies and entropies were obtained for many of these species, and a third-law value, ΔH298o, of 186 ± 2 kJ·mol-1 was obtained for Cs2Te. (author)

  15. 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

  16. Fission product separation from seawater by electrocoagulation method

    International Nuclear Information System (INIS)

    At the Fukushima Daiichi nuclear power station, seawater was urgently injected into the reactor core. Therefore a large amount of seawater containing highly radioactive fission products (FP) accumulated and its treatment has been a serious problem. FP such as Cs, Sr and I in water are generally removed by an ion exchanger such as zeolite and separated with column or chemical precipitation methods. An alternative electrocoagulation method, which efficiently separates fine particles from the liquid phase without a chemical reagent is expected to be part of a useful separation system that can reduce the amount of waste, decrease processing time and simplify the process. In this study, powdered adsorbents, such as ferrocyanide and zeolite, were added to seawater containing simulated FP, and the electrocoagulation effect with Al alloy electrodes were investigated. More than 99 % of Cs and 90 % of I were removed by potassium nickel hexacyanoferrate(II) and silver zeolite, respectively. Sedimentation was promoted by electrocoagulation and addition of an inorganic cohesion promoter further increased the sedimentation rate. Moreover, rapid dissolution reaction with heating of the aggregation substance was not observed, so the thermal risk of aqueous processing of it would be low. In addition, thermal analyses showed that the electrocoagulation process did not lead to thermal decomposition. Therefore, if the electrocoagulation method is applied to a decontamination system, it has the potential to thermally stabilize and reduce waste. (author)

  17. 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

  18. 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

  19. 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

  20. 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)

  1. Thermal decomposition of fission product nitrates and their reaction with glass batch additives Part A

    International Nuclear Information System (INIS)

    Thermal decomposition of fission product nitrates constituting a 'Purex' waste and their reaction with silica, alumina and aluminosilicate (Al2O3.2SiO2) have been studied. Oxides of nitrogen may be released from fission product nitrates in the temperature range (150 - 850degC) either due to their thermal decomposition or due to the reaction of alkali nitrates with silica or alumina of the glass batch additives. Fission product nitrates, CaNO3 and RbNO3 excepting, react very slowly with silica, the principal glass forming additive for a silicate glass, at temperatures of 1000degC. Complete reaction of Gr.II-Gr.IV metal nitrates with silica may be achieved at temperatures of 1400degC which is too high for a fission product immobilisation process. The alkali nitrates react very rapidly with boric oxide, silica, alumina and metakaolin (Al2O3.2SiO2) at subsolidus temperatures forming the alkali metaborate, alkali disilicate, alkali monoaluminate and alkali alumino disilicate respectively. A phenomenological description of the heating-up processes for fission product nitrate with silicate glass batch additives has been included. It has been very strongly recommended, that, if a silicate matrix is selected for incorporation of fission products, the glass melting should be carried out until a homogeneous or nearly homogeneous product is obtained and should not be stopped at a stage where a sintered or semivitreous mass is the resultant product. (author)

  2. 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

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

    International Nuclear Information System (INIS)

    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

  4. Strategic recycling of fission products in nuclear fuel cycle as for hydrogen production catalyst

    International Nuclear Information System (INIS)

    The catalytic electrolytic extraction (CEE) method has been studied as a separation tool for rare metal fission products [RMFP - Ru, Rh, Pd*, Tc*, Se* and Te* (*LLFP)] in spent nuclear fuel. In an employed CEE process, Pd2+ cation itself would not only be easily deposited from various nitric acid solutions, but would also enhance the deposition of co-existing RuNO3+, ReO-4 and 99TcO-4 by acting as a catalyst (as Pdadatom). The quaternary, Pd-Ru-Rh-Re, deposit Pt or Ti electrode, fabricated by CEE, suggested the highest cathodic current corresponding to the hydrogen generation reaction in both alkaline solution and seawater. The advanced ORIENT cycle, where ion exchange chromatography using tertiary pyridine resin and the CEE is employed as a mainstay separation technology, will enhance separation and utilisation of actinide and fission products, and thus be expected to realize an ultimate reduction of radioactive wastes. (authors)

  5. Experience in monitoring the BWR fuel behaviour and fission product releases during off-normal conditions

    International Nuclear Information System (INIS)

    Tarapur Atomic Power Station has accumulated over 33 reactor years of operating experience in monitoring Boiling Water Reactor fuel behaviour. The sudden and sharp increases in the fission product releases were experienced in the earlier years due to gross fuel failures caused by mechanical damage (lifting of certain core internals) or due to certain operating practices and transients. Data on fission product releases under such gross fuel failure conditions is presented and discussed with evaluation of the incident and corrective actions taken. An apparent correlation observed in incidents of fuel failures and certain operating system transients are discussed. Conventionally for the BWRs sum of six fission gas release rate measured at Steam Jet Air Ejectors is correlated with fission gas radiation monitor reading to work out alarm and trip settings - modifications are suggested to improve reliability and effectiveness in monitoring of fission gas release rates. Appropriate data for fission product deposition and characterisation of coolant crud long lived fission products is also presented. (author). 6 refs, 15 figs, 9 tabs

  6. A comparison of fusion breeder/fission client and fission breeder/fission client systems for electrical energy production

    International Nuclear Information System (INIS)

    A parametric study that evaluated the economic performance of breeder/client systems is described. The linkage of the breeders to the clients was modelled using the stockpile approach to determine the system doubling time. Since the actual capital costs of the breeders are uncertain, a precise prediction of the cost of a breeder was not attempted. Instead, the breakeven capital cost of a breeder relative to the capital cost of a client reactor was established by equating the cost of electricity from the breeder/client system to the cost of a system consisting of clients alone. Specific results are presented for two breeder/client systems. The first consisted of an LMFBR with LWR clients. The second consisted of a DT fusion reactor (with a 238U fission suppressed blanket) with LWR clients. The economics of each system was studied as a function of the cost of fissile fuel from a conventional source. Generally, the LMFBR/LWR system achieved relatively small breakeven capital cost ratios; the maximum ratio computed was 2.2 (achieved at approximately triple current conventional fissile material cost). The DTFR/LWR system attained a maximum breakeven capital cost ratio of 4.5 (achieved at the highest plasma quality (ignited device) and triple conventional fissile cost)

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

    OpenAIRE

    Äystö J.; Penttilä H.; Gorelov D.; Rubchenya V.A.

    2012-01-01

    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-compoun...

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

    International Nuclear Information System (INIS)

    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

  9. 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)

  10. 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.)

  11. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel cladding material. The release of fission products from the fuel plate at temperature below 500 deg. C was found negligible. The first rapid release of fission products was observed with the occurrence of blistering at 561±1 deg. C on the plates. The next release at 585. C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 deg. C of U-Alx. The released material was mostly xenon, but small amounts of iodine and cesium were observed. (author)

  12. Analysis of reactor coolant system depressurization effect of ex-vessel release of fission products

    International Nuclear Information System (INIS)

    Coupling model of thermal-hydraulic, fission products behavior and radiological consequences assessment was constructed. Based on high-pressure core melt severe accidents of SB-LOCA, SGTR, SBO, and LOFW, the influence of reactor coolant system (RCS) depressurization on ex-vessel release of fission products was studied, including mitigation effect on ex-vessel release of fission products and other negative effects. It is shown that RCS depressurization can mitigate ex-vessel release of fission products for high-pressure core melt accident sequences, while airborne activity in the early phase with RCS depressurization is higher than that of the base cases without RCS depressurization. The research results can give support to establish severe accident management guideline. (authors)

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

    International Nuclear Information System (INIS)

    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

  14. 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)

  15. 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)

  16. 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)

  17. A model for non-volatile fission product release during reactor accident conditions

    International Nuclear Information System (INIS)

    An analytical model has been developed to describe the release kinetics of non-volatile fission products (e.g., Mo, Ce, Ru and Ba) from uranium dioxide fuel under severe reactor accident conditions. The present treatment considers the rate-controlling process of release in accordance with diffusional transport in the fuel matrix and fission product vaporization from the fuel surface into the surrounding gas atmosphere. The effect of the oxygen potential in the gas atmosphere on the chemical form and volatility of the fission product is considered. A correlation is also developed to account for the trapping effects of Sb and Te in the Zircaloy cladding. This model has been used to interpret the release behaviour of fission products observed in the CEA experiments conducted in the HEVA/VERCORS facility at high temperature in a hydrogen and steam atmosphere. (author)

  18. 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

  19. 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

  20. 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

  1. 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)

  2. 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 \

  3. Investigation of Inconsistent ENDF/B-VII.1 Independent and Cumulative Fission Product Yields with Proposed Revisions

    International Nuclear Information System (INIS)

    A recent implementation of ENDF/B-VII.1 independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear schemes in the decay sub-library that are not reflected in legacy fission product yield data. Recent changes in the decay data sub-library, particularly the delayed neutron branching fractions, result in calculated fission product concentrations that do not agree with the cumulative fission yields in the library as well as with experimental measurements. To address these issues, a comprehensive set of independent fission product yields was generated for thermal and fission spectrum neutron-induced fission for 235,238U and 239,241Pu in order to provide a preliminary assessment of the updated fission product yield data consistency. These updated independent fission product yields were utilized in the ORIGEN code to compare the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. Another important outcome of this work is the development of fission product yield covariance data necessary for fission product uncertainty quantification. The evaluation methodology combines a sequential Bayesian method to guarantee consistency between independent and cumulative yields along with the physical constraints on the independent yields. This work was motivated to improve the performance of the ENDF/B-VII.1 library for stable and long-lived fission products. The revised fission product yields and the new covariance data are proposed as a revision to the fission yield data currently in ENDF/B-VII.1

  4. Investigation of inconsistent ENDF/B-VII.1 independent and cumulative fission product yields with proposed revisions

    Energy Technology Data Exchange (ETDEWEB)

    Pigni, Marco T [ORNL; Francis, Matthew W [ORNL; Gauld, Ian C [ORNL

    2015-01-01

    A recent implementation of ENDF/B-VII. independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear scheme in the decay sub-library that is not reflected in legacy fission product yield data. Recent changes in the decay data sub-library, particularly the delayed neutron branching fractions, result in calculated fission product concentrations that are incompatible with the cumulative fission yields in the library, and also with experimental measurements. A comprehensive set of independent fission product yields was generated for thermal and fission spectrum neutron induced fission for 235,238U and 239,241Pu in order to provide a preliminary assessment of the updated fission product yield data consistency. These updated independent fission product yields were utilized in the ORIGEN code to evaluate the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. An important outcome of this work is the development of fission product yield covariance data necessary for fission product uncertainty quantification. The evaluation methodology combines a sequential Bayesian method to guarantee consistency between independent and cumulative yields along with the physical constraints on the independent yields. This work was motivated to improve the performance of the ENDF/B-VII.1 library in the case of stable and long-lived cumulative yields due to the inconsistency of ENDF/B-VII.1 fission p;roduct yield and decay data sub-libraries. The revised fission product yields and the new covariance data are proposed as a revision to the fission yield data currently in ENDF/B-VII.1.

  5. Production and release of the fission gas in (Th U)O2 fuel rods

    International Nuclear Information System (INIS)

    The volume, composition and release of the fission gas products were caculated for (Th, U)O2 fuel rods. The theorectical calculations were compared with experimental results available on the literature. In ThO2 + 5% UO2 fuel rods it will be produced approximated 5% more fission gas as compared to UO2 fuel rods. The fission gas composition or Xe to Kr ratio has showed a decreasing fuel brunup dependence, in opposition to that of UO2. Under the same fuel rod operational conditions, the (Th, U)O2 fission gas release will be smaller as compared to UO2. This behaviour of (Th, U)O2 fuel comes from smallest gas atom difusivity and higher activation energies of the processes that increase the fission gas release. (Author)

  6. 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

  7. Experiments on the high-temperature behaviour of neutron-irradiated uranium dioxide and fission products

    International Nuclear Information System (INIS)

    Tanke, R.H.J. 1990 Experiments on the high temperature behaviour of neutron-irradiated uranium dioxide and fission products - Kema Scientific and Technical Reports 8(1): 1-79. ASSN 0167-8590. ISBN 90-353-0078-5. The objective of the study was to determine the release rate of fission products from overheated UO2 spheres of approx. 1 mm diameter, irradiated in the High-Flux Reactor at Petten, were used for the experiments. The chemical forms of the particles released from the spheres during evaporation were determined by mass spectrometry and the release rate of the fission products was determined by γ-spectrometry. A γ-tomographer was developed to determine the change with temperature in the 3-D distribution of radioactive fission products in the spheres. The experiments showed no clear relationship between the stoichiometry of the speres and uranium consumption, a finding which is contradictory to predictions made with a commonly used model. The absence of a clear relationship is explained by the - still not understood - influence of the microstructure of the nuclear fuel. A diffusion model was used to determine the activation energy for the diffusion of fission products. Evaporation of the UO2 matrix is the main mechanism for the release of all fission products at temperatures above 2300 K. Barium can be as volatile as iodine. Niobium and lanthanum can be volatile. Molecular combinations of the fission products, iodine, cesium and tellurium, are highly unlikely to be present inside the fuel. Barium and niobium may form compounds with oxygen and are then released as simple oxides. Fission products are released from overheated UO2 as atoms or as oxides. Finally, a new model is proposed for describing the behaviour of oxygen in irradiated nuclear fuel. (author). 87 refs.; 112 figs.; 13 tabs

  8. Fission product transport in the primary circuit and in the containment in severe nuclear accidents

    OpenAIRE

    Kalilainen, Jarmo

    2015-01-01

    The fission product transport in the primary circuit and in the containment was investigated in nuclear reactor severe accident conditions, where the reactor core is damaged and core materials are released from the reactor pressure vessel. The re-vaporization of fission products and core materials from the circuit surfaces was investigated with special emphasis on the effects of re-volatilization on iodine speciation and transport. Secondly, aerosol transport in the containment was studied in...

  9. Experimental Measurements and Computer Simulation of Fission Product Gamma-Ray Spectra

    OpenAIRE

    Sanderson, D.C.W; Cresswell, A.; Allyson, J.D.; McConville, P.

    1997-01-01

    Airborne gamma ray spectrometry using high volume scintillation detectors, optionally in conjunction with Ge detectors, has potential for making rapid environmental measurements in response to nuclear accidents. An experimental investigation and computer simulation have been used to characterise the response of such detectors to short lived fission products. Small samples of 235U were irradiated in a research reactor for short periods, to generate fission product sources. Gamma ray spectra w...

  10. Dose received after a single absorption of a fission product mixture (1962)

    International Nuclear Information System (INIS)

    During an accident a person can absorb a certain quantity of fission products. The doses received by the entire body or by a particular organ depend on the time spent in the reactor by the fuel element and the cooling time of the fission products. The calculations have been made for an irradiation time of 300 days and for a cooling time of between 1 day and 3 years. (author)

  11. Evaluation of fission product release rates during postulated core meltdown accidents

    International Nuclear Information System (INIS)

    An investigation of fission product release rates under severe accident conditions is presented for a pressurized water reactor with a large dry containment. In the calculation of fission product source terms, the numeical scheme by using core node temperatures during heatup and meltdown is introduced and incorporated into the MARCH code. The results indicated that the significant differences between the numerical integration method and WASH-1400 method are observed, especially Ru and La release categories. (Author)

  12. Performance of RELAP/SCDAPSIM Code on Fission Products Transport Prediction

    International Nuclear Information System (INIS)

    Fission product transport in the piping system of primary circuits is an important area of study in field of the severe accidents. Fission product transport comprises all phenomenon occurring from the nuclear core to the containment release site. Once released in the flow channels, fission products can condense on the piping walls, nucleate aerosols, which can agglomerate and/or deposit on the piping walls. The phenomenology occurs in a steam-hydrogen convective environment. A model (FPTRAN) was developed for the program RELAP/SCDAPSIM that calculates all phenomenon related to the fission product transport through the piping system. The model solves a set of differential equations. The coefficients in these equations represent the processes at which several states change among them. The processes considered were vapor adsorption and condensation on the piping walls, aerosol formation and growth (condensation and agglomeration), and aerosol deposition. The model also controls the aerosol particle size distribution. The PHEBUS experiments compose the most complete experimental program ever conducted for the understanding of fission product behavior in Reactor Cooling System and containment. It employs a reactor to generate fission products, which are transported through a scaled piping system simulating the primary circuit of a pressurized water reactor (PWR). Along the piping system, several instruments are installed to measure the amount of fission products deposited and their states. This paper describes the modeling of the experiment Phebus FPT-01 using RELAP/SCDAPSIM and compares simulation and experimental results to assess the performance of the FPTRAN module on the fission products transport prediction. These results can be considered satisfactory, except for iodine. This inconsistency of iodine is probably due to an incorrect chemical form assumed for iodine. (author)

  13. 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

  14. Analytical method and result of fission product release from core during a depressurization accident of HTTR

    International Nuclear Information System (INIS)

    This paper describes the analytical method and the result of the fission product release from the core during the depressurization accident of the HTTR (High Temperature Engineering Test Reactor) to be constructed by Japan Atomic Energy Research Institute. In the safety evaluation of the HTTR, a double-ended rupture of the coaxial pipe of the primary cooling system is postulated as the design basis depressurization accident. The thermal response of the core after a depressurization in the graphite-moderated gas-cooled reactor such as the HTTR is very slow because of the large heat capacity of the graphite. Therefore, the fuel temperature remains high during about hundred hours after the depressurization accident, which induces additional fission product release from the heated fuel by diffusion. Thus, the amount of the additional fission products released should be taken into account in the safety evaluation of the HTTR. In order to calculate the amount of the additional fission products released from the core during the depressurization accident, the HTCORE code has been developed. It determines the time-dependent amounts of fission product nuclides that escape from the core. The amount of the additional fission products released from the core into the containment vessel during the depressurization accident of the HTTR is calculated by HTCORE code to be 4.8 x 1013 MeV · Bq for noble gases, 5.5 x 1013 Bq for I-131 and 2.6 x 1012 Bq for Cs-137. (author). 7 refs, 10 figs

  15. 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

  16. 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

  17. Recent Experiments on the Beta Activity of Fission Products from the Thermal-Neutron Fission of U233, U235 and Pu239

    International Nuclear Information System (INIS)

    The build-up of beta-activity of fission products from the low-energy fission of U233 and U235 has been measured as a function of the lime t after starting the irradiation of the fissionable material with thermal neutrons. By differentiating the beta-activity divided by the constant fission product rate one obtains the mean beta-decay rate β(t) per fission as a function of the time t after fission. The uranium targets were mounted within a 4π proportional counter, coincident conversion electrons and γ-rays being thereby eliminated. The background of fast neutrons and γ-rays was strongly reduced by using a neutron beam tube with a liquid nitrogen-cooled bismuth single crystal filter. The fission product rate was measured with the same counter. The measurements cover the time interval from 0.01 s to 10 h. The results are compared with other experimental data and existing theories. The total number of beta-decays per fission was found to be 5.25 ± 0.2 and 5.93 ± 0.2 for U233 and U235 respectively compared with theoretical values of 5.27 and 6.10 respectively. (author)

  18. Diffusion of Zr, Ru, Ce, Y, La, Sr and Ba fission products in UO2

    International Nuclear Information System (INIS)

    The diffusivity of the solid fission products (FP) Zr (Zr4+), Ru (Ru4+, Ru3+), Ce (Ce4+), Y (Y3+), La (La3+), Sr (Sr2+) and Ba (Ba2+) 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 Y3+, 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 Ru3+ and Ru4+ diffusing even faster than Xe for T < 1200 K. The diffusivities correlate with the observed fission product solubility in UO2, and the tendency to form metallic and oxide second phase inclusions

  19. Nuclear criticality safety basics for personnel working with nuclear fissionable materials. Phase I

    International Nuclear Information System (INIS)

    DOE order 5480.1A, Chapter V, ''Safety of Nuclear Facilities,'' establishes safety procedures and requirements for DOE nuclear facilities. The ''Nuclear Criticality Safety Basic Program - Phase I'' is documented in this report. The revised program has been developed to clearly illustrate the concept of nuclear safety and to help the individual employee incorporate safe behavior in his daily work performance. Because of this, the subject of safety has been approached through its three fundamentals: scientific basis, engineering criteria, and administrative controls. Only basics of these three elements were presented. 5 refs

  20. Analysis of fission product behavior in the Saclay Spitfire Loop Test SSL-1. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, D.D.; Haire, M.J.; Ballagny, A.

    1978-02-01

    The behavior of the fission metal cesium and the fission gases krypton and xenon in the Saclay Spitfire Loop SSL-1 test has been compared to that predicted using General Atomic reference data and computer code models. This is the first in a series of analyses planned in order to provide quantitative validation of HTGR fission product design methods. In this analysis, the first attempt to rigorously verify fission product design methods, the FIPERQ code was used to model the diffusion of cesium graphite and release to the coolant stream. The comparisons showed that the cesium profile shape in the graphite web and the partition coefficient between fuel rod matrix material and fuel element graphite were correctly modeled, although the overall release was significantly underpredicted. Uncertainties in the source term (fissile particle failure fraction) and total release to the coolant precluded an accurate appraisal of the validity of FIPERQ. However, several recommendations are presented to improve the applicability of future in-pile test data for the validation of fission metal release codes. The half-life dependence of fission gas release during irradiation was found to be in good agreement with the model used in the reference design materials, providing assurance that this aspect of the fission gas release predictions is properly modeled.

  1. 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

  2. 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.

  3. Calculation of prompt fission product average cross sections for neutron-induced fission of 235U and 239Pu

    International Nuclear Information System (INIS)

    Yield-weighted average cross sections of neutron radiative capture, (n,2n), and (n,3n) reactions over prompt fission products (FPs) from 235U and 239Pu are calculated. The prompt fission production yields are taken from the ENDF/B-VII.0 library. The FPs for each fissile material exist over a range of approximately 1000 neutron-rich nuclides. Several nuclear reaction codes are utilized for calculating the cross sections on each individual fission product - EMPIRE-2.19, TALYS-1.0, GNASH, and CoH. The influence of the FP isomers on the average cross sections is examined with TALYS. We investigate the dependence of the average cross sections on the number of FPs taken for averaging. It is shown that the average capture cross section is much more sensitive to the number of FPs included, compared with the (n,2n) and (n,3n) reactions. An intercomparison of the calculated cross sections with the different reaction codes is carried out. In the capture reaction, EMPIRE predicted lower cross section than TALYS and CoH owing to different default assumptions used in the γ-ray strength function modeling. Moreover, the pre-equilibrium models implemented in each code give different predictions for the neutron-emission reactions, although the differences are relatively small. We also discuss a difference between the macroscopic and microscopic calculation options in TALYS for the pre-equilibrium model, optical potential model, and γ-ray strength function. The predictive capability of the reaction codes for the capture reaction is examined by comparing their calculations with the ENDF data, which are based on measurements. Compared with the historic Foster and Arthur's evaluation, our new (n,2n) predictions are similar, although our capture predictions are almost an order of magnitude higher. Recommended cross sections for use in applications have been tabulated in ENDF-formatted files. (author)

  4. 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)

  5. Isotopic fission product release from nuclear fuel under severe core damage accident conditions

    International Nuclear Information System (INIS)

    Isotopic fission gas release behavior during SFD tests 1-1, 1-3, and 1-4 is strongly dependent on the pre-test fuel history. For SFD 1-1, where the majority of all the fission products were generated during the preconditioning period, very little difference in isotopic release behavior between short- and long-lived species is predicted. For the SFD 1-3 and 1-4 tests, where the majority of all short-lived fission gases decayed away during the 4-year cooling period, differences between the behavior of long- and short-lived species are predicted. Most of the intragranular fission product release has been shown to be due to a grain-growth/grain-boundary-sweeping mechanism. In addition, fuel liquefaction/dissolution processes can lead to increased release under these degraded-core-accident conditions. These predictions follow the trend of the observed phenomena

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

    International Nuclear Information System (INIS)

    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

  7. 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)

  8. Thermal release of volatile fission products from irradiated nuclear fuel

    International Nuclear Information System (INIS)

    An effective procedure for removing 3H, Xe and Kr from irradiated fuels was demonstrated using Shippingport UO2 fuel. The release characteristics of 3H, Kr, Xe, and I from irradiated nuclear fuel have been determined as a function of temperature and gaseous environment. Vacuum outgassing and a flowing gas stream have been used to vary the gaseous environment. Vacuum outgassing released about 99% of the 3H and 20% of both Kr and Xe within a 3 h at 15000C. Similar results were obtained using a carrier gas of He containing 6% H2. However, a carrier gas containing only He resulted in the release of approximately 80% of the 3H and 99% of both Kr and Xe. These results indicate that the release of these volatile fission products from irradiated nuclear fuel is a function of the chemical composition of the gaseous environment. The rate of tritium release increased with increasing temperature (1100 to 15000C) and with the addition of hydrogen to the gas stream. Using crushed UO2 fuel without cladding and He as the carrier gas, Kr was completely released at 15000C in 2.5 h. Below 13500C, no Kr-Xe release was observed. Approximately 86% of the 129I and 95% of the cesium was released from a piece (3.9 g) of UO2 fuel at 15000C in He. The zirconium cladding was observed to fracture during heat treatment. A large-scale thermal outgassing system was conceptually designed by the General Atomic Company from an engineering analysis of available experimental data. The direct cost of a 0.5 metric/ton day thermal outgassing system is estimated to be $1,926,000 (1982 dollars), including equipment, installation, instrumentation and controls, piping, and services. The thermal outgassing process was determined to be a technically feasible and cost-competitive process to remove tritium in the head-end portion of a LWR fuel reprocessing plant. Additional laboratory-scale development has been recommended

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

    OpenAIRE

    Asner, D. M.; Burns, K; Campbell, L. W.; Greenfield, B.; Kos, M. S.; Orrell, J. L.; Schram, M.; VanDevender, B.; Wood, 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 measurem...

  10. Experimental determination of the antineutrino spectrum of the fission products of 238U

    International Nuclear Information System (INIS)

    Fission of 238U contributes about 10 % to the antineutrino emission of a pressurized water reactor. In the present thesis, the beta spectrum of the fission products of 238U 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.

  11. Research on behaviour of the irradiated uranium silicide for fission Mo-99 production

    International Nuclear Information System (INIS)

    This paper shows the results of purification of Mo-99 obtained by U-235 fission contained in uranium silicide (Si2U3) targets. These are the first tests carried out with irradiated targets. The separation method was previously developed employing non-irradiated uranium silicide with and with tracer addition. These tests were made trying to preserve the stages of the present method for fission Mo-99 production in the Argentine Republic. (author)

  12. 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.

  13. Considerations on the influence of fission products in whole core accidents

    International Nuclear Information System (INIS)

    If the hypothetical Whole Core Accidents which are taken into account in reactor safety analysis can change from one country to another, there is nevertheless a general agreement over their description and main phases. Furthermore the important parameters have also been identified by every laboratory. During the development of such core accidents the role of the fission products in essential. It is not the purpose of this paper to give an exhaustive description of the phases which can be influenced by the fission products, we will try however to focus this study on the most important ones. In a second step we will discuss the equation of state of irradiated fuels; here again one principal preoccupation being to quantify the influence of fission products on reactor accidents. It is not our purpose to enter on the fundamental aspects of the equation of state. The studies and the experimental program launched at the CEA will then be described. Special attention will be directed towards the eventual role of fission products in molten fuel-coolant interactions (MFCls) or the events leading to the initiation of whole core accidents. This paper will be limited to oxide fuels. Whether the whole core accident is initiated by a reactivity defect or a coolant coast-down, one has to deal with four great categories of phenomena. Loss of flow: the power is around the nominal value, while the coolant flow has been reduced by a factor of 5 to 10. This induces boiling and clad weakening. Will the plenum pressure lead to a clad rupture? In case of a rupture, what will be the effect on the voiding of the channel? Transient over power: influence of gases from gaseous and volatile fission products on the fuel movements? MFCIs: Influence of the fission products in the mode of contact between fuel and coolant? Influence on the fuel characteristics. Sodium vapour bubble expansion: influence of the fission products on the heat transfer and eventual condensation of the bubble?

  14. 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

  15. Transmutation of fission product 93 Zr by fusion neutron source

    International Nuclear Information System (INIS)

    The problem of 93 Zr accumulation and its transmutation is addressed. It is shown that neutron flux required for 93 Zr transmutation is unachievable in fission reactor. Therefore transmutation of 93 Zr in the fusion based facility is proposed. It is demonstrated that current ITER fusion reactor project can be sufficiently served as a transmuter to resolve the problem of 93 Zr accumulation within the time period of several decades

  16. 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)

  17. 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)

  18. 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…

  19. 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

  20. Implementation of a Gibbs energy minimizer in a fission-product release computer program

    International Nuclear Information System (INIS)

    SOURCE 2.0 is the Canadian computer program for calculating fractional release of fission products from the UO2 fuel matrix. In nuclear accidents, fission-product release from fuel is one of the physical steps required before radiation dose from fission products can affect the public. Fission-product release calculations are a step in the analysis path to calculating dose consequences to the public from postulated nuclear accidents. SOURCE 2.0 contains a 1997 model of fission-product vaporization by B.J. Corse et al. based on lookup tables generated with the FACT computer program. That model was tractable on computers of that day. However, the understanding of fuel thermochemistry has advanced since that time. Additionally, computational resources have significantly improved since the time of the development of the Corse model and now allow incorporation of the more-rigorous thermodynamic treatment. Combining the newer Royal Military College of Canada (RMC) thermodynamic model of irradiated uranium dioxide fuel, a new model for fission-product vaporization from the fuel surface, a commercial user-callable thermodynamics subroutine library (ChemApp), an updated nuclide list, and updated nuclear physics data, a prototype computer program based on SOURCE IST 2.0P11 has been created that performs thermodynamic calculations internally. The resulting prototype code (with updated and revised data) provides estimates of 140La releases that are in better agreement with experiments than the original code version and data. The improvement can be quantified by a reduction in the mean difference between experimental and calculated release fractions from 0.70 to 0.07. 140La is taken to be representative of “low-volatile” fission products. To ensure that the existing acceptable performance for noble gases and volatile fission products is not adversely affected by the changes, comparisons were also made for a representative noble gas, 85Kr, and a representative volatile

  1. Simulation of COMEDIE Fission Product Plateout Experiment Using GAMMA-FP

    International Nuclear Information System (INIS)

    FThis phenomenon is particularly important under a VHTR design with vented low pressure confinement (VLPC), because the vent allows the prompt release of fission products accumulated within the primary circuit to environment during an initial blow-down phase after pipe break accidents. In order to analyze the fission product plateout, an numerical model was developed by Yoo et al. and incorporated into the GAMMA-FP code in the past. The GAMMA-FP model was validated against two experiment data, i.e., VAMPYR-1 and OGL, during the development phase. One of the well-known experiments for fission product plateout is the COMEDIE experiment. In this work, the COMEDIE experiment has been simulated using the GAMMA-FP code to investigate the reliability and applicability of the plateout model of GAMMA-FP. The COMEDIE experiment for fission product plateout was simulated using the GAMMA-FP code in this work. A good agreement was achieved between the measured and predicted plateout activities. The existing solution scheme was modified to allow larger time step size for fission product analysis in order to speed-up the computational time. Nevertheless, the modification of the existing numerical model of GAMMA-FP is necessary when a simulation capability of a long duration of plateout period (e.g., 60 years) is targeted

  2. 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

  3. Highlights from the IAEA coordinated research programme on fuel performance and fission product data

    International Nuclear Information System (INIS)

    Seven countries are cooperating with the objectives (i) to document the status of the experimental data base and of the predictive methods for Gas-Cooled Reactor fuel performance and fission product behaviour; (ii) to verify and validate methods in fuel performance and fission product retention prediction. These countries are China, France, Germany, Japan, Russia, USA and the UK. Duration of the programme is 1993-96. The technology areas addressed in this IAEA Coordinated Research Programme are: Fuel design and manufacture, Normal operation fuel performance and fission product behaviour, Accident condition fuel performance and fission product behaviour, -core heatup, -fast transients, -oxidising conditions (water and air ingress), Plateout, re-entrainment of plateout, fission product behaviour in the reactor building, and Performance of advanced fuels. Work performed so far has generated a 300-page draft document with important information for normal operations (Germany, Japan, China, Russia) and accident conditions (USA, Japan, Germany, Russia) and, additionally, a special chapter on advanced fuels (Japan). (author)

  4. Prediction of fission product and aerosol behaviour during a postulated severe accident in a LWR

    International Nuclear Information System (INIS)

    Lack of appropriate energy removal causes fuel elements in a reactor core to overheat and may eventually cause core to degrade. Fission products will be emitted from a degraded reactor core. Aerosols are generated when the vapours of various fuel and structural materials reach a cold environment and nucleate. In addition to the fission products release and aerosol generation taking place in the reactor vessel, some more fission products release and aerosol generation will occur when the molten core debris leaves the pressure vessel bottom head and comes in contact with the pedestal concrete floor. Fission products, if they are released to environment from the containment boundary, exert a great danger to public health. A source term is defined as the quantity, timing, and characteristics of the release of radionuclide material to the environment following a postulated severe accident. At PSI a considerable effort hase been spent in investigating and establishing a source term assessment methodology in order to predict the source term for a given Light Water Reactor (LWR) accident scenario. This report introduces the computer programs and the methods associated with the release of the fission products, generation of the aerosols and behaviour of the aerosols in LWR compartments used for a source term assessment analysis at PSI. (author) 4 figs., 5 tabs., 28 refs

  5. Analysis on Fission Product Inventory with the Consideration of Axial Power Distribution

    International Nuclear Information System (INIS)

    Most of the regulatory guides have proposed taking the assumptions that can lead to more severe results than those that are realistically expected. In the case of a Loss of Coolant Accident (LOCA), the fission product inventory in the core is generally estimated based on the average burnup of a fuel assembly, regardless of the accident occurrence time and degree of fuel meltdown. However, the actual fuel burnup is non-uniform along the core height with axial power distributions, and fluctuated with fuel cycles as shown in. In this study, the variation of fission product inventory in the core is analyzed by considering the degree of fuel meltdown at the Beginning of Cycle (BOC), the Middle of Cycle (MOC), and the End of Cycle (EOC). For this study, the ORIGEN-ARP in the SCALE 6.1 package code system and the axial power distributions of the Ulchin (Hanul) unit 6 were used for the depletion calculation. The variation of fission product inventory in the core was analyzed based upon the degree of fuel meltdown, which reflects the axial power distribution of the fuel at BOC, MOC, and EOC. When considering some major fission products released into the containment, the calculation with the assumptions recommended by regulatory guides does not lead to results more conservative than for over approximately a 30% meltdown. Therefore, the fission products released into the containment need to be evaluated with a consideration of the accident occurrence point and the degree of fuel meltdown

  6. Fission product recycling as catalysts for hydrogen production by water electrolysis

    International Nuclear Information System (INIS)

    An electrolytic extraction method has been studied to separate fission products (Ru, Rh, Pd, Tc, Se, Te, etc) from the nuclear spent fuel. Yet they are rare metal fission products (RMFP), most are long-lived (LLFP; Pd, Tc, Se, Te). In the applied separation process, Pd2+ cation itself would not only be easily deposited from various nitric acid solutions, but enhance also the deposition of RuNO3+ and ReO4- by acting as a catalyst (as Pdadatom). Such Catalytic Electrolytic Extraction (CEE) method was found to be applicable in the case of TcO4- deposition, too. The quaternary-, Pd-Ru-Rh-Re, deposit on the Pt electrodes show the highest cathodic current, ca twice superior to that of the Pt electrode both in artificial sea water as well as in alkaline solution. The promising utilization of RMFP will be as 'FP-catalyst' for hydrogen production by water electrolysis. RMFP would be circulating material to bridge nuclear and hydrogen energy systems. (author)

  7. Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (235U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required 99Mo/99mTc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99

  8. Process for packaging fission products resulting from reprocessing by dry methods

    International Nuclear Information System (INIS)

    A process is described for packaging fission products obtained by the dry method with a view to making a resulting product that offers great innocuousness with respect to the environment, the characteristic being that a mix is made comprising a major proportion of a zeolite of formula nSiO2, mAl2O3, MO, zH2O and a minor proportion of processed primary fission products; this mix is brought to its melting temperature at an appropriate temperature rise rate, this temperature being maintained until a homogeneous phase is obtained, after which it is allowed to cool

  9. 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.

  10. 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.

  11. 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

  12. Determination of fission product yields in the 14 MeV photon (Bremsstrahlung) induced fission of 232Th

    International Nuclear Information System (INIS)

    The cumulative yields of various fission products in the 232Th(γ,f) with end-point Bremsstrahlung energy of 14 MeV having have been determined using off-line γ-ray spectrometric technique. The end-point Bremsstrahlung energy of 14 MeV was generated by impinging the electron beam on a solid graphite beam dump of the 20 electron LINAC (ELBE) at Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Dresden, Germany. From the cumulative fission yields, the mass chain yields were obtained by using charge distribution correction of medium energy. The fine structure in the mass yield distribution was interpreted from the point of nuclear structure effect such as shell closure proximity and even-odd effect. The mass yield distribution in 232Th(γ,f) is triple humped unlike 238U(γ,f), where it is double humped. This different behaviour in between 232Th* and 238U* was explained from the point of different potential energy surfaces between two systems. (author)

  13. Selective Determination of Gaseous Fission Products in a Radioactive Decay Chain

    International Nuclear Information System (INIS)

    We have developed equipment for the selective determination of gaseous fission products in a radioactive decay chain. A circulating pump is used to pass the gases being analysed through equipment consisting of an input filter for removing solid particles, a sealed enclosure, referred to as the disintegration chamber, in which some of the radioactive gases disintegrate, and a filter at the outlet of the enclosure for retaining the daughter products produced in the disintegration chamber. After suction, the activity of the daughter products collected by the filter is analysed. We have studied the families of seven fission gases whose half-lives range between 2.8 hours and 10 secs: 88Kr, 138Xe, 89Kr, 139Xe, 90Kr, 140Xe, 91Kr. 88Kr, 138Xe and 89Kr are measured by cycling without a source; this procedure involves three operations: the fission gases are formed in the generator; after the production of fission gases ceases, the gases with the shortest half-lives are eliminated during a delay time; the air containing the remaining fission gases is recycled in the equipment. By altering the irradiation, delay and cycling times, one improves the relative concentration of the gas daughter product one wants to determine. The gases with the shortest half-lives, i.e. 139Xe, 90Kr, 140Xe and 91Kr, are measured in a single pass: as the fission gases are produced, they are passed through the installation and vented to the surroundings. In this case, the suction time and the delay time after suction are varied to improve the relative concentration of the gas daughter product being measured. The activity of the outlet filter is measured by gamma spectrometry with subtraction of the Compton effect. It is thus possible to separate the isotopes of solid elements which are daughter products of the different gases. (author)

  14. 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)

  15. Method of separation of fission and corrosion products and of corresponding isotopes from liquid waste

    International Nuclear Information System (INIS)

    A method of separating fission and corrosion products and corresponding stable isotopes from liquid waste is described. Mycelia of fungi are used as sorbents for retaining these products on their surface and within their pores. Methods of activation or regeneration of the sorbent are outlined. 11 claims

  16. Rare metal fission products in nuclear spent fuel as catalysts for hydrogen production by water electrolysis

    International Nuclear Information System (INIS)

    Separation and utilization of rare metal fission products (RMFP) in nuclear spent fuel were studied to apply them as a catalyst for hydrogen generation by water electrolysis. The RMFP, namely Pd, Ru, Rh and Tc, etc, are abundant, more than ca. 30kg per metric ton of a typical fast reactor spent fuel. The RMFP can be selectively separated from high level liquid waste (HLLW) by catalytic electrolytic extraction (CEE) method. Specific metallic cations such as Pd2+, which originate in the solutions, may act as promoters (i.e., Pdadatom) or mediators, thereby accelerating electrochemical deposition of RuNO3+, Rh3+ and ReO4- (simulator TcO4-). In utilizing CEE method, electrodeposited electrodes were prepared, and successively dedicated to the water (alkaline or artificial sea water) electrolysis tests. Among the RMFP deposited electrodes, maximum potential shifting for hydrogen evolution to noble side was observed for the quaternary, Pd-Ru-Rh-Re (3.5:4:1:1), deposit Pt electrode, with suggesting the highest cathodic currents for hydrogen evolution both in alkaline solution and artificial sea water. The electro analytic activity of quaternary, Pd-Ru-Rh-Re (3.5:4:1:1), deposit Pt electrode exceeded that of Pt electrode by ca. twice both in alkaline solution and artificial sea water. The paper conclusively proposes RMFP generated by nuclear fission to utilize as an alternative material for hydrogen production with a novel vision to bridge nuclear and hydrogen energy systems. (author)

  17. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    The transport and fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behaviour of simulant fission product species such as caesium iodide, caesium hydroxide and tellurium, in terms of their vapour deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high-density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO/sub 2/ clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 25000C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapour phase, and specific data using this technique are reported

  18. Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

    International Nuclear Information System (INIS)

    An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

  19. Simulated fission product oxide behavior in Triso-coated HTGR fuel

    International Nuclear Information System (INIS)

    Several combinations of Trisco-coated UO2 particles with additions of simulated fission product oxides were investigated. They were first heat-treated in the laboratory; then their behavior was examined by metallography, radiography, the scanning electron microscope, and electron microprobe x-ray analysis. Pressures of the various gaseous species within the particles were calculated and displayed as Ellingham diagrams. It appears to be essential that Triso-coated fuel have impermeable inner high-density pyrocarbon (iLTI) layers, because the fission product strontium, in gaseous form, will interact with SiC. As oxides, the rare earth fission products redistributed slightly within the buffer layer but did not interact with the SiC layers

  20. 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.

  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. Augmentation of ENDF/B fission product gamma-ray spectra by calculated spectra

    International Nuclear Information System (INIS)

    Gamma-ray spectral data of the ENDF/B-V fission product decay data file have been augmented by calculated spectra. The calculations were performed with a model using beta strength functions and cascade gamma-ray transitions. The calculated spectra were applied to individual fission product nuclides. Comparisons with several hundred measured aggregate gamma spectra after fission were performed to confirm the applicability of the calculated spectra. The augmentation was extended to a preliminary ENDF/B-VI file, and to beta spectra. Appendix C provides information on the total decay energies for individual products and some comparisons of measured and aggregate values based on the preliminary ENDF/B-VI files. 15 refs., 411 figs

  3. 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.

  4. Fission product migration in intact fuel rods. S176 experiments 1-5: Fission product deposits on clad specimens and their thermal stability

    International Nuclear Information System (INIS)

    This report is the fifth of a group which present the results obtained during the first 5 experiments in the S176 series of irradiation experiments. The aim of this series is to provide informtion on the distribution of fission products in intact irradiated fuel rods, both within the UO2 fuel and on the inside of the Zircaloy clad. Fuel rods, previously irradiated to appreciable burnups in the Aagesta R3 reactor, after cooling, are re-irradiated in the Studsvik R2 test reactor for short periods to build up significant inventories of short-lived fission products of interest. Examination of sections of fuel and clad is performed within a short time after removal from the reactor

  5. Fission product migration in intact fuel rods. S176 experiments 1-5: Gamma spectrometric and beta autoradiography results on radial fission product distribution

    International Nuclear Information System (INIS)

    This report is the third of a group which present the results obtained during the first 5 experiments in the S176 series of irradiation experiments. The aim of this series is to provide information on the distribution of fission products in intact irradiated fuel rods, both within the UO2, fuel and on the inside of the Zircaloy clad. Fuel rods, previously irradiated to appreciable burnups in the Aagesta R3 reactor, after cooling, are reirradiated in the Studsvik R2 test reactor for short periods to build up significant inventories of short-lived fission products of interest. The irradiation conditions are well defined. Examination of sections of fuel and clad within a short time after removal from the reactor is performed by means of high-resolution Ge(Li) gamma spectrometry. The fuel sections are also examined metallographically to determine fuel structure, and by β autoradiography

  6. 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)

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

    International Nuclear Information System (INIS)

    Three experiments (one from Oak Ridge and two from Los Alamos), in which samples of 235U and 238Pu 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. 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.

  9. Mechanical behaviour and fission product release in overpower transients at entended burnup

    International Nuclear Information System (INIS)

    An important factor for the utilization of water reactor fuel at extended burnup is the PCI behaviour during operational transients. The data from the internationally sponsored RISOE Fission Gas Project have provided significant information regarding the fission product release and swelling of UO2-Zr fuel pins with burnup in the range of 38,000-44,000 MWd/tU, peak pellet, during mild overpower transients (bump tests). The pellet cladding mechanical interaction during the transient tests was large enough to cause permanent deformation down to local heat ratings around 300 W/cm. The mechanical interaction was promoted by very rapid gaseous swelling, which developed during the transients. The local swelling increase during the short transient tests was measured by QIA and found to be as high as 8% at local, calculated temperatures of 800 deg. C. The local release of the volatile fission products Cs and I was measured by radial gamma scanning and found to be very similar to the measured release of noble fission gas. Radial EMPA measurements showed that the Xe release started at local, calculated temperatures of about 700 deg. C. Above 1100 deg. C a release of about 95% was found. Although significant mechanical interaction as well as release of volatile fission products were observed, only one of the twelve tests resulted in cladding failure. This is attributed to the slow approach to the full test power levels, where a power increase rate of 5 W/cm/h (peak pellet) was generally used. (author)

  10. Freedom: a transient fission-product release model for radioactive and stable species

    International Nuclear Information System (INIS)

    A microstructure-dependent fission-gas release and swelling model (FREEDOM) has been developed for UO2 fuel. The model describes the transient release behaviour for both the radioactive and stable fission-product species. The model can be applied over the full range of operating conditions, as well as for accident conditions that result in high fuel temperatures. The model accounts for lattice diffusion and grain-boundary sweeping of fusion products to the grain boundaries, where the fission gases accumulate in grain-face bubbles as a result of vacancy diffusion. Release of fission-gas to the free void of the fuel element occurs through the interlinkage of bubbles and cracks on the grain boundaries. This treatment also accounts for radioactive chain decay and neutron-induced transmutation effects. These phenomena are described by mass balance equations which are numerically solved using a moving-boundary, finite-element method with mesh refinement. The effects of grain-face bubbles on fuel swelling and fuel thermal conductivity are included in the ELESIM fuel performance code. FREEDOM has an accuracy of better than 1% when assessed against an analytic solution for diffusional release. The code is being evaluated against a fuel performance database for stable gas release, and against sweep-gas and in-cell fission-product release experiments at Chalk River for active species

  11. Nuclear charge distribution of fission products originated from fission of 238U nuclei induced by45-69 MeV protons

    Directory of Open Access Journals (Sweden)

    Houshyar Noshad

    2007-12-01

    Full Text Available  Fission of 238U nuclei was performed by 45-69 MeV protons at the Cyclotron and Radioisotope Center of Tohoku University in Japan. The fission products originated in the reaction were identified by using gamma spectroscopy. The experimental data show that the charge distribution of isobar fission products follows a Gaussian distribution with a standard deviation independent of the selected mass number. The standard deviations were measured for the reaction 238U(p, f with 45, 55, 65 and 69 MeV protons. For Ep = 45 MeV, the standard deviation obtained from the experiment is in agreement with the existing data and satisfies the prediction of the Hauser-Feshbach statistical model. For other proton energies, measurement of this quantity has not been reported in the literature. The experimental results show that the value of standard deviation increases, when the excitation energy of the fissioning nucleus increases. Furthermore, the most probable charge was determined for the isobar fission products detected in the experiment. The results are consistent with the prediction of the minimum potential energy (MPE model. Moreover, the experimental data show that nuclear charge polarization occurs in the fission process.

  12. Sensitivity analysis of the effect of various key parameters on fission product concentration (mass number 120 to 126)

    International Nuclear Information System (INIS)

    An analytical sensitivity analysis has been made of the effect of various parameters on the evaluation of fission product concentration. Such parameters include cross sections, decay constants, branching ratios, fission yields, flux and time. The formulae are applied to isotopes of the Tin, Antimony and Tellurium series. The agreement between analytically obtained data and that derived from a computer evaluated model is good, suggesting that the analytical representation includes all the important parameters useful to the evaluation of the fission product concentrations

  13. Transmutation of fission products in reactors and accelerator-driven systems

    International Nuclear Information System (INIS)

    Energy flows and mass flows in several scenarios are considered. Economical and safety aspects of the transmutation scenarios are compared. It is difficult to find a sound motivation for the transmutation of fission products with accelerator-driven systems. If there would be any hesitation in transmuting fission products in nuclear reactors, there would be an even stronger hesitation to use accelerator-driven systems, mainly because of their lower energy efficiency and their poor cost effectiveness. The use of accelerator-driven systems could become a 'meaningful' option only if nuclear energy would be banished completely. (orig./HP)

  14. Determination of 235U isotope abundance by measuring selected pairs of fission products

    International Nuclear Information System (INIS)

    A new method to determine experimentally the isotopic abundance of uranium has been developed by measuring the ratios of selected fission product pairs. The relationship between the 235U abundance and the ratios of 14 MeV neutron-induced fission products, which are N(3892Sr)/N(44105Ru) and N(3892Sr)/N(53135I), have been obtained. The results of 4 test samples show that the relative deviations between the measured and the reference 235U abundances are less than ±3%.

  15. Benchmarking a fission-product release computer program containing a Gibbs energy minimizer

    International Nuclear Information System (INIS)

    The computer program SOURCE IST 2.0 contains a 1997 model of fission-product vaporization, developed by B.J. Corse et al. That model was tractable on computers of that day. However, the understanding of fuel thermochemistry has advanced since that time. A new prototype computer program was developed with: a) newer Royal Military College of Canada thermodynamic model of uranium dioxide fuel, b) new model for fission-product vaporization from the fuel surface, c) a user-callable thermodynamics subroutine library, d) an updated nuclear data library, and e) an updated nuclide generation and depletion algorithm. The prototype has been benchmarked against experimental results. (author)

  16. Modelling of fission product release from TRISO fuel during accident conditions : benchmark code comparison / Ramlakan A.

    OpenAIRE

    Ramlakan, Alastair Justin

    2011-01-01

    This document gives an overview of the proposed MSc study. The main goal of the study is to model the cases listed in the code benchmark study of the International Atomic Energy Agency CRP–6 fuel performance study (Verfondern & Lee, 2005). The platform that will be employed is the GETTER code (Keshaw & van der Merwe, 2006). GETTER was used at PBMR for the release calculations of metallic and some non–metallic long–lived fission products. GETTER calculates the transport of fission products ...

  17. Fission product migration in intact fuel rods S176 experiments 1-5: Metallography results

    International Nuclear Information System (INIS)

    This work is part of a project, the aim of which is to provide information on the distribution of fission products in intact irradiated fuel rods, both within the UO2 fuel and on the inside of the zircaloy clad. Fuel rods, previously irradiated to appreciable burnups in the Aagesta R3 reactor, after cooling, are re-irradiated in the Studsvik R2 test reactor for short periods to build up significant inventories of short-lived fission products of interest. The irradiation conditions are well defined. In this report the results from the ceramographic analyses of the fuel rods are presented together with the beta autoradiography results

  18. Calculated leaching of certain fission products from a cylinder of French glass

    International Nuclear Information System (INIS)

    The probable total leaching of the most important fission products and actinides have been tabulated for a cylinder of French HLW glass with approximately 9 percent fission products. The calculations cover the period between 30 and 10000 years after removal from the reactor. The cylinder is of the type planned for the introduction of the HLW into Swedish crystalline rocks. All the components are supposed to have the same leach rate. The calculations also include the probable thickness of eroded glass layer/year. (author)

  19. 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)

  20. 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.

  1. Direct and preequilibrium effects in the fission-product mass range

    International Nuclear Information System (INIS)

    Until recently inelastic scattering did not gain the proper attention in fission-product cross section evaluations. In many existing evaluations global spherical optical models have been used, neglecting direct and pre-equilibrium effects. There are also few experimental data relevant to inelastic scattering in fission products. This paper is focussed on the anomalously high inelastic scattering cross sections observed in even-mass nuclei near mass A=100 at low energies. Both more data and more refined theoretical analyses are required. A number of suggestions for relevant coupled-channel calculations is made. (author). 29 refs., 5 figs., 1 tab

  2. 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\

  3. 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

  4. 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

  5. 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.)

  6. 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)

  7. Evaluation of fission product afterheat. Quarterly report, 1 October 1976--31 December 1976

    International Nuclear Information System (INIS)

    Reported here is work for the period indicated under the subject contract. Specific tasks covered in the report are: (1) a re-examination of fission product yield uncertainties to be used as input to merging sensitivity calculations, (2) a tabulation of recommended uncertainties on fission product decay energies to accompany the decay energy of ENDF/B-IV, (3) a description of the ROPEY computing system, and (4) a reduction of ORNL decay heat data to pulse irradiation conditions and the merging into it of reliable older data of reasonable quality. The Oak Ridge experiment dominates the final results

  8. Modelling of whole-core release of fission products in PWR core melt accidents: Chapter 13

    International Nuclear Information System (INIS)

    The computer code FISREL combines the thermal history of a reactor core with experimentally-based release rate constants to calculate whole-core release histories of fission products in PWR core melt accidents. Predictions of the code for releases of volatile fission products during large-break, small-break and transient initiated sequences are presented, and the sensitivities of results to input data examined. A preliminary assessment of the limitations imposed by mass transport on release of vaporized materials in high pressure sequences is given, and the implications of the results for primary system transport are discussed

  9. Fission product behaviour - in particular Cs-137 - in HTR-TRISO-coated particle fuel

    International Nuclear Information System (INIS)

    This work is performed between 1977 and 1979. The main task is to determine a temperature dependent diffusion coefficient of the fission product Cs-137 in the silicon carbide interlayer of HTR particles. The raw material is laso presented as the used measuring techniques and computer codes. The results are discussed in detail and some critical remarks are made about the efficiency of the silicon carbide interlayer to retent fission products including Ag-110m, Sr-90, and Ru-106, which temperature dependent diffusion coefficient is also been determined. (orig.)

  10. 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)

  11. Determination of fission and activation products in nearly fresh fuel elements by self-calibration

    International Nuclear Information System (INIS)

    Gamma spectra at several axial positions of the BR02 fuel element X133 have been measured and the concentration of 137Cs has been determined directly. On basis of this determination and some assumptions of the irradiation history, the concentrations of other fission products, relevant for the cold reprocessing process, have been calculated. On the basis of ORIGEN-type calculations the concentration of some relevant transuranics has been estimated. Both the concentrations of the fission products and of the transuranics appeared to be below the threshold values. Therefore it could be decided that cold reprocessing was feasible for all old BR02 fuel elements. (author)

  12. Experimental studies on removal of airborne fission products methyl iodide by sprays in containment

    International Nuclear Information System (INIS)

    For reducing the amount of fission products leaked to environment under accident conditions of PWR, the experimental studies on the removal of airborne fission products methyl iodide by sprays in containment was carried out on the basis of the theoretical work in a simulation facility. Inactive methyl iodide was used for the experiment so the experiment facility was simplified. A gas chromatography was employed to measure the aerosol concentration of methyl iodide. A series of experiments on the removal of methyl iodide by sprays under different temperatures and various chemical additives has been made. The experimental results are useful for rationally selecting parameters of containment spray system of PWR

  13. 235U fission product gamma spectra: a comparison between experiment and calculation

    International Nuclear Information System (INIS)

    Recent measurements of 235U fission product gamma spectra (ORNL), after reduction to a broad group structure, were compared with results of corresponding summation calculations which were made with the UKFPDD-1 fission product data base. In order to facilitate convenient integration of the summation results, weighted sums of decaying exponentials were accurately fitted to them using an iterative least-squares method described. The comparisons between experiment and calculation reveal significant deficiencies in data for short-lived nuclides which prevent the accurate calculation of short-term ( 5 MeV are also examined briefly and possible causes for them are discussed. (author)

  14. 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.

  15. 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

  16. The different facilities of the reactor PHENIX for radio isotope production and fission product burner

    International Nuclear Information System (INIS)

    During the last few years different tests have been made to optimize the blanket of the reactor. Year after year the breeding ratio has lost a part of interest regarding the production and availability of plutonium in the world. A characteristic of a fast reactor is to have important neutron leaks from the core. The spectrum of those neutrons is intermediate, the idea was to find a moderator compatible with sodium and stable in temperature. After different tests we kept as a moderator the calcium hydride and as a samply support, a cluster which is separated from the carrier. At the end we present the model used for thermalized calculations. The scheme is then applied to a heavy nuclide transmutation example (Np237 Pu238) and to fission product transmutation (Tc99). (author)

  17. 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)

  18. Design of an on-line selective ion source for fission products

    International Nuclear Information System (INIS)

    This project is part of a more vast project for producing radioactive ions starting from the reaction 238U(γ,f). A drawback of this type of production by fission is the difficulty of separating efficiently the simultaneously produced isobars, except for a few particular elements (rare gases, alkali metals, halogens). For alkali metals and halogens one makes use of the high sensitivity of their surface ionization. The rare gases can be easily separated from other elements due to their high volatility related to the lack of reactivity. The new design goes still farther in this direction extending the method to elements of medium or low volatility. The apparatus is inserted between the target-containing crucible where the nuclides under study are generated and the ionization chamber in which the beam is generated. The difference in volatility of the elements will be used by the ionization chamber in selecting the isobaric elements. The paper presents the principle and basic hypotheses on which a calculation was done to compare the evaporation of tin and antimony. The results are discussed. A practical solution of the project was found by using a model and Compromising between the numerous thermal, geometrical and temporal parameters and by making use of a computerized model, a practical solution of the project was found

  19. 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.

  20. Upgrading DRACULA setup to be used for light products - fission fragments coincidence measurements

    International Nuclear Information System (INIS)

    At low bombarding energy (E/A 238 U give rise to a number of fission processes, all leading to very similar fission products. Therefore, in order to understand the fission processes in this energy domain it is of interest to determine the amount of fission occurring after a peripheral interaction relative to that originating from compound nucleus formation. Although the detection of a projectile residue (PLF) in coincidence with the fission fragments is a very promising probe for the macroscopic features of the mechanism of induced fission, at incident energies in the vicinity of the Coulomb barrier (E/A 2 cross section area uses the phoswich technique by coupling a thin fast NE102A plastic scintillator to a 10 cm long BaF2 crystal of hexagonal section. The BaF2 crystal detectors have been successfully used in modular multielement detector ARGOS in the context of GANCT and HOTCT researches at LNS. The light response of the phoswich configuration as a function of the plastic thickness and of the energy and charge of the incident ion has been studied at Tandem energies. Both arrays will be placed in separate vacuum chambers attached to the remaining large angular opening windows of the reaction chamber. By rotating the whole device the fission fragment detection arrays will cover a range of 96 angle in the horizontal plane. The main advantage of this setup is that it allows to perform continuous measurements in energy and angle of the reaction products. The geometry of the whole device has been tested by Monte Carlo calculations using the code ELPHIC. The coincidence condition is completely fulfilled for the first two positions of the setup and partially for the third one. Measurements are intended to be performed at the SMP Tandem from LNS-Catania using light beams (16 O, 19 F, 20 Ne, 32 S) at ∼ 6 MeV/A on high fissility parameter targets. (authors)

  1. 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

  2. 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

  3. Short term fission product and actinide decay heat for a PWR

    International Nuclear Information System (INIS)

    This note gives the results of best estimate calculations of the decay heat following reactor trip for the UK PWR using UK recommended methods. It is intended that these values, together with the uncertainties identified, should be used for the analysis of reactor transients following shutdown. This requires the use of the computer code FISPIN (or a similar code FISP) together with the First UK Library of Fission Product Decay Data (UKPFDD-1), the Crouch 2 fission yields and group averaged fission product capture cross sections recommended individually for each reactor type. The calculations reported here conform to this standard. Decay heat from heavy elements (identified as actinides in this report) is also calculated in FISPIN. (U.K.)

  4. 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.))

  5. 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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. 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)

  10. The use of recoil for the separation of uranium fission products; Utilisation du recul pour la separation des produits de fission de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Henry, R.; Herczec, C. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The recoil distance of fission fragments in U{sub 3}O{sub 8} is about 8 microns. By using highly diluted suspensions of uranium oxide particles having dimension much smaller than this figure (mean diameter 0,5 micron), we were able to study the re-adsorption of fission products on uranium oxide. Separation results have been studied as a function of the nature of the irradiation medium (solid or liquid) and the separation medium, of particle size and of concentration of particles in the dispersing medium. Decay curves can be used to discriminate between {sup 239}Np and mixed fission products. Most of the {sup 239}Np is found in the U{sub 3}O{sub 8} particles. The location of fission products in solid dispersing media has been determined, fission products being found always inside the dispersing medium particles. The results obtained can be applied to the rapid separation of short-lived fission products from a uranium-free starting material. (author) [French] Le parcours de recul des fragments de fission est en moyenne de 8 microns dans l'U{sub 3}O{sub 8}. En prenant des suspensions d'oxyde d'uranium dont les particules, tres diluees, ont des dimensions nettement inferieures a cette valeur (diametre moyen 0,5 micron), on a pu etudier directement la readsorption des produits de fission sur l'oxyde d'uranium. Les resultats de separation ont ete etudies en fonction de la nature du milieu d'irradiation (solide ou liquide) et du milieu de separation, de la taille des particules d'oxyde et de leur concentration dans le milieu dispersant. Les courbes de decroissance permettent de determiner la perturbation apportee dans les mesures par le {sup 239}Np qui reste en majorite dans les grains d'U{sub 3}O{sub 8}. On a determine enfin l'emplacement des produits de fission dans le cas des melanges solides; ils se trouvent toujours a l'interieur des grains du milieu recepteur. Les resultats obtenus permettent d'envisager la

  11. Modeling of in-vessel fission product release including fuel morphology effects for severe accident analyses

    International Nuclear Information System (INIS)

    A new in-vessel fission product release model has been developed and implemented to perform best-estimate calculations of realistic source terms including fuel morphology effects. The proposed bulk mass transfer correlation determines the product of fission product release and equiaxed grain size as a function of the inverse fuel temperature. The model accounts for the fuel-cladding interaction over the temperature range between 770 K and 3000 K in the steam environment. A separate driver has been developed for the in-vessel thermal hydraulic and fission product behavior models that were developed by the Department of Energy for the Modular Accident Analysis Package (MAAP). Calculational results of these models have been compared to the results of the Power Burst Facility Severe Fuel Damage tests. The code predictions utilizing the mass transfer correlation agreed with the experimentally determined fractional release rates during the course of the heatup, power hold, and cooldown phases of the high temperature transients. Compared to such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation, the mass transfer correlation resulted in lower and less rapid releases in closer agreement with the on-line and grab sample data from the Severe Fuel Damage tests. The proposed mass transfer correlation can be applied for best-estimate calculations of fission products release from the UO2 fuel in both nominal and severe accident conditions. 15 refs., 10 figs., 2 tabs

  12. 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

  13. Atomic scale simulation of fission product trapping and migration in UO2+x and U3O8-z

    International Nuclear Information System (INIS)

    The paper discusses simulations, performed at atomic scale of fission product trapping and migration in UO2+x and U3O8-z. The solution and the migration of a range of volatile fission products including iodine, cesium, ruthenium in stoichiometric and nonstoichiometric UO2 and U3O8 are considered

  14. Neutronic analysis for the fission Mo-99 production by irradiation of a LEU target at RECH-1 reactor

    International Nuclear Information System (INIS)

    For the purpose of developing the capability to produce fission 99Mo, the Chilean Nuclear Energy Commission is participating in the IAEA Coordinated Research Project: 'Developing Techniques for Small Scale Indigenous Mo-99 Production using LEU Fission or Neutron Activation'. Fission 99Mo will be produced irradiating, at RECH-1 reactor, a target made of a LEU metallic uranium foil held between two concentric aluminum tubes. KAERI will provide the LEU foil. Neutronic calculations were performed to estimate the fission products activity for a 13 grams LEU foil annular target, which will be irradiated at the level power of 5 MW during 48 hours. (author)

  15. A brief history of the fission products and transuranium elements research in Japan

    International Nuclear Information System (INIS)

    The history of Japanese participation in the research of fission products and transuranium elements from the installation of the first cyclotron in 1937 through the physical and radiochemical research efforts concerning the atomic bomb disaster in 1945 to the investigations on nuclear energy to solve present problems of environmental toxicology is briefly overviewed

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

    International Nuclear Information System (INIS)

    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.

  17. Effect of fission products accumulation on thermophysical properties of oxide fuels for fast reactors

    International Nuclear Information System (INIS)

    It is important to understand the behavior of fission products under irradiation. In this paper, recent activities for obtaining the fundamental findings concerning the effect of FPs accumulation on the thermophysical properties of oxide fuels for fast reactors are presented. (author)

  18. Fission product yields measurement of 232Th induced by 14.8 MeV neutron

    International Nuclear Information System (INIS)

    The relative cumulative yields of 62 fission products were determined using γ-spectrum method for 232Th induced by 14.8 MeV neutron. Using chain yields summation 200% normalization method, 47 chain yields were given, and the data precision is better than 10%. (authors)

  19. 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)

  20. 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

  1. The production and transmission of covariance in the evaluation processing of fission yield data

    International Nuclear Information System (INIS)

    The production and transmission of correlation in the evaluation processing of fission yield data, including average with weight, ratio and sum consistence adjusting, are researched. The variation of the averaged and adjusted yields and/or rations with the correlation coefficient of the input data are investigated. The results obtained are reasonable in physics

  2. Corrections and supplements to the fission product data libraries BIBFP and BIBGRFP

    International Nuclear Information System (INIS)

    Corrections and addenda between 1972 and Feb 15, 1977 are presented to the BIBFP library on fission products half-lives. Tabulated data include the names of the isotopes, the value of half-life listed in the BIBFP library, data taken from the Karlsruhe Nuklidkarte and in the CEA-N-1822 report, and the recommended half-life values. (J.B.)

  3. Influence of radiation on formation of fission product aerosols during LWR degraded core accidents

    International Nuclear Information System (INIS)

    Purpose of this paper is to construct a model for estimating the number density of ions produced by the high radiation levels in reactor core and upper plenum and to use this estimate to determine the effect of ions on the formation of fission product aerosols

  4. Fission product transport in the reactor coolant system for a spectrum of interfacing system LOCA scenarios

    International Nuclear Information System (INIS)

    One of the most important potential severe accident sequences for any pressurized water reactor (PWR) is a loss of coolant accident (LOCA), or V-sequence, in one of the interfacing systems. As initially described in the reactor safety study WASH-1400, interfacing system LOCAs involved the failure of check valves in emergency core cooling systems (ECCS), but could also involve the residual heat removal (RHR) systems. The check valves protect the low-pressure portions of these systems from the high pressures of the reactor coolant system (RCS) to which they are connected to provide cold leg injection. A consequent break in the low-pressure piping outside the containment may result in core damage and a direct pathway for fission products to be transported from the core, through the RCS and ECCS or RHR to the auxiliary building, from which they can escape to the environment. This paper addresses the retention and transport of fission products (specifically, CsI) in the RCS in V-sequence scenarios. It summarizes some of the major differences between models resulting from the latest version of the industry degraded core rulemaking (IDCOR) MAAP Computer Program, MAAP 3.0B. Discussed are the differences in: fission product transport and retention in small, medium, and large ECCS pipe breaks, as well as the effect of ECCS and auxiliary feedwater (AFW) system operation and fission product retention in the various regions of the RCS as calculated by MAAP 3.0B and the STCP

  5. 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.

  6. Fission-product transport and retention in the PHTS under accident conditions

    International Nuclear Information System (INIS)

    The current CANDU safety analysis methodology for predicting the release of radionuclides into containment is based on the bounding assumption that fission products released from the fuel go directly into containment. Allowing for FP retention in the PHTS will help achieve the following objectives: (1) improved estimates of doses to safety equipment in environmental qualification (EQ) analyses, (2) improved estimates of public and operator doses from an improved assessment of less volatile radionuclide behaviour, (3) improved ability to perform best-estimate safety analyses, (4) improved post-accident management plans from a better knowledge of FP location, and (5) less restrictive exclusion area boundary (EAB) designs from better source term estimates. Two LWR fission-product behaviour codes, VICTORIA and SOPHAEROS, have been assessed for their ability to provide a CANDU PHTS FP transport and retention modelling capability. The assessment of VICTORIA and SOPHAEROS was conducted by comparing the features of the two codes with the requirements for CANDU PHTS fission-product transport software, and performing simulations representative of the Loss-of-Coolant Accident with additional Loss of Emergency Coolant Injection (LOCA/LOECI) and stagnation feeder break scenarios with both codes. Based on this assessment, SOPHAEROS is better suited for simulating fission-product transport and retention in the PHTS for CANDU safety and licensing analysis, and VICTORIA should be retained to support more detailed calculations and R and D activities. (author)

  7. 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.

  8. Nuclear data project in Korea and resonance parameter evaluation of fission products

    International Nuclear Information System (INIS)

    Nuclear data activities in the fields of evaluation, processing, measurement, and service in Korea are presented in this paper. As one of the current activities, the neutron resonance parameters for stable or long-lived nineteen fission products have been evaluated and the results are presented here. (author)

  9. 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.)

  10. FISSION-PRODUCT SEPARATION BASED ON ROOM-TEMPERATURE IONIC LIQUIDS

    Science.gov (United States)

    The objectives of this project are (a) to synthesize new ionic liquids tailored for the extractive separation of Cs + and Sr 2+; (b) to select optimum macrocyclic extractants through studies of complexation of fission products with macrocyclic extractants and transport in new ext...

  11. Use of ELOCA.Mk5 to calculate transient fission product release from CANDU fuel elements

    International Nuclear Information System (INIS)

    A change in fuel element power output, or a change in heat transfer conditions, will result in an immediate change in the temperature distribution in a fuel element. The temperature distribution change will be accompanied by concomitant changes in fuel stress distribution that lead, in turn, to a release of fission products to the fuel-to-sheath gap. It is important to know the inventory of fission products in the fuel-to-sheath gap, because this inventory is a major component of the source term for many postulated reactor accidents. ELOCA.Mk5 is a FORTRAN-77 computer code that has been developed to estimate transient releases to the fuel-to-sheath gap in CANDU reactors. ELOCA.Mk5 is an integration of the FREEDOM fission product release model into the ELOCA fuel element thermo-mechanical code. The integration of FREEDOM into ELOCA allows ELOCA.Mk5 to model the feedback mechanisms between the fission product release and the thermo-mechanical response of the fuel element. This paper describes the physical model, gives details of the ELOCA.Mkt code, and describes the validation of the model. We demonstrate that the model gives good agreement with experimental results for both steady state and transient conditions

  12. 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

  13. High-flux fusion neutron source for transmutation of fission products

    International Nuclear Information System (INIS)

    Transmutation of long-lived fission products is often referred to as a crucial step toward harmonized Self-Consistent Nuclear Energy System. The feasibility of their incineration is determined by available neutron excess in the nuclear energy system as a whole and appropriate transmutation environment (neutron spectra and flux) in a dedicated transmuter. The present paper highlights the remarkable transmutation environment of Fusion Neutron Source with ITER-like plasma parameters in approaching the transmutation characteristics which are superior to that of fission and accelerator based transmuters. Quantitative example of zirconium and cesium transmutation is addressed. (author)

  14. 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)

  15. 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)

  16. Compilation of Data on Radionuclide Data for Specific Activity, Specific Heat and Fission Product Yields

    Energy Technology Data Exchange (ETDEWEB)

    Gibbs, A.; Thomason, R.S.

    2000-09-05

    This compilation was undertaken to update the data used in calculation of curie and heat loadings of waste containers in the Solid Waste Management Facility. The data has broad general use and has been cross-checked extensively in order to be of use in the Materials Accountability arena. The fission product cross-sections have been included because they are of use in the Environmental Remediation and Waste Management areas where radionuclides which are not readily detectable need to be calculated from the relative fission yields and material dispersion data.

  17. Compilation of Data on Radionuclide Data for Specific Activity, Specific Heat and Fission Product Yields

    International Nuclear Information System (INIS)

    This compilation was undertaken to update the data used in calculation of curie and heat loadings of waste containers in the Solid Waste Management Facility. The data has broad general use and has been cross-checked extensively in order to be of use in the Materials Accountability arena. The fission product cross-sections have been included because they are of use in the Environmental Remediation and Waste Management areas where radionuclides which are not readily detectable need to be calculated from the relative fission yields and material dispersion data

  18. Effects of time and other variables on fission product release rates

    International Nuclear Information System (INIS)

    The releases of krypton and cesium from highly irradiated LWR fuel have been examined in detail. The main interest has been the effect of time on the rate of release and the effects of heatup and cooldown cycles. The minute-by-minute release rates for fission product 85Kr from commercial fuel irradiated in the H.B. Robinson PWR are shown. The release rate, fraction per minute, is calculated in the same manner as release rates given in NUREG-0772; the fission gas, cesium, and iodine release rate curve from that report is also shown

  19. 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

  20. 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

  1. 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

  2. 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

  3. Determination of short-lived fission product yields with gamma spectrometry

    International Nuclear Information System (INIS)

    The majority of fission yield measurements to date have examined cumulative yields of long-lived nuclides. We present a method for determining independent as well as cumulative fission yields using gamma spectrometry and a Bayesian inverse analysis. This paper outlines the impetus for new fission product yield measurements, the methodology developed to measure these and other nuclear parameters, and initial experimental results for long-lived nuclides and sensitivity analyses. In initial scoping measurements, the cumulative yield of 140Ba was estimated as 4.9966±0.3309 %, and the independent yield of 140La was estimated to be 0.0045±0.0022 %. These estimated values are commensurate with existing literature values. (author)

  4. 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.

  5. Interface between thermal-hydraulics and fission product transport in severe accident analysis

    International Nuclear Information System (INIS)

    Studies of the effectiveness of the containment following hypothetical severe accidents in LWRs have traditionally been split into two separate disciplines. The first is the thermal-hydraulics of the containment atmosphere, leading to an estimate of the magnitudes of the threats to the integrity of the containment building. The second is the study of the transport of the fission products released from the core within and from the building, the ultimate output of which, given a building damage state, is a radiological source term to the environment. In fact, the transport of fission products is strongly influenced by the prevailing thermal-hydraulic conditions, and there are some ways in which the fission products have an influence back on the thermal-hydraulics. The paper describes work funded and co-ordinated by the Commission of the European Communities (CEC) to investigate how the two sorts of calculation, the thermal-hydraulic and the fission product transport, can be made more responsive to the needs of each other. Recent theoretical and experimental work has indicated that the deposition of fission product aerosols is markedly enhanced by the condensation of steam on the particles. The rate of such condensation is a sensitive function of the thermodynamic conditions in the atmosphere, and the release of latent heat because of condensation can in turn significantly change these conditions. The rates of condensation are increased if the particles contain chemical components which are water soluble. To model this effect one requires considerable knowledge of the thermodynamic properties of the relevant aqueous solutions. Recent experience has shown that considerable differences occur between the predictions of different codes. In order to resolve these discrepancies, the CEC has organised a code comparison exercise based on the LACE LA-4 experiment and is currently organising another code comparison based on one of the experiments in the DEMONA series. (author

  6. 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)

  7. Measured and calculated fission-product poisoning in neutron-irradiated uranium-233

    International Nuclear Information System (INIS)

    Samples of 233U and of natural thorium have been irradiated in high neutron-flux facilities, in both soft and hard neutron spectra, and for both short and long exposure times. Included are exposures resulting in depletions of more than 90 percent of the 233U in the fissile material and burnups of more than 30,000 MWd/MT in the fertile material. Fission-product poison cross sections in two energy groups (thermal and epithermal) exhibit differences between measurement and calculation that are believed to be attributable to a lack of adequate information on important fission products in the literature. Experimental results for transient absorbers in irradiated 233U give at least 20,000 b for the neutron absorption resonance integral of 149Pm. This is a factor of 15 higher than that obtained by a 1/v extrapolation of the thermal cross sections. For transient 135Xe, the measured absorption is 7.5 percent higher than that calculated using ENDF/B-IV data. Information is also provided concerning such matters as fission yields and neutron absorption of neodymium isotopes, the existence of significant transient fission-product poisons other than 135Xe and 149Sm, and the shielding of 233U by 232Th. Such shielding suggests the need for a change in the energy dependence of the 232Th thermal-neutron cross section

  8. A Covariance Generation Methodology for Fission Product Yields

    OpenAIRE

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

    2016-01-01

    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 reac...

  9. 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.

  10. Sorption of radioactive products of corrosion and fissions on montmorillonite and kaolinite

    International Nuclear Information System (INIS)

    Sorption of radioactive products of corrosion of 59Fe, 58Co, 54Mn and fission products of 137Cs+137mBa, 91Y, 95Nb, 125Sb and other radionuclides on montmorillonite and kaolinite at different ratio and in a wide range of ph medium is studied. Optimal conditions of extraction of radionuclides from solutions of NO3-, SO42-, F-, H2PO4-, C2O42-are defined.

  11. Separation of valuable fission products from High Level Liquid Waste (HLLW)

    International Nuclear Information System (INIS)

    High Level Liquid Waste (HLLW) generated during spent fuel reprocessing (PUREX) is an important source of valuable fission products viz. 137Cs, 90Sr (and its daughter product 90Y), platinum group metals like 106Ru , Pd etc. We present a short overview of the work carried out in our laboratory on the separation of the valuable radionuclides from actual research reactor HLLW employing a host of separation techniques like solvent extraction, precipitation, liquid membrane etc. (author)

  12. Revision and implementation of ISO 9001 of the 99Mo fission production process

    International Nuclear Information System (INIS)

    The work involves the survey of the plant production processes in order to analyze their strengths and weaknesses related to process improvements.Such a task includes the documentation system - procedures, instruction manual, descriptions, documents, operation plans, trials and control and the quality plan with the ultimate goal of obtaining certification according to ISO 9001:2000, quality management standard for the fission 99Mo production plant at Ezeiza Atomic Center

  13. 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

  14. 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

  15. Experimental tests of the production process of Mo-99 fission in the hot cell

    International Nuclear Information System (INIS)

    The production method of 99 Mo of fission obtains to this with a specific activity of several orders of great magnitude to the one obtained by other methods (as that of irradiation of a target constituted by an alloy or that of irradiation with neutrons of targets of molybdenum of natural isotopic composition or enriched with 98 Mo) and perhaps the most important it is that by this method hundred of Ci of 99 Mo can be obtained by production process. The development of the production process of 99 Mo of fission, is closely linked with the development of techniques for the handling of high radioactivities, particularly the handling of radioactive gases, also with the deposit and elimination of radioactive wastes and with the construction of safety targets for its irradiation in a nuclear reactor. (Author)

  16. 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...

  17. Some topics on improvement in production technology and fission product behavior of HTTR coated particle fuel

    International Nuclear Information System (INIS)

    In collaborations of the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries, Ltd. (NFI), various improvements have been made on the production processes of coated fuel particles and fuel compacts, in order to reduce the failure fraction of the fuel for the High Temperature Engineering Test Reactor (HTTR). Irradiation performance of the fuels recently produced by the improved technology was successfully demonstrated by the 13th OGL-1 fuel assembly irradiation, performed in an in-pile gas loop, OGL-1. In the field of fission product behavior, a peculiar distribution of 110mAg was noticed in its radial profile in the graphite sleeve; the concentration of this nuclide was markedly reduced with increasing concentration of metallic impurities including 60Co. This behavior may be ascribed to competitive co-sorption of Ag and the impurities for the trapping sites on the surface of open pores in the graphite material. Obtained results so far suggest that this nuclide should be retained effectively in the highly-purified graphite sleeve under the HTTR conditions. As for the ZrC-coated particle fuel being developed as an advanced HTTR fuel, much better retention of Cs was observed than for SiC-coated particles in isochronal annealing experiments at temperatures between 1500 and 2300degC; while relatively poor retention of Ru was observed for ZrC-coated particles in this temperature range. (author)

  18. Preliminary evaluation of NMCA effects on corrosion and fission products transport in BWRs

    International Nuclear Information System (INIS)

    Under hydrogen water chemistry conditions in BWR-type reactors, the enhanced N-16 activity transport through the main steam line to the turbine system becomes the major problem in plant operation. To overcome the N-16 problem and to mitigate of intergranular stress corrosion cracking, the noble metal chemical addition technology is employed to lower the electrochemical corrosion potential of structure material with very little hydrogen addition in the feed water. This paper reviews the fundamentals of this technology and provides evaluation of its effects on the corrosion and fission products transport behavior in the BWR primary system. Noble metals, Pt or Rh particulates deposited on the stainless steel surfaces, dramatically improve the hydrogen and oxygen recombination rate thus lowering electrochemical corrosion potential in the presence of excess hydrogen. The paper then describes effects on corrosion product transport, such as variation of Co-60 concentration in reactor water, variation in out-of-core radiation fields, crud spalling from fuel cladding deposits, recoil fission product activity release, and fission product iodine release. Furthermore, laboratory tests on the formation and stability of (Zn, Co)-chromates with hydrogen peroxide are suggested in order to clarify the role of Cr behavior in corrosion product transport in the BWR system. (S. Ohno)

  19. 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

  20. Instrumental neutron activation analysis in geochemistry. Emphasis on spectral and uranium fission product interferences

    International Nuclear Information System (INIS)

    Full text: Since the advent of semiconductor detectors, several contributions to the INAA methodology allowed to resolve and analyse complex gamma ray spectra enhancing thus, the reliability of this analytical technique. Despite attainable performances, some difficulties due to spectral and fission product interferences remain affecting thus the reliability of analytical results. A typical case is that discussed by Landsberger et al. on the determination of Sm in the presence Gd, U and Th. In practice spectral interferences in INAA are resolved by allowing shorter-lived radionuclides to decay and counting the remaining long-lived radionuclides. It is also proceeded to the subtraction of contributions of interfering radionuclides using their interference free analytical photopeaks. Both of them are tedious and time consuming, particularly if radionuclides of interest have similar γray spectra, as is the case for 153Sm and 153Gd to which further interferences come from 239Np and 233Pa. On the other hand, the presence of fissionable products such as U and Th in geological samples, and particularly in high grade U and Th samples, enhances difficulties in the accuracy attainable by instrumental neutron activation analysis. They give rise, during irradiation, to identical (n,γ) products from isotopes of natural elements. The present contribution deals with 2 aspects of resolution of interferences in instrumental neutron activation analysis: 1. The application of a multicomponent method for the resolution of spectral interferences in gamma spectrometry using simultaneous equations method. Assume m elements are to be determined by INAA whose (n,γ) products interfere at their n analytical γ-rays. An overall analytical response can be assumed to consist of several additive individual responses from m interfering radionuclides. The mathematical terms can be expressed by means of the following equation: Pn=k+Σain Aiεi Pn = Photopeak area at the specific interfering n