WorldWideScience

Sample records for transmuting fission products

  1. Transmutation of fission products through accelerator

    International Nuclear Information System (INIS)

    Nakamura, H.; Tani, S.; Takahashi, T.; Yamamura, O.

    1995-01-01

    The transmutation of fission products through particle accelerators has been studied under the OMEGA program. The photonuclear reaction has also been investigated to be applied to transmuting long-lived fission products, such as Cesium and Strontium, which have difficulties on reaction with neutrons due to its so small cross section. It is applicable for the transmutation if the energy balance can be improved with a monochromatic gamma rays in the range of the Giant Dipole Resonance generated through an excellent high current electron linear accelerator. The feasibility studies are being conducted on the transmutation system using it through an electron accelerator. (authors)

  2. 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.)

  3. 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.)

  4. 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.)

  5. 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.)

  6. Actinide and fission product separation and transmutation

    International Nuclear Information System (INIS)

    1991-01-01

    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.)

  7. Actinide and Fission Product Partitioning and Transmutation

    International Nuclear Information System (INIS)

    2015-06-01

    The benefits of partitioning and transmutation (P and T) have now been established worldwide and, as a result, many countries are pursuing R and D programmes to advance the technologies associated with P and T. In this context, the OECD Nuclear Energy Agency (NEA) has organised a series of biennial information exchange meetings to provide experts with a forum to present and discuss state-of-the-art developments in the field of partitioning and transmutation since 1990. The OECD Nuclear Energy Agency Information Exchange Meeting on Actinides and Fission Products Partitioning and Transmutation is a forum for experts to present and discuss the state-of-the-art development in the field of P and T. Thirteen meetings have been organised so far and held in Japan, the United States, France, Belgium, Spain, the Republic of Korea and the Czech Republic. This 13. meeting was hosted by Seoul National University (Seoul, Republic of Korea) and was organised in co-operation with the International Atomic Energy Agency (IAEA) and the European Community (EC). The meeting covered strategic and scientific developments in the field of P and T such as: fuel cycle strategies and transition scenarios, the role of P and T in the potential evolution of nuclear energy as part of the future energy mix; radioactive waste management strategies; transmutation fuels and targets; advances in pyro and aqueous separation processes; P and T specific technology requirements (materials, spallation targets, coolants, etc.); transmutation systems: design, performance and safety; impact of P and T on the fuel cycle; fabrication, handling and transportation of transmutation fuels. A total of 103 presentations (39 oral and 64 posters) were discussed among the 110 participants from 19 countries and 2 international organisations. The meeting consisted of one plenary session where national and international programmes were presented followed by 5 technical sessions: - Fuel Cycle Strategies and Transition

  8. Status report on actinide and fission product transmutation studies

    International Nuclear Information System (INIS)

    1997-06-01

    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

  9. Transmutation of fission products and actinide waste at Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Daemen, L.L.; Pitcher, E.J.; Russell, G.J. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    The authors studied the neutronics of an ATW system for the transmutation of the fission products ({sup 99}Tc in particular) and the type of actinide waste stored in several tanks at Hanford. The heart of the system is a highly-efficient neutron production target. It is surrounded by a blanket containing a moderator/reflector material, as well as the products to be transmuted. The fission products are injected into the blanket in the form of an aqueous solution in heavy water, whereas an aqueous actinides slurry is circulated in the outer part of the blanket. For the sake of definiteness, the authors focussed on {sup 99}Tc (the most difficult fission product to transmute), and {sup 239}Pu, {sup 237}Np, and {sup 241}Am. Because of the low thermal neutron absorption cross-section of {sup 99}Tc, considerable care and effort must be devoted to the design of a very efficient neutron source.

  10. High flux transmutation of fission products and actinides

    International Nuclear Information System (INIS)

    Gerasimov, A.; Kiselev, G.; Myrtsymova, L.

    2001-01-01

    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

  11. Transmutation of long-lived fission products

    International Nuclear Information System (INIS)

    Abrahams, K.

    1994-01-01

    The time-accumulated dose related to technetium dominates the leakage doses in most scenarios for imperturbed geological disposal. If human intrusion into geologically stable repositories or other disturbances is taken into account, the actinides determine the maximum value of the expected individual dose rates of shorter storage times. Therefore actinides dominate the discussion on transmutation of nuclear waste. In principle current LWRs could be used for a massive transmutation of Tc and perhaps I. Fast reactors and HWRs have attractive potential with respect to transmutation in moderated assemblies. HWRs like CANDU have easy refuelling possibilities. (orig.)

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

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    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., 137 Cs, 90 Sr, 129 I, 99 Tc, 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/cm 3 are required for a practical fusion-based fission product transmutation system

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

    International Nuclear Information System (INIS)

    Janssen, A.J.

    1994-01-01

    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. Transmutation of nuclear waste. Status report RAS programme 1994: Recycling and transmutation of actinides and fission products

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Gruppelaar, H.; Franken, W.M.P.

    1995-07-01

    This report describes the status and progress of the Dutch RAS programme on 'Recycling and Transmutation of Actinides and Fission Products' over the year 1994, which is the first year of the second 4-year programme. This programme is outlined and a short progress report is given over 1994, including a listing of 23 reports and publications over the year 1994. Highlights of 1994 were: The completion of long-lived fission-product transmutation studies, the initiation of small-scale demonstration experiments in the HFR on Tc and I, the issue of reports on the potential of the ALMR (Advanced Liquid Metal Reactor) for transmutation adn the participation and international cooperation on irradiation experiments with actinides in inert matrices. The remaining chapters contain more extended contributions on recent developments and selected topics, under the headings: Benefits and risks of partitioning and transmutation, Perspective of chemical partitioning, Inert matrices, Evolutionary options (MOX), Perspective of heavy water reactors, Perspective of fast burners, Perspective of accelerator-based systems, Thorium cycle, Fission-product transmutation, End scenarios, and Executive summary and recommendations. (orig.)

  15. Transmutation of nuclear waste. Status report RAS programme 1994: Recycling and transmutation of actinides and fission products

    Energy Technology Data Exchange (ETDEWEB)

    Cordfunke, E H.P.; Gruppelaar, H; Franken, W M.P.

    1995-07-01

    This report describes the status and progress of the Dutch RAS programme on `Recycling and Transmutation of Actinides and Fission Products` over the year 1994, which is the first year of the second 4-year programme. This programme is outlined and a short progress report is given over 1994, including a listing of 23 reports and publications over the year 1994. Highlights of 1994 were: The completion of long-lived fission-product transmutation studies, the initiation of small-scale demonstration experiments in the HFR on Tc and I, the issue of reports on the potential of the ALMR (Advanced Liquid Metal Reactor) for transmutation adn the participation and international cooperation on irradiation experiments with actinides in inert matrices. The remaining chapters contain more extended contributions on recent developments and selected topics, under the headings: Benefits and risks of partitioning and transmutation, Perspective of chemical partitioning, Inert matrices, Evolutionary options (MOX), Perspective of heavy water reactors, Perspective of fast burners, Perspective of accelerator-based systems, Thorium cycle, Fission-product transmutation, End scenarios, and Executive summary and recommendations. (orig.).

  16. Determination of procedures for transmutation of fission product wastes by fusion neutrons. Volume 2. Final report

    International Nuclear Information System (INIS)

    Lang, G.P.

    1980-12-01

    This study is concerned with the engineering aspects of the transmutation of fission products utilizing neutrons generated in fusion reactors. It is assumed that fusion reactors, although not yet developed, will be available around the turn of the century. Therefore, early studies of this type are appropriate as a guide to the large amount of further investigations that will be needed to fully evaluate this concept. Not all of the radioactive products from light water reactors can be economically transmuted, but it appears that the most hazardous can. This requires that fission-product wastes must first be separated into a number of fractions, and in some instances this must be accomplished with extremely high separation factors. A review of current commercial separation processes and of promising methods that are now in the laboratory stage indicate that the necessary processes can most likely be developed but will require an active and sustained development program. Current fusion reactor concepts were examined as to their suitability for transmuting the separated fission wastes. It was concluded that the long-lived fission products were most amenable to transmutation. The medium-lived fission products, Cs-137 and Sr-90, require higher neutron fluxes than are available in the most developed fusion reactor concepts. Concepts which are less developed may eventually be adaptable as transmuters of these fission products

  17. Transmutation of long-lived fission product (137Cs, 90Sr) by a reactor-accelerator system

    International Nuclear Information System (INIS)

    Toyama, Shin-ichi; Takashita, Hirofumi; Konashi, Kenji; Sasao, Nobuyuki; Sato, Isamu.

    1990-01-01

    The report discusses the transmutation of long-lived fission products by a reactor and accelerator. It is important to take some criteria into consideration in transmutation disposal. To satisfy the criteria, a combined system of a reactor and an accelerator is proposed for the transmutation. An outline of the transmutation reactor and the accelerator is presented. The transmutation reactor has the ability to transmute a large quantity of fission products. However, it is desirable to have a high transmutation rate as well as a large disposal ability. Besides the transmutation property, it is necessary to investigate the physics of the transmutation reactor such as nuclear characteristics and burnup properties in order to obtain the most suitable, high performance core concept. A study on those properties is also presented. A high power accelerator is required for the transmutation. So a test linac is developed to accelerate high intensity beams. (N.K.)

  18. Transmutations of nuclear waste. Progress report RAS programme 1995: Recycling and transmutation of actinides and fission products

    International Nuclear Information System (INIS)

    Gruppelaar, H.; Cordfunke, E.H.P.; Konings, R.J.M.; Bultman, J.H.; Dodd, D.H.; Franken, W.M.P.; Kloosterman, J.L.; Koning, A.J.; Wichers, V.A.

    1996-04-01

    This report describes the progress of the Dutch RAS programme on 'Recycling and Transmutation of Actinides and Fission Products' over the year 1995, which is the second year of the 4-year programme 1994-1997. An extensive listing of reports and publications from 1991 to 1995 is given. Highlights in 1995 were: -The completion of the European Strategy Study on Nuclear Waste Transmutation as a result of which the understanding of transmutation of plutonium, minor actinides and long-lived fission products in thermal and fast reactors has been increased significantly. Important ECN contributions were given on Am, 99 Tc and 129 I transmutation options. Follow-up contracts have been obtained for the study of 100% MOX cores and accelerator-based transmutation. - Important progress in the evaluation of CANDU reactors for burning very large amounts of transuranium mixtures in inert matrices. - The first RAS irradiation experiment in the HFR, in which the transmutation of technetium and iodine was examined, has been completed and post-irradiation examination has been started. - A joint proposal of the EFTTRA cooperation for the 4 th Framework Programme of the EU, to demonstrate the feasibility of the transmutation of americium in an inert matrix by an irradiation in the HFR, has been granted. - A bilateral contract with CEA has been signed to participate in the CAPRA programme, and the work in this field has been started. - The thesis work on Actinide Transmutation in Nuclear Reactor Systems was succesfully defended. New PhD studies on Pu burning in HTGR, on nuclear data for accelerator-based systems, and on the SLM-technique for separation of actinides were started. - A review study of the use of the thorium cycle as a means for nuclear waste reduction, has been completed. A follow-up of this work is embedded in an international project for the 4th Framework Programme of the EU. (orig./DG)

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

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    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., 137 Cs, 90 Sr, 129 I, 99 Tc, 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. Role of (n,2n) reactions in transmutation of long-lived fission products

    Energy Technology Data Exchange (ETDEWEB)

    Apse, V. A.; Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Kulikov, E. G. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    The conditions under which (n,γ) and (n,2n) reactions can help or hinder each other in neutron transmutation of long-lived fission products (LLFPs) are considered. Isotopic and elemental transmutation for the main long-lived fission products, {sup 79}Se, {sup 93}Zr, {sup 99}Tc, {sup 107}Pd, {sup 126}Sn, {sup 129}I, and {sup 135}Cs, are considered. The effect of (n,2n) reactions on the equilibrium amount of nuclei of the transmuted isotope and the neutron consumption required for the isotope processing is estimated. The aim of the study is to estimate the influence of (n,2n) reactions on efficiency of neutron LLFP transmutation. The code TIME26 and the libraries of evaluated nuclear data ABBN-93, JEF-PC, and JANIS system are applied. The following results are obtained: (1) The effect of (n,2n) reactions on the minimum number of neutrons required for transmutation and the equilibrium amount of LLFP nuclei is estimated. (2) It is demonstrated that, for three LLFP isotopes ({sup 126}Sn, {sup 129}I, and {sup 135}Cs), (n,γ) and (n,2n) reactions are partners facilitating neutron transmutation. The strongest effect of (n,2n) reaction is found for {sup 126}Sn transmutation (reduction of the neutron consumption by 49% and the equilibrium amount of nuclei by 19%).

  1. Evaluation of transmutation performance of long-lived fission products with a super fast reactor

    International Nuclear Information System (INIS)

    Lu, Haoliang; Han, Chiyoung; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    The performance of the Super Fast Reactor for transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super Fast Reactor. First is in the blanket assembly due to the ZrH 1.7 layer which can slow down the fast neutrons. Second is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected of transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe·y and 2.79%/GWe.y can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the outputs from 11.8 and 6.2 1000MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained by the Super Fast Reactor. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super Fast Reactor. It turns out that the 135 Cs transmutation is a challenge not only for the Super Fast Reactor but also for other commercial fast reactors. (author)

  2. Transmutation of fission products with the use of an accelarator

    International Nuclear Information System (INIS)

    Kase, T.; Harada, H.; Takahashi, T.

    1995-01-01

    The three transmutation methods with the use of an accelerator, the proton method, the spallation neutron method and the μCF method, are employed for the transmutation of long-lived nuclides in high level radioactive wastes. The transmutation energies and the effective half-lives of 99 Tc and 137 Cs for these transmutation methods are calculated by the Monte Carlo simulation codes for particle transport. The transmutation energies of the proton method are larger than those of the spallation neutron method and the μCF method under the condition of the same effective half life. The proton method is difficult to meet energy balance criterion. On the other hand, the spallation neutron method and the μCF method have possibility to meet the energy balance criterion. (author)

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

    International Nuclear Information System (INIS)

    Warin, D.

    2003-01-01

    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)

  4. Actinide and fission product partitioning and transmutation. Status and assessment report

    International Nuclear Information System (INIS)

    1999-01-01

    Implementation and partitioning technology is intended to reduce the inventory of actinides and long-lived fission products in nuclear waste. Such technology can decrease hazards of pre-disposal waste management and of physical disturbance of a waste repository. An authoritative analysis is given of the technical, radiological and economic consequences of the proposed partitioning and transmutation operations on the present and future fuel cycle options. The report is subdivided to a general part for non-specialist readers, and to a technical systems analysis discussing issues on partitioning, transmutation and long-term waste management. (R.P.)

  5. Calculation characterization of spent fuel hazard related to partitioning and transmutation of minor actinides and fission products

    International Nuclear Information System (INIS)

    Gerasimov, A. S.; Bergelson, B. R.; Tikhomirov, G.V.; Volovik, A.I. . E-mail of corresponding author: geras@itep.ru; Gerasimov, A.S.)

    2005-01-01

    Radiotoxicity is one of important characteristics of radwaste hazard. Radiotoxicity of actinides and fission products from spent fuel of VVER-1000 reactor for processes of burnup, long-term storage, and transmutation is discussed. (author)

  6. High-power proton linac for transmuting the long-lived fission products in nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, G.P.

    1991-01-01

    High power proton linacs are being considered at Los Alamos as drivers for high-flux spallation neutron sources that can be used to transmute the troublesome long-lived fission products in defense nuclear waste. The transmutation scheme being studied provides a high flux (> 10{sup 16}/cm{sup 2}{minus}s) of thermal neutrons, which efficiently converts fission products to stable or short-lived isotopes. A medium-energy proton linac with an average beam power of about 110 MW can burn the accumulated Tc99 and I129 inventory at the DOE's Hanford Site within 30 years. Preliminary concepts for this machine are described. 3 refs., 5 figs., 2 tabs.

  7. Partitioning and transmutation of actinides and fission products

    International Nuclear Information System (INIS)

    Baetsle, L.H.

    1993-01-01

    The world's nuclear power plants have a total installed capacity of approximately 340 GWe. They give rise to an annual volume of approximately 9000 t of radioactive waste, which is reprocessed, separated from its plutonium content, contained, and stored in repositories to close the nuclear fuel cycle. Direct disposal is being discussed as an alternative to this procedure. As repositories in suitable types of host rock are not operational, the only viable solution is long-term interim storage above ground. If the volumes of radioactive waste are to be reduced, the longlived actinides and fission products must be partitioned. Isotope partitioning in accelerators, though still sounding like science fiction, may soon be indispensable as the third way of treating radioactive waste. The use of mixed oxide fuel in light water reactors and fast breeder reactors both help to limit waste arisings and protect the long-term continuity of raw materials supply. However, both require public acceptance if they are to succeed. (orig.) [de

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

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    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., 137 Cs or 90 Sr. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n, 2n), (n, α), etc.) that convert the long lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product

  9. Hyper fuse: a novel inertial confinement system utilizing hypervelocity projectiles for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1979-01-01

    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 a target 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., 137 Cs or 90 Sr. The 14 MeV fusion neutrons released during the pellet burn cause transmutation reactions [e.g., (n, 2n), (n, α), etc.] that convert the long lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product

  10. Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

    Science.gov (United States)

    Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

    2017-10-24

    Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

  11. Transmutation of nuclear waste. Status report RAS programme 1993: Recycling and transmutation of actinides and fission products

    International Nuclear Information System (INIS)

    Abrahams, K.; Bultman, J.H.; Cordfunke, E.H.P.; Gruppelaar, H.; Janssen, A.J.; Franken, W.M.P.; Klippel, K.T.; Kloosterman, J.L.; Konings, R.J.M.; Smit, J.

    1994-11-01

    The term ''nuclear transmutation'' means a conversion of long-lived radioactive nuclides into short-lived or stable nuclides and ''recycling'' means re-use of fissile material to generate energy in power reactors. With these two processes a reduction of the radiotoxicity and of its duration may be achieved, thus reducing the potential hazard to future generations. Firstly, the report gives a survey of the present situation regarding nuclear waste: its components, how the waste is produced in current LWR and possible options for interim and final storage. Then the objective of the RAS programme, the working methods and the state of the art of the research are considered. Two chapters deal with preliminary results of national and international research. A rather tentative prediction for the future is formulated. Some conclusions are drawn: It seems to be in the best interests of the Netherlands to continue the established line of reprocessing nuclear waste, should new reactors be introduced. It may be advisable to make international agreements so that in the future fission products will contain as few traces of transuranic actinides and long-lived components as possible. Consequently, nuclear waste would become cleaner in terms of long-lived components. For the transmutation of products separated in foreign countries, the Netherlands could pursue an active policy, perform research and also consider the use of MOX fuel in future Dutch reactors. Further contributions towards the solution of these problems can only be made by the Netherlands on an international level. As such, the research and study performed within the framework of the RAS-programme represents a useful international contribution. The possibilities offered by the HFR are particularly of great value. Finally, the choice of a new generation of nuclear reactors should be made not based only on the safety aspects, but also on the extent of waste production and on the transmutation possibilities (application

  12. Transmutation of nuclear waste. Status report RAS programme 1993: Recycling and transmutation of actinides and fission products

    Energy Technology Data Exchange (ETDEWEB)

    Abrahams, K; Bultman, J H; Cordfunke, E H.P.; Gruppelaar, H; Janssen, A J; Franken, W M.P.; Klippel, K T; Kloosterman, J L; Konings, R J.M.; Smit, J

    1994-11-01

    The term ``nuclear transmutation`` means a conversion of long-lived radioactive nuclides into short-lived or stable nuclides and ``recycling`` means re-use of fissile material to generate energy in power reactors. With these two processes a reduction of the radiotoxicity and of its duration may be achieved, thus reducing the potential hazard to future generations. Firstly, the report gives a survey of the present situation regarding nuclear waste: its components, how the waste is produced in current LWR and possible options for interim and final storage. Then the objective of the RAS programme, the working methods and the state of the art of the research are considered. Two chapters deal with preliminary results of national and international research. A rather tentative prediction for the future is formulated. Some conclusions are drawn: It seems to be in the best interests of the Netherlands to continue the established line of reprocessing nuclear waste, should new reactors be introduced. It may be advisable to make international agreements so that in the future fission products will contain as few traces of transuranic actinides and long-lived components as possible. Consequently, nuclear waste would become cleaner in terms of long-lived components. For the transmutation of products separated in foreign countries, the Netherlands could pursue an active policy, perform research and also consider the use of MOX fuel in future Dutch reactors. Further contributions towards the solution of these problems can only be made by the Netherlands on an international level. As such, the research and study performed within the framework of the RAS-programme represents a useful international contribution. Finally, the choice of a new generation of nuclear reactors should be made not based only on the safety aspects, but also on the extent of waste production and on the transmutation possibilities (application of MOX, etc.). (orig./HP).

  13. High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products

    International Nuclear Information System (INIS)

    Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

    1993-01-01

    An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ''engineered safety features,'' which, along with the use of high temperature capable materials further enhance its safety characteristics

  14. Fission product yield data for the transmutation of minor actinide nuclear waste

    International Nuclear Information System (INIS)

    2008-04-01

    A report issued by an international study group for the transmutation of nuclear waste using accelerator driven systems has highlighted the need for specific sets of nuclear data. These authoritative requirements include fission product yields at an intermediate incident neutron energy of up to 150 MeV. Before the start of the present CRP on fission product yield data for the transmutation of nuclear waste, only four types of evaluated fission yield data sets existed, namely for spontaneous fission, and for fission induced by thermal, fast (or fission) spectrum, and by 'high energy' (14-15 MeV) neutrons. A new type of evaluation for energy dependent neutron induced fission yields was required for this project. In view of the scarcity of experimental data, such an evaluation has to be based on systematics and theoretical model calculations. Unlike fission cross-sections, where nuclear models are being used successfully for the calculation of unmeasured cross-section ranges, such models or theories existed only for low energy fission yields. Hence the CRP participants entered a completely new field of research for which the progress and outcome were unpredictable. Clearly the ultimate goal of such an effort, namely an evaluation of energy dependent fission yields, could not be realized within the perceived lifetime of a CRP. The main emphasis of the CRP was on the development of adequate systematics and models for the calculation of energy dependent fission yields up to 150 MeV incident neutron energy. Several problems had to be solved, such as the correct choice of model parameters and multiplicity distributions of emitted neutrons, and the effect of multi-chance fission. Models and systematics have been tested for lower energy yields, but they failed to reproduce recent experimental data, particularly at higher energies, and the parameters had to be modified. Other models have been developed from the analysis of experimental data in order to derive systematic

  15. Proceedings of the Eleventh Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation

    International Nuclear Information System (INIS)

    2012-01-01

    Partitioning and transmutation (P and T) is one of the key technologies for reducing the radiotoxicity and volume of radioactive waste arisings. Recent developments indicate the need for embedding P and T strategies in advanced fuel cycles considering both waste management and economic issues. In order to provide experts a forum to present and discuss state-of-the-art developments in the P and T field, the OECD/NEA has been organising biennial information exchange meetings on actinide and fission product partitioning and transmutation since 1990. The previous meetings were held in Mito (Japan) in 1990, at Argonne (United States) in 1992, in Cadarache (France) in 1994, in Mito (Japan) in 1996, in Mol (Belgium) in 1998, in Madrid (Spain) in 2000, in Jeju (Korea) in 2002, in Las Vegas (United States) in 2004, in Nimes (France) in 2006 and in Mito (Japan) in 2008. They have often been co-sponsored by the European Commission (EC) and the International Atomic Energy Agency (IAEA). The 11. Information Exchange Meeting was held in San Francisco, California, United States on 1-4 November 2010, comprising a plenary session on national P and T programmes and six technical sessions covering various fields of P and T. The meeting was hosted by the Idaho National Laboratory (INL), United States. The information exchange meetings on P and T form an integral part of NEA activities on advanced nuclear fuel cycles. The meeting covered scientific as well as strategic/policy developments in the field of P and T, such as: fuel cycle strategies and transition scenarios; radioactive waste forms; the impact of P and T on geological disposal; radioactive waste management strategies (including secondary wastes); transmutation fuels and targets; pyro and aqueous separation processes; materials, spallation targets and coolants; transmutation physics, experiments and nuclear data; transmutation systems (design, performance and safety); handling and transportation of transmutation fuels; and

  16. A repository released-dose model for the evaluation of long-lived fission product transmutation effectiveness

    International Nuclear Information System (INIS)

    Davidson, J.W.

    1995-01-01

    A methodology has been developed to quantify the total integrated dose due to a radionuclide species i emplaced in a geologic repository; the focus is on the seven long-lived fission products (LLFPs). The methodology assumes continuous exposure water contaminated with species i at the accessible environment (i.e., just beyond the geologic barrier afforded by the geologic repository). The dose integration is performed out to a reference post-release time. The integrated dose is a function of the total initial inventory of radionuclide i the repository, the time at which complete and instantaneous failure of the engineered barrier (e.g., waste canister) in, a geologic repository occurs, the fractional dissolution rate (from waste solid form) of radionuclide i in ground water, the ground water travel time to the accessible environment, the retardation factor (sorption on the geologic media) for radionuclide i, the time after radionuclide begins to enter the biosphere. In order to assess relative dose, the ratio of total integrated dose to that for a reference LLFP species j (e.g., 99 Tc) was defined. This ratio is a measure of the relative benefit of transmutation of other LLFPs compared to 99 Tc. This methodology was further developed in order to quantify the integrated dose reduction per neutron utilized for LLFP transmutation in accelerator-driven transmutation technologies (ADTT). This measure of effectiveness is a function of the integrated dose due to LLFP species i, the number of total captures in LLFP species i chain per LLFP nuclide fed to the chain at equilibrium, and the number of total captures in related transmutation product (TP) chains per capture in the LLFP species i chain. To assess relative transmutation effectiveness, the ratio of integrated dose reduction per neutron utilization to that for a reference LLFP species j (e.g., 99 Tc) was defined. This relative measure of effectiveness was evaluated LLFP transmutation strategy

  17. Flibe blanket concept for transmuting transuranic elements and long lived fission products

    International Nuclear Information System (INIS)

    Gohar, Y.

    2000-01-01

    A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

  18. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  19. Concept and experimental studies on fuel and target for minor actinides and fission products transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Prunier, C; Guerin, Y [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d` Etudes des Combustibles; Salvatores, M [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires; Zaetta, A [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d` Etudes des Reacteurs

    1994-12-31

    High activity long-lived radionuclides in nuclear wastes, namely minor actinides (americium and neptunium) are in large amount generated by current nuclear reactive. The destruction of these radionuclides is a part of the French SPIN (Partitioning and Burning) program consistent with the determination to send a minimum amount of harmful products for final storage. Transmutation concepts are defined for neptunium and americium taking into account fuel cycle strategies. Neptunium destruction does not pose any major problems. It`s a by-product of uranium consumption, as plutonium and in despite of a slight gamma activity due to the protactinium 233 it`s quite easy to handle. Diluting neptunium in the mixed oxide fuels (MOX) should not be an obstacle for fabrication, in-pile behaviour and reprocessing either. Consequently we make the proposal of homogeneous mode of neptunium in MOX which should be soon explored in the experimental OSIRIS reactor and in the Phenix and Superphenix reactors. The analysis is more complex for the multi isotope americium. Its destruction is difficult because of gamma radioactivity which complicates fabrication. Experiments in Phenix and calculation showed that Phenix reactor offers a good potential for americium incineration, but similar data do not exist for PWR. It will remain a well known difficulty for fabrication and reprocessing. In this case we have to put a real new face to the fabrication flow-sheet of americium compounds and we propose to develop the heterogeneous mode. Targets choice are defined in term of: -safety, considering fuel reaction with cladding and water sodium, -transmutation rate, limited by target behaviour, in FR`s (Phenix), PWR`s (OSIRIS) and HFR (Petten), -reprocessing, checking the solubility of such targets by Purex process. So, at the beginning of our program the account has been on improving fuel and targets properties related to safety and fuel cycle. (authors). 4 figs.

  20. Transmutation of Tc-99 in fission reactors

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1994-12-01

    Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

  1. Safety and environmental aspects of partitioning and transmutation of actinides and fission products. Proceedings of a technical committee meeting held in Vienna, 29 November - 2 December 1993

    International Nuclear Information System (INIS)

    1995-01-01

    There is considerable interest in many countries in the partitioning and transmutation of long lived radionuclides as a potential complement to the closed fuel cycle. Recognizing this, the IAEA organized a Technical Committee Meeting on Safety and Environmental Aspects of Partitioning and Transmutation of Actinides and Fission Products, to review the current status of progress of national and international programmes and identify the most important directions of co-operation. The results of the Technical Committee meeting are presented in this document. Refs, figs and tabs

  2. Safety and environmental aspects of partitioning and transmutation of actinides and fission products. Proceedings of a technical committee meeting held in Vienna, 29 November - 2 December 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-01

    There is considerable interest in many countries in the partitioning and transmutation of long lived radionuclides as a potential complement to the closed fuel cycle. Recognizing this, the IAEA organized a Technical Committee Meeting on Safety and Environmental Aspects of Partitioning and Transmutation of Actinides and Fission Products, to review the current status of progress of national and international programmes and identify the most important directions of co-operation. The results of the Technical Committee meeting are presented in this document. Refs, figs and tabs.

  3. Monte Carlo analysis of the long-lived fission product neutron capture rates at the Transmutation by Adiabatic Resonance Crossing (TARC) experiment

    International Nuclear Information System (INIS)

    Abánades, A.; Álvarez-Velarde, F.; González-Romero, E.M.; Ismailov, K.; Lafuente, A.; Nishihara, K.; Saito, M.; Stanculescu, A.; Sugawara, T.

    2013-01-01

    Highlights: ► TARC experiment benchmark capture rates results. ► Utilization of updated databases, included ADSLib. ► Self-shielding effect in reactor design for transmutation. ► Effect of Lead nuclear data. - Abstract: The design of Accelerator Driven Systems (ADS) requires the development of simulation tools that are able to describe in a realistic way their nuclear performance and transmutation rate capability. In this publication, we present an evaluation of state of the art Monte Carlo design tools to assess their performance concerning transmutation of long-lived fission products. This work, performed under the umbrella of the International Atomic Energy Agency, analyses two important aspects for transmutation systems: moderation on Lead and neutron captures of 99 Tc, 127 I and 129 I. The analysis of the results shows how shielding effects due to the resonances at epithermal energies of these nuclides affects strongly their transmutation rate. The results suggest that some research effort should be undertaken to improve the quality of Iodine nuclear data at epithermal and fast neutron energy to obtain a reliable transmutation estimation.

  4. Monte Carlo analysis of the long-lived fission product neutron capture rates at the Transmutation by Adiabatic Resonance Crossing (TARC) experiment

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, A., E-mail: abanades@etsii.upm.es [Grupo de Modelizacion de Sistemas Termoenergeticos, ETSII, Universidad Politecnica de Madrid, c/Ramiro de Maeztu, 7, 28040 Madrid (Spain); Alvarez-Velarde, F.; Gonzalez-Romero, E.M. [Centro de Investigaciones Medioambientales y Tecnologicas (CIEMAT), Avda. Complutense, 40, Ed. 17, 28040 Madrid (Spain); Ismailov, K. [Tokyo Institute of Technology, 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Lafuente, A. [Grupo de Modelizacion de Sistemas Termoenergeticos, ETSII, Universidad Politecnica de Madrid, c/Ramiro de Maeztu, 7, 28040 Madrid (Spain); Nishihara, K. [Transmutation Section, J-PARC Center, JAEA, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Saito, M. [Tokyo Institute of Technology, 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Stanculescu, A. [International Atomic Energy Agency (IAEA), Vienna (Austria); Sugawara, T. [Transmutation Section, J-PARC Center, JAEA, Tokai-mura, Ibaraki-ken 319-1195 (Japan)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer TARC experiment benchmark capture rates results. Black-Right-Pointing-Pointer Utilization of updated databases, included ADSLib. Black-Right-Pointing-Pointer Self-shielding effect in reactor design for transmutation. Black-Right-Pointing-Pointer Effect of Lead nuclear data. - Abstract: The design of Accelerator Driven Systems (ADS) requires the development of simulation tools that are able to describe in a realistic way their nuclear performance and transmutation rate capability. In this publication, we present an evaluation of state of the art Monte Carlo design tools to assess their performance concerning transmutation of long-lived fission products. This work, performed under the umbrella of the International Atomic Energy Agency, analyses two important aspects for transmutation systems: moderation on Lead and neutron captures of {sup 99}Tc, {sup 127}I and {sup 129}I. The analysis of the results shows how shielding effects due to the resonances at epithermal energies of these nuclides affects strongly their transmutation rate. The results suggest that some research effort should be undertaken to improve the quality of Iodine nuclear data at epithermal and fast neutron energy to obtain a reliable transmutation estimation.

  5. Consultancy to review and finalize the IAEA publication 'Compendium on the use of fusion/fission hybrids for the utilization and transmutation of actinides and long-lived fission products'. Working material

    International Nuclear Information System (INIS)

    2004-01-01

    In addition to the traditional fission reactor research, fusion R and D activities are becoming of interest also to nuclear fission power development. There is renewed interest in utilizing fusion neutrons, Heavy Liquid Metals, and molten salts for innovative systems (energy production and transmutation). Indeed, for nuclear power development to become sustainable as a long-term energy option, innovative fuel cycle and reactor technologies will have to be developed to solve the problems of resource utilization and long-lived radioactive waste management. In this context Member States clearly expressed the need for comparative assessments of various transmutation reactors. Both the fusion and fission communities are currently investigating the potential of innovative reactor and fuel cycle strategies that include a fusion/fission system. The attention is mainly focused on substantiating the potential advantages of such systems: utilization and transmutation of actinides and long-lived fission products, intrinsic safety features, enhanced proliferation resistance, and fuel breeding capabilities. An important aspect of the ongoing activities is the comparison with the accelerator driven subcritical system (spallation neutron source), which is the other main option for producing excess neutrons. Apart from comparative assessments, knowledge preservation is another subject of interest to the Member States: the goal, applied to fusion/fission systems, is to review the status of, and to produce a 'compendium' of past and present achievements in this area

  6. Consultancy on the potential of fusion/fission sub-critical neutron systems for energy production and transmutation. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    The Workshop on Sub-critical Neutron Production held at the University of Maryland and the Eisenhower Institute on 11-13 October 2004 brought together members of fusion, fission and accelerator technical communities to discuss issues of spent fuel, nonproliferation, reactor safety and the use of neutrons for sub-critical operation of nuclear reactors. The Workshop strongly recommended that the fusion community work closely with other technical communities to ensure that a wider range of technical solutions is available to solve the spent fuel problem and to utilize the current actinide inventories. Participants of the Workshop recommended that a follow-on Workshop, possibly under the aegis of the IAEA, should be held in the first half of the year 2005. The Consultancy Meeting is the response to this recommendation. The objectives of the Consultancy meeting were to hold discussions on the role of fusion/fission systems in sub-critical operations of nuclear reactors. The participants agreed that development of innovative (fourth generation) fission reactors, advanced fuel cycle options, and disposition of existing spent nuclear fuel inventories in various Member Sates can significantly benefit from including sub-critical systems, which are driven by external neutron sources. Spallation neutrons produced by accelerators have been accepted in the past as the means of driving sub-critical reactors. The accelerator community deserves credit in pioneering this novel approach to reactor design. Progress in the design and operation of fusion devices now offers additional innovative means, broadening the range of sub-critical operations of fission reactors. Participants felt that fusion should participate with accelerators in providing a range of technical options in reactor design. Participants discussed concrete steps to set up a small fusion/fission system to demonstrate actinide burning in the laboratory and what advice should be given to the Agency on its role in

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

    International Nuclear Information System (INIS)

    Abe, A.Y.

    1989-01-01

    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)

  8. Conceptual design of a fusion-fission hybrid reactor for transmutation of high level nuclear waste

    International Nuclear Information System (INIS)

    Qiu, L.J.; Wu, Y.C.; Yang, Y.W.; Wu, Y.; Luan, G.S.; Xu, Q.; Guo, Z.J.; Xiao, B.J.

    1994-01-01

    To assess the feasibility of the transmutation of long-lived radioactive waste using fusion-fission hybrid reactors, we are studying all the possible types of blanket, including a comparison of the thermal and fast neutron spectrum blankets. Conceptual designs of a small tokamak hybrid blanket with small inventory of actinides and fission products are presented. The small inventory of wastes makes the system safer. The small hybrid reactor system based on a fusion core with experimental parameters to be realized in the near future can effectively transmute actinides and fission products at a neutron wall loading of 1MWm -2 . An innovative energy system is also presented, including a fusion driver, fuel breeder, high level waste transmuter, fission reactor and so on. An optimal combination of all types of reactor is proposed in the system. ((orig.))

  9. Transmutation of Tc-99 and I-129 in fission reactors. A calculational study

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1995-03-01

    The HWR is a better candidate for large-scale transmutation of long-lived fission products. When target pins containing either Tc-99 or I-129 are positioned in the centre of each fuel bundle of a 935 MW e CANDU reactor, the transmutation half lives are 44 and 20 years, respectively, and the gross transmutation rates 60 and 48 kg/a. The positive coolant void coefficient is reduced in both cases with about 30%. When Tc-99 target pins are positioned in the moderator between the fuel bundles, the transmutation half life becomes 25 years and the gross transmutation rate 106 kg/a. This means that one HWR can serve four PWRs with equal power. The fast reactor seems most promising. When Tc-99 target pins are irradiated in moderated subassemblies in the inner core of Superphenix (∼1240 MW e ), a transmutation half life of 15 years is obtained with a gross transmutation rate of 122 kg/a. These values become 18 years and 101 kg/a when non-moderated subassemblies are used for the irradiation. This implies that one fast reactor can serve four to five PWRs with equal power. The PWR seems not very effective for transmutation of Tc-99. Large inventories are needed to obtain a Tc-99 transmutation rate equal to the production rate (18 kg/a for a 900 MW e PWR). When all guide tubes of an UO 2 fuelled PWR are filled with Tc-99 with density of 5 g cm -3 , the transmutation half life is 39 years and the gross transmutation rate 64 kg/a. (orig./GL)

  10. Project 'Installation of a stand at the horizontal channel of the MARIA Research Reactor, Otwock-Swierk, Poland, for the research of transmutation of minor actinides and fission products'

    International Nuclear Information System (INIS)

    Szuta, M.

    2006-01-01

    As a long range objective we would like to focus on management of the fuel economy in the sub-critical assembly of the accelerator driven system (ADS) in terms of long lived fission products (LLFP) and minor actinides (MA) transmutation. Transmutation of the radioactive waste (RW) is an important element within the technical objective of the optimal management of the fuel economy in the sub-critical assembly of the accelerator driven system (ADS). Analysis of possible ways of reduction of radioactive wastes by transmutation of radioactive long-lived fission products such as 99 Tc, 129 I and 135 Cs and by burning up of transuranic nuclides implies that the sub-critical assembly of the accelerator driven system should consist of three zones. The requirement of three zones comes out of the fact that each radioactive isotope to be reduced is to be located in a different spectrum of thermal, epithermal and high energy neutron fluxes. High flux thermal neutron environment (≥10 16 n/cm 3 ·s) is expected as the best way for the transmutation of most of the radioactive waste to stable or short-lived nuclides and for increasing the probability for fission such actinides as 237 Np and 238 Np. The concept of ADS system for energy production and for transmutation is quite new to some extent and from this reason it requires many theoretical and experimental studies. The research of transmutation is a very large area of study requiring a significant experimental and financial support, so it can be performed only within the international cooperation. Specifically, the proposed research within this CRP requires important means, in terms of high-energy proton beams, spallation targets, sub-critical assembly, measurement instrumentation, post-irradiation characterisation and its testing and, of course, manpower for the interpretation of results, modelling observed phenomena, and programme management. The personal involved in the research is to be a skilled personal of

  11. Genetic effects of decay of fuel fission products in cells of yeast Saccharomyces cerevisial. Role of transmutations

    International Nuclear Information System (INIS)

    Shanshiashvili, T.A.; Gracheva, L.M.

    1981-01-01

    Lethal and mutagenic effects and the nature of mutations induced by 91 Y decay in cells of Saccharomyces cerevisial yeasts were studied. An inactivation probability was (9.6+-4.0)x10 -5 decaysup(-I) and mutation probability in ade 1, ade 2 genes was (0.65+-0.53)x10 -8 decaysup(-I). It is shown that 95% of the effect is caused by local transmutation aftereffects of the decay of the isotope in the cell, and only 5% - by β-radiation. It is also shown that 91 Y is one of the most effective physical matagens. Its mutagenic efficiency (M/1 nS) is the same as of 32 P, and 3.5 times higher than the mutagenic efficiency of γ-rays, and 8 times - than of neutrons with E=1.3 MeV. According to the results of the analysis of the nature of mutations, 91 Y is a nonspecific mutagen which induces mutations of different types. Among 4B analyzed ade 2 mutations there have been detected 146(35%) transitions AT→GC, 139(34%) transitions GC→AT, 64(16%) mutations of reading frame shift, 63 (15%) transversions and 1 deletion. The reasons for such significant genetic effect of 91 Y on the cell are discussed [ru

  12. Conceptual design study of Hyb-WT as fusion–fission hybrid reactor for waste transmutation

    International Nuclear Information System (INIS)

    Siddique, Muhammad Tariq; Kim, Myung Hyun

    2014-01-01

    Highlights: • Conceptual design study of fusion-fission hybrid reactor for waste transmutation. • MCNPX and MONTEBURNS are compared for transmutation performance of WT-Hyb. • Detailed neutronic performance of final optimized Hyb-WT design is analyzed. • A new tube-in-duct core design is implemented and compared with pin type design. • Study shows many aspects of hybrid reactor even though scope was limited to neutronic analysis. - Abstract: This study proposes a conceptual design of a hybrid reactor for waste transmutation (Hyb-WT). The design of Hyb-WT is based on a low-power tokamak (less than 150 MWt) and an annular ring-shaped reactor core with metal fuel (TRU 60 w/o, Zr 40 w/o) and a fission product (FP) zone. The computational code systems MONTEBURNS and MCNPX2.6 are investigated for their suitability in evaluating the performance of Hyb-WT. The overall design performance of the proposed reactor is determined by considering pin-type and tube-in-duct core designs. The objective of such consideration is to explore the possibilities for enhanced transmutation with reduced wall loading from fusion neutrons and reduced transuranic (TRU) inventory. TRU and FP depletion is analyzed by calculating waste transmutation ratio, mass burned per full power year (in units of kg/fpy), and support ratio. The radio toxicity analysis of TRUs and FPs is performed by calculating the percentage of toxicity reduction in TRU and FP over a burn cycle

  13. APEX accelerator cycle for transmutation of long-lived fission wastes

    International Nuclear Information System (INIS)

    Powell, J.; Steinberg, M.; Takahashi, H.; Grand, P.; Botts, T.; Kouts, H.J.C.

    1980-01-01

    Based on preliminary studies, some conclusions can be drawn concerning the Accelerator Fuel Enricher and Fission Product Exterminator (APEX). APEX-1 and APEX-2 systems can destroy TU's, 137 Cs, and 90 Sr at acceptable cost and efficiency. The principal difference between APEX-1 and APEX-2 is the in-reactor and in-circuit inventory of 137 Cs and 90 Sr. Stable and low hazard wastes can be disposed of by burial. Accelerator breeders can effectively sustain a fission reactor economy indefinitely. Military waste can be blended into commercial fuel cycle for transmutation. Accelerator and target technologies appear practical and could be developed in a few years. More detailed studies are needed to better define the technical and economic features of the LAFER and APEX cycles, so that comparative assessments can be made between these cycles, as well as with other transmutation and waste disposal concepts

  14. Fission products collecting devices

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi

    1979-01-01

    Purpose: To enable fission products trap with no contamination to coolants and cover gas by the provision of a fission products trap above the upper part of a nuclear power plant. Constitution: Upon fuel failures in a reactor core, nuclear fission products leak into coolants and move along the flow of the coolants to the coolants above the reactor core. The fission products are collected in a trap container and guided along a pipeline into fission products detector. The fission products detector monitors the concentration of the fission products and opens the downstream valve of the detector when a predetermined concentration of the fission products is detected to introduce the fission products into a waste gas processing device and release them through the exhaust pipe. (Seki, T.)

  15. Helium and fission gas behaviour in magnesium aluminate spinel and zirconia for actinide transmutation

    NARCIS (Netherlands)

    Damen, P.M.G.

    2003-01-01

    In order to reduce the long-term radiotoxicity of spent nuclear fuel, many studies are performed on partitioning and transmutation of actinides. In such a scenario, the long-lived radio-isotopes (mostly actinides) are partitioned from the nuclear waste, and subsequently transmuted or fissioned in a

  16. HAMCIND, Cell Burnup with Fission Products Poisoning

    International Nuclear Information System (INIS)

    Abe, Alfredo Y.; Dos Santos, Adimir

    2002-01-01

    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

  17. ORIGEN-S: scale system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    ORIGEN-S computes time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, input feet rates and physical or chemical removal rates. The calculations may pertain to fuel irradiation within nuclear reactors, or the storage, management, transportation or subsequent chemical processing of removed fuel elements. The matrix exponential expansion model of the ORIGIN code is unaltered in ORIGEN-S. Essentially all features of ORIGEN were retained, expanded or supplemented within new computations. The primary objective of ORIGEN-S, as requested by the Nuclear Regulatory Commission, is that the calculations may utilize the multi-energy group cross sections from any currently processed standardized ENDF/B data base. This purpose has been implemented through the prior execution of codes within either the SCALE System or the AMPX System, developed at the Oak Ridge National Laboratory. These codes compute flux-weighted cross sections, simulating conditions within any given reactor fuel assembly, and convert the data into a library that can be input to ORIGEN-S. Time-dependent libraries may be produced, reflecting fuel composition variations during irradiation. Presented in the document are: detailed and condensed input instructions, model theory, features available, range of applicability, brief subroutine descriptions, sample input, and I/O requirements. Presently the code is operable on IBM 360/370 computers and may be converted for CDC computers. ORIGEN-S is a functional module in the SCALE System and will be one of the modules invoked in the SAS2 Control Module, presently being developed, or may be applied as a stand alone program. It can be used in nuclear reactor and processing plant design studies, radiation safety analyses, and environmental assessments

  18. Research of the Mass Spectra of the Fission Products and Yields of (n, gamma) and (n, 2n) Reactions in a Model Subcritical Uranium Blanket of the Electronuclear System "Energy Plus Transmutation" on Proton Beam of the Dubna Synchrophasotron at 1.5 Ge

    CERN Document Server

    Chultem, D; Krivopustov, M I; Gerbish, S; Tumendemberel, B; Pavlyuk, A B; Zaveryukha, O S

    2002-01-01

    This paper is devoted to the research of the spatial distributions of the yields of (n, f), (n, gamma) and (n, 2n) reactions in a two-section model of the uranium blanket electronuclear installation constructed at the Laboratory of High Energies, JINR (Dubna) for experiments according to the program "Research of physical aspects of the electronuclear method of energy production and of radioactive waste transmutation in atomic power-engineering on beams of the synchrophasotron and nuclotron" - project "Energy plus Transmutation". The mass spectrum of the fission products and yields of above reactions in uranium activation detectors placed on the radii of the so-called detector plates is determined. The experimental results testify that the fission of nuclei in the uranium blanket is made by fast neutrons. This conclusion coincides with the result obtained with track integrators of uranium fission.

  19. Special scientific programme on use of high energy accelerators for transmutation of actinides and power production

    International Nuclear Information System (INIS)

    1994-09-01

    Various techniques for the transmutation of radioactive waste through the use of high energy accelerators are reviewed and discussed. In particular, the present publication contains presentations on (i) requirements and the technical possibilities for the transmutation of long-lived radionuclides (background paper); (ii) high energy particle accelerators for bulk transformation of elements and energy generation; (iii) the resolution of nuclear energy issues using accelerator-driven technology; (iv) the use of proton accelerators for the transmutation of actinides and power production; (v) the coupling of an accelerator to a subcritical fission reactor (with a view on its potential impact on waste transmutation); (vi) research and development of accelerator-based transmutation technology at JAERI (Japan); and (vii) questions and problems with regard to accelerator-driven nuclear power and transmutation facilities. Refs, figs and tabs

  20. Preliminary analyses of neutronics schemes for three kinds waste transmutation blankets of fusion-fission hybrid

    International Nuclear Information System (INIS)

    Zhang Mingchun; Feng Kaiming; Li Zaixin; Zhao Fengchao

    2012-01-01

    The neutronics schemes of the helium-cooled waste transmutation blanket, sodium-cooled waste transmutation blanket and FLiBe-cooled waste transmutation blanket were preliminarily calculated and analysed by using the spheroidal tokamak (ST) plasma configuration. The neutronics properties of these blankets' were compared and analyzed. The results show that for the transmutation of "2"3"7Np, FLiBe-cooled waste transmutation blanket has the most superior transmutation performance. The calculation results of the helium-cooled waste transmutation blanket show that this transmutation blanket can run on a steady effective multiplication factor (k_e_f_f), steady power (P), and steady tritium production rate (TBR) state for a long operating time (9.62 years) by change "2"3"7Np's initial loading rate of the minor actinides (MA). (authors)

  1. Fission product yields

    International Nuclear Information System (INIS)

    Valenta, V.; Hep, J.

    1978-01-01

    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 235 U, 239 Pu, 241 Pu and for fast fission (approximately 1 MeV) of 235 U, 238 U, 239 Pu, 241 Pu; these values are included into the 5th version of the YIELDS library, supplementing the BIBFP library. A comparison is made of experimental data and possible improvements of calculational methods are suggested. (author)

  2. Transmutation and neutron flux studies with fission chambers in the MEGAPIE target

    International Nuclear Information System (INIS)

    Chabod, S.; Foucher, Y.; Letourneau, A.; Marie, F.; Toussaint, J.C.; Blandin, Ch.; Chartier, F.; Fioni, G.

    2005-01-01

    Eight fission micro chambers will be inserted inside the central rod of the 1 MW liquid Pb-Bi MEGAPIE target in order to study the transmutation of two major actinides and to measure the neutron flux at a level of 5%. These chambers were developed for high neutron fluxes and tested at Laue Langevin Institute. (authors)

  3. Reduction of the radio-toxicity of long-lived nuclear wastes. Theoretical and strategic studies of the transmutation of minor actinides and fission products in electronuclear reactors

    International Nuclear Information System (INIS)

    Sala, Stephanie

    1995-01-01

    The first objective of this research thesis is to establish an assessment of the current situation regarding long-lived nuclear wastes: their identification, quantities produced in electronuclear reactors and during the fuel cycle, and their toxicity on the long term (until millions of years). The author then proposes a synthesis of possible solutions of management, particularly the solution based on separation and transmutation of these materials in electronuclear reactors, as well as of the consequences on the core and fuel cycle parameters. An application study is performed on an electronuclear fleet which, according to different scenarios, may comprise Pressurised Water Reactors and Fast Breeder Reactors, in order to assess the opportunities and constraints of such a solution, as well as expected benefits on assessments regarding materials, activity and radio-toxicity on the long term while taking present technologies into account [fr

  4. Power reactors and sub-critical blanket systems with lead and lead-bismuth as coolant and/or target material. Utilization and transmutation of actinides and long lived fission products

    International Nuclear Information System (INIS)

    2003-05-01

    High level radioactive waste disposal is an issue of great importance in the discussion of the sustainability of nuclear power generation. The main contributors to the high radioactivity are the fission products and the minor actinides. The long lived fission products and minor actinides set severe demands on the arrangements for safe waste disposal. Fast reactors and accelerator driven systems (ADS) are under development in Member States to reduce the long term hazard of spent fuel and radioactive waste, taking advantage of their incineration and transmutation capability. Important R and D programmes are being undertaken in many Member States to substantiate this option and advance the basic knowledge in this innovative area of nuclear energy development. The conceptual design of the lead cooled fast reactor concept BREST-OD-300, as well as various other conceptual designs of lead/lead-bismuth cooled fast reactors have been developed to meet enhanced safety and non-proliferation requirements, aiming at both energy production and transmutation of nuclear waste. Some R and D studies indicate that the use of lead and lead-bismuth coolant has some advantages in comparison with existing sodium cooled fast reactor systems, e.g.: simplified design of fast reactor core and BOP, enhanced inherent safety, and easier radwaste management in related fuel cycles. Moreover, various ADS conceptual designs with lead and lead-bismuth as target material and coolant also have been pursued. The results to date are encouraging, indicating that the ADS has the potential to offer an option for meeting the challenges of the back end fuel cycle. During the last decade, there have been substantial advances in several countries with their own R and D programme in the fields of lead/lead-bismuth cooled critical and sub-critical concepts. coolant technology, and experimental validation. In this context, international exchange of information and experience, as well as international

  5. Transmutation of radioactive wastes: how and why?

    International Nuclear Information System (INIS)

    Patarin, L.

    2004-01-01

    After having evoked the natural or spontaneous transmutation of natural or artificial radioactive atoms, the author describes how this transmutation is technically obtained, indicates the two main families of atoms present in a used nuclear fuel and for which transmutation is to be investigated (long-lived fission residues or products, and transuranium elements) and of which the behaviour in neutron fluxes must be explored. He discusses the industrial means required for artificial transmutation. He discusses the interest of performing such a transmutation

  6. Fission product detection

    International Nuclear Information System (INIS)

    Liatard, E.; Akrouf, S.; Bruandet, J.F

    1987-01-01

    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 [fr

  7. Ability of Accelerator-Driven Systems (ADS) to Transmute Long Lived Fission Fragments

    International Nuclear Information System (INIS)

    Nguyen Mong Giao; Nguyen Thi Ai Thu; Tu Thanh Danh; Tran Thanh Dung; Huynh, Thi Kim Chi

    2010-12-01

    This paper presents the research results of the possibility to transmute the long-lived radioactive isotopes into stable or short-lived, mainly the long-lived fission fragments as 99 Tc, 127 I, 129 I, 181 Ta, 107 Ag, 109 Ag by accelerator-driven systems. We use semi-empirical formulas to establish our calculating code with the support of computer programs. (author)

  8. Transmutation of Isotopes --- Ecological and Energy Production Aspects

    Science.gov (United States)

    Gudowski, Waclaw

    2000-01-01

    This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional

  9. Fission Product Library and Resource

    Energy Technology Data Exchange (ETDEWEB)

    Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Padgett, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-09-29

    Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

  10. Fusion-Fission Transmutation Scheme-Efficient destruction of nuclear waste

    International Nuclear Information System (INIS)

    Kotschenreuther, M.; Valanju, P.M.; Mahajan, S.M.; Schneider, E.A.

    2009-01-01

    A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion-fission hybrid assembly uniquely equipped to burn the worst thermal nonfissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The center piece of this fuel cycle, the high power density compact fusion neutron source (100 MW, outer radius <3 m), is made possible by a new divertor with a heat-handling capacity five times that of the standard alternative. The number of hybrids needed to destroy a given amount of waste is an order of magnitude below the corresponding number of critical fast-spectrum reactors (FRs) as the latter cannot fully exploit the new fuel cycle. Also, the time needed for 99% transuranic waste destruction reduces from centuries (with FR) to decades

  11. Fission neutron irradiation of copper containing implanted and transmutation produced helium

    DEFF Research Database (Denmark)

    Singh, B.N.; Horsewell, A.; Eldrup, Morten Mostgaard

    1992-01-01

    High purity copper containing approximately 100 appm helium was produced in two ways. In the first, helium was implanted by cyclotron at Harwell at 323 K. In the second method, helium was produced as a transmutation product in 800 MeV proton irradiation at Los Alamos, also at 323 K. The distribut......High purity copper containing approximately 100 appm helium was produced in two ways. In the first, helium was implanted by cyclotron at Harwell at 323 K. In the second method, helium was produced as a transmutation product in 800 MeV proton irradiation at Los Alamos, also at 323 K...... as well as the effect of the presence of other transmutation produced impurity atoms in the 800 MeV proton irradiated copper will be discussed....

  12. Accelerator technology for Los Alamos nuclear-waste-transmutation and energy-production concepts

    International Nuclear Information System (INIS)

    Lawrence, G.P.; Jameson, R.A.; Schriber, S.O.

    1991-01-01

    Powerful proton linacs are being studied at Los Alamos as drivers for high-flux neutron sources that can transmute long-lived fission products and actinides in defense nuclear waste, and also as drivers of advanced fission-energy systems that could generate electric power with no long-term waste legacy. A transmuter fed by an 800-MeV, 140-mA cw conventional copper linac could destroy the accumulated 99 Tc and 129 I at the DOE's Hanford site within 30 years. A high-efficiency 1200-MeV, 140-mA niobium superconducting linac could drive an energy-producing system generating 1-GWe electric power. Preliminary design concepts for these different high-power linacs are discussed, along with the principal technical issues and the status of the technology base. 9 refs., 5 figs., 4 tabs

  13. Transmutation Capability of a Once-Through Molten-Salt and Other Transmuting Reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Lowenthal, M.; Barnes, D.; Kawasaki, D.; Kimball, D.; Matsumoto, H.; Sagara, H.; Vietez, E.R.

    2002-01-01

    A preliminary assessment is done of the transmutation characteristics of three reactor technologies: a multi-batch liquid metal (LM) cooled transmuter, a once-through molten-salt (MS) transmuter and a pebble bed (PB) transmuter. It was found that for the same fractional transmutation and same k eff drop with burnup (Δk effBU ), lead-bismuth offers smaller peak-to-average core power density, and it requires a smaller pumping power but a larger and heavier core than a sodium cooled transmuter. 99 Tc cannot effectively serve as a burnable absorber to reduce Δk effBU of LM transmuters. However, addition of thorium can greatly flatten k eff and almost double the fractional transmutation of the LWR spent fuel from ∼20% to ∼40%. If the 'once-through' MS transmuter is operated with continuous complete removal of fission products, it can achieve ∼85% fractional transmutation provided that the equilibrium concentration of actinides in the MS can reach 4 mole %. If the fission products are not actively removed, the fractional transmutation is reduced to ∼75%. The fractional transmutation of a PB transmuter can exceed 40%. More thorough analysis is required to better quantify the transmutation capability of the different transmuter technologies. (authors)

  14. Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

    Science.gov (United States)

    Bowman, Charles D.

    1992-01-01

    Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

  15. Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

    Science.gov (United States)

    Bowman, C.D.

    1992-11-03

    Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

  16. Utilization of fast reactor excess neutrons for burning long-lived fission products

    International Nuclear Information System (INIS)

    Kawashima, K.; Kobayashi, K.; Kaneto, K.

    1995-01-01

    An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

  17. Nuclear transmutations

    International Nuclear Information System (INIS)

    Mikulaj, V.

    1992-01-01

    Two types of nuclear transmutations are outlined, namely the radioactive transmutations and nuclear reactions. The basic characteristics are given of radioactive transmutations (gamma transmutations and isomeric transitions, beta, alpha transmutations, spontaneous fission and spontaneous emission of nucleons), their kinetics and the influence of the physical and chemical state of the radionuclide on the transmutation rate. The basic characteristics are described of nuclear reactions (reactions of neutrons including fission, reactions induced by charged particles and photons), their kinetics, effective cross sections and their mechanism. Chemical reactions caused by nuclear transmutations are discussed (recoil energy, properties of hot atoms, Szilard-Chalmers effect). A brief information is given on the behavior of radionuclides in trace concentrations. (Z.S.) 2 tabs., 19 figs., 12 refs

  18. Fission-product source terms

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1981-01-01

    This presentation consists of a review of fission-product source terms for light water reactor (LWR) fuel. A source term is the quantity of fission products released under specified conditions that can be used to calculate the consequences of the release. The source term usually defines release from breached fuel-rod cladding but could also describe release from the primary coolant system, the reactor containment shell, or the site boundary. The source term would be different for each locality, and the chemical and physical forms of the fission products could also differ

  19. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1984-09-01

    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

  20. Extraction chromatography of fission products

    International Nuclear Information System (INIS)

    Bonnevie-Svendsen, M.; Goon, K.

    1978-01-01

    Various cases of using extraction chromatography during analysis of fission products are reviewed. The use of the extraction chromatography method is considered while analysing reprocessed products of nuclear fuel for quantitative radiochemical analysis and control of fission product and actinoide separation during extraction and their chemical state in production solutions. The method is used to obtain pure fractions of typical burnup monitors (neodymium, molybdenum, cerium, cesium, europium, lanthanides) during determination of nuclear fuel burnup degree. While studying the nature of nuclear reactions the method is used to separate quickly short-life isotopes, to purify β-radiator fractions before measuring their half-life periods, to enrich isotopes forming with low output during fission. Examples of using extraction chromatography are given to separate long half-life or stable fission products from spent solutions, to control environment object contamination

  1. Chemical Production using Fission Fragments

    International Nuclear Information System (INIS)

    Dawson, J. K.; Moseley, F.

    1960-01-01

    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) [fr

  2. Transmutation Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    Song, T. Y.; Park, W. S.; Kim, Y. H. (and others)

    2007-06-15

    The spent fuel coming from the PWR is one of the most difficult problems to be solved for the continuous use of nuclear power. It takes a few million years to be safe under the ground. Therefore, it is not easy to take care of the spent fuel for such a long time. Transmutation technology is the key technology which can solve the spent fuel problem basically. Transmutation is to transmute long-lived radioactive nuclides in the spent fuel into short-lived or stable nuclide through nuclear reactions. The long-lived radioactive nuclides can be TRU and fission products such as Tc-99 and I-129. Although the transmutation technology does not make the underground disposal totally unnecessary, the period to take care of the spent fuel can be reduced to the order of a few hundred years. In addition to the environmental benefit, transmutation can be considered to recycle the energy in the spent fuel since the transmutation is performed through nuclear fission reaction of the TRU in the spent fuel. Therefore, transmutation technology is worth being developed in economical aspect. The results of this work can be a basis for the next stage research. The objective of the third stage research was to complete the core conceptual design and verification of the key technologies. The final results will contribute to the establishment of Korean back end fuel cycle policy by providing technical guidelines.

  3. A Los Alamos concept for accelerator transmutation of waste and energy production (ATW)

    International Nuclear Information System (INIS)

    1990-01-01

    This document contains the diagrams presented at the ATW (Accelerator Transmutation of Waste and Energy Production) External Review, December 10-12, 1990, held at Los Alamos National Laboratory. Included are the charge to the committee and the presentations for the committee's review. Topics of the presentations included an overview of the concept, LINAC technology, near-term application -- high-level defense wastes (intense thermal neutron source, chemistry and materials), advanced application of the ATW concept -- fission energy without a high-level waste stream (overview, advanced technology, and advanced chemistry), and a summary of the research issues

  4. [Fission product yields of 60 fissioning reactions]. Final report

    International Nuclear Information System (INIS)

    Rider, B.F.

    1995-01-01

    In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ''Evaluation and Compilation of Fission Product Yields 1993,'' LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

  5. Fission 99Mo production technology

    International Nuclear Information System (INIS)

    Miao Zengxing; Luo Zhifu; Ma Huimin; Liang Yufu; Yu Ningwen

    2003-01-01

    This paper describes a production technology of fission 99 Mo in the Department Isotope, CIAE. The irradiation target is tubular U-Al alloy containing highly enriched uranium. The target is irradiated in the swimming pool reactor core. The neutron flux is about 4x10 13 /cm 2 .sec. The production scale is 3.7-7.4 TBq (100-200Ci) of fission 99 Mo per batch. Total recovery of 99 Mo is more than 70%. The production practice proves that the process and equipment are safe and reliable. (author)

  6. PKA distributions: Contributions from transmutation products and from radioactive decay

    Directory of Open Access Journals (Sweden)

    M.R. Gilbert

    2016-12-01

    Full Text Available The neutrons generated in fusion plasmas interact with materials via nuclear reactions. The resulting transmutations and atomic displacements have life-limiting consequences for fusion reactor components. A detailed understanding of the production, evolution and material consequences of the damage created by cascades of atomic displacements requires, as a vital primary input, a complete description of the energy-spectrum of initial (prompt atomic displacement events (the primary knock on atoms or PKAs produced by direct neutron nuclear interactions. There is also the possibility that the radionuclides produced under transmutation will create further PKAs as they decay, and so the rate of these must also be quantified. This paper presents the latest results from the analysis of PKA spectra under neutron irradiation, focussing particularly on the variation in PKA distributions due to changes in composition under transmutation, but also on the PKA contributions from radioactive decay of materials that become activated under irradiation.

  7. NEACRP thermal fission product benchmark

    International Nuclear Information System (INIS)

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

    1989-09-01

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

  8. Influence of transmutation and high neutron exposure on materials used in fission-fusion correlation experiments

    International Nuclear Information System (INIS)

    Garner, F.A.

    1990-07-01

    This paper explores the response of three different materials to high fluence irradiation as observed in recent fusion-related experiments. While helium at fusion-relevant levels influences the details of the microstructure of Fe--Cr--Ni alloys somewhat, the resultant changes in swelling and tensile behavior are relatively small. Under conditions where substantially greater-than-fusion levels of helium are generated, however, an extensive refinement of microstructure can occur, leading to depression of swelling at lower temperatures and increased strengthening at all temperatures studied. The behavior of these alloys is dominated by their tendency to converge to saturation microstructures which encourage swelling. Irradiations of nickel are dominated by its tendency to develop a different type of saturation microstructure that discourages further void growth. Swelling approaches saturation levels that are remarkably insensitive to starting microstructure and irradiation temperature. The rate of approach to saturation is very sensitive to variables such as helium, impurities, dislocation density and displacement rate, however. Copper exhibits a rather divergent response depending on the property measured. Transmutation of copper to nickel and zinc plays a large role in determining electrical conductivity but almost no role in void swelling. Each of these three materials offers different challenges in the interpretation of fission-fusion correlation experiments

  9. Transmutation detectors

    Energy Technology Data Exchange (ETDEWEB)

    Viererbl, L., E-mail: vie@ujv.c [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Lahodova, Z. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Klupak, V. [Nuclear Research Institute Rez plc (Czech Republic); Sus, F. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Kucera, J. [Research Centre Rez Ltd. (Czech Republic); Nuclear Physics Institute, Academy of Sciences of the Czech Republic (Czech Republic); Kus, P.; Marek, M. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic)

    2011-03-11

    We have designed a new type of detectors, called transmutation detectors, which can be used primarily for neutron fluence measurement. The transmutation detector method differs from the commonly used activation detector method in evaluation of detector response after irradiation. Instead of radionuclide activity measurement using radiometric methods, the concentration of stable non-gaseous nuclides generated by transmutation in the detector is measured using analytical methods like mass spectrometry. Prospective elements and nuclear reactions for transmutation detectors are listed and initial experimental results are given. The transmutation detector method could be used primarily for long-term measurement of neutron fluence in fission nuclear reactors, but in principle it could be used for any type of radiation that can cause transmutation of nuclides in detectors. This method could also be used for measurement in accelerators or fusion reactors.

  10. Transmutation detectors

    International Nuclear Information System (INIS)

    Viererbl, L.; Lahodova, Z.; Klupak, V.; Sus, F.; Kucera, J.; Kus, P.; Marek, M.

    2011-01-01

    We have designed a new type of detectors, called transmutation detectors, which can be used primarily for neutron fluence measurement. The transmutation detector method differs from the commonly used activation detector method in evaluation of detector response after irradiation. Instead of radionuclide activity measurement using radiometric methods, the concentration of stable non-gaseous nuclides generated by transmutation in the detector is measured using analytical methods like mass spectrometry. Prospective elements and nuclear reactions for transmutation detectors are listed and initial experimental results are given. The transmutation detector method could be used primarily for long-term measurement of neutron fluence in fission nuclear reactors, but in principle it could be used for any type of radiation that can cause transmutation of nuclides in detectors. This method could also be used for measurement in accelerators or fusion reactors.

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

    Science.gov (United States)

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

    2014-09-01

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

  12. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, G.

    1976-05-01

    The purpose of this series is to inform scientists working on Fission Product Nuclear Data, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. This report consists of reproductions of essentially unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat

  13. User's manual for computer code RIBD-II, a fission product inventory code

    International Nuclear Information System (INIS)

    Marr, D.R.

    1975-01-01

    The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)

  14. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Auvinen, A.; Maekynen, J.; Valmari, T.

    1998-01-01

    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

  15. Device for measuring fission product density

    International Nuclear Information System (INIS)

    Kaneda, Mitsunori.

    1980-01-01

    Purpose: To determine the fission product density of xenon or the like and enable measurement of real time of fission product density in a reactor by calculating the disintegration and annihilation of the fission product on the basis of neutron detected output. Constitution: The neutron flux in a reactor is detected by a detector, and applied to first and second density calculators. Second fission product density signal of xenon or the like outputted from first device is again inputted to the device to form an annihilation signal due to disintegration to determine the present density of the second fission product of xenon or the like corresponding to the decrease of the neutron due to the poison of xeron or the like. Similarly, second device determines the first fission product density of iodine or the like. (Sekiya, K.)

  16. Transmutation of radioactive nuclear waste – present status and ...

    Indian Academy of Sciences (India)

    Transmutation of long-lived actinides and fission products becomes an im- ... Similar approach was performed for sub critical fast reactor core with Pu/MA .... The same might be addressed to masses of nuclei (the use of experimental values.

  17. Attachment of gaseous fission products to aerosols

    International Nuclear Information System (INIS)

    Skyrme, G.

    1985-01-01

    Accidents may occur in which the integrity of fuel cladding is breached and volatile fission products are released to the containment atmosphere. In order to assess the magnitude of the subsequent radiological hazard it is necessary to know the transport behaviour of such fission products. It is frequently assumed that the fission products remain in the gaseous phase. There is a possibility, however, that they may attach themselves to particles and hence substantially modify their transport properties. This paper provides a theoretical assessment of the conditions under which gaseous fission products may be attached to aerosol particles. Specific topics discussed are: the mass transfer of a gaseous fission product to an isolated aerosol particle in an infinite medium; the rate at which the concentration of fission products in the gas phase diminishes within a container as a result of deposition on a population of particles; and the distribution of deposited fission product between different particle sizes in a log-normal distribution. It is shown that, for a given mass, small particles are more efficient for fission product attachment, and that only small concentrations of such particles may be necessary to achieve rapid attachment. Conditions under which gaseous fission products are not attached to particles are also considered, viz, the competing processes of deposition onto the containment walls and onto aerosol particles, and the possibility of the removal of aerosols from the containment by various deposition processes, or agglomeration, before attachment takes place. (author)

  18. Process for the extraction of fission products

    International Nuclear Information System (INIS)

    Anav, M.; Chesne, A.; Leseur, A.; Miquel, P.; Pascard, R.

    1979-01-01

    A process is described for the extraction of fission products contained in irradiated nuclear fuel elements which have been subject to a temperature of at least 1200 0 C during their irradiation prior to dissolving the fuel by the wet process. After mechanically treating the elements in order to decan and/or cut them they are brought into contact with water in order to pass the fission products into aqueous solution. The treated elements are then separated from the thus obtained aqueous solution. At least one of the fission products is then recovered from the aqueous solution. The fission products are iodine, cesium, rubidium and tritium

  19. The different facilities of the reactor Phenix for radio isotope production and fission product burner

    International Nuclear Information System (INIS)

    Coulon, P.; Clerc, R.; Tommasi, J.

    1993-01-01

    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). (authors). 9 figs

  20. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  1. Downstream behavior of fission products

    International Nuclear Information System (INIS)

    Johnson, I.; Farahat, M.K.; Settle, J.L.; Johnson, C.E.; Ritzman, R.

    1986-01-01

    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 (1000 0 to 200 0 C) 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

  2. Aerosols and fission product transport

    International Nuclear Information System (INIS)

    Megaw, W.J.

    1987-12-01

    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

  3. Nuclear transmutation by flux compression

    International Nuclear Information System (INIS)

    Seifritz, W.

    2001-01-01

    A new idea for the transmutation of minor actinides, long (and even short) lived fission products is presented. It is based an the property of neutron flux compression in nuclear (fast and/or thermal) reactors possessing spatially non-stationary critical masses. An advantage factor for the burn-up fluence of the elements to be transmuted in the order of magnitude of 100 and more is obtainable compared with the classical way of transmutation. Three typical examples of such transmuters (a subcritical ringreactor with a rotating reflector, a sub-critical ring reactor with a rotating spallation source, the socalled ''pulsed energy amplifier'', and a fast burn-wave reactor) are presented and analysed with regard to this purpose. (orig.) [de

  4. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, G.

    1975-01-01

    This is the first issue of a report series on Fission Product Nuclear Data (FPND), published every six months by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). Its purpose 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 cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat. The present issue includes contributions which were received by NDS before 1 November 1975

  5. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1981-06-01

    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)

    Lammer, M.

    1982-07-01

    This is the eighth 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. Therefore, the IAEA cannot be held responsible for the information contained nor for any consequences resulting from the use of this information. 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 seventh issue of this series has been published in July 1981 as INDC(NDS)-116. The present issue includes contributions which were received by NDS between 1 August 1981 and 15 June 1982

  7. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1983-08-01

    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

  8. Estimation of the outlooks for large-scale transmutation of fission-produced iodine

    CERN Document Server

    Galkin, B Y; Kolyadin, A B; Kocherov, N P; Lyubtsev, R I; Hosov, A A; Rimskij-Korsakov, A A

    2002-01-01

    To obtain data necessary for estimating sup 1 sup 2 sup 9 I transmutation efficiency in nuclear reactors the effective neutron capture cross section on sup 1 sup 2 sup 9 I isotope in neutral spectrum of the WWR-M reactor was determined. The calculated value of sup 1 sup 2 sup 9 I capture cross section, averaged by neutron spectrum in beryllium reflector of the WWR-M reactor, made up 17.8+-3.2 barn. On the basis of experimental data and estimations it was shown that in neutron flux 10 sup 1 sup 4 1/(cm sup 2 s) transmutation of iodine -129 loaded in the course of one year can amount to approximately 25%

  9. New data libraries for transmutation studies

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Interfaculty Reactor Inst., Delft (Netherlands)

    1995-06-01

    The fuel depletion code ORIGEN-S is often used for transmutation studies. It uses three different working libraries for actinides, fission products, and light elements, which contain decay data, cross-section data and fission product yields. These data have been renewed with data based on the JEF2.2 and the EAF3 evaluated files. Furthermore, data for 201 fission products have been added to the libraries. The new data libraries are particular suitable for parameter studies and other introductory calculations. For more accurate calculations, it is advised to regularly update the cross sections of the most important actinides and fission products during the burnup sequence. (orig.).

  10. New data libraries for transmutation studies

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Interfaculty Reactor Inst., Delft (Netherlands)

    1995-12-31

    The fuel depletion code ORIGEN-S is often used for transmutations studies. It uses three different working libraries for actinides, fission products, and light elements, which contain decay data, cross-section data and fission product yields. These data have renewed with data based on the JEF2.2 and the EAF3 evaluated files. Furthermore, data for 201 fission products have been added to the libraries. The new data libraries are particular suitable for parameter studies and other introductory calculations. For more accurate calculations, it is advised to regularly update the cross sections of the most important actinides and fission products during the burnup sequence. (author) 9 refs.

  11. Separation of actinides and their transmutation

    International Nuclear Information System (INIS)

    Bouchard, M.; Bathelier, M.; Cousin, M.

    1978-08-01

    Neutron irradiation of long-half-life actinides for transmutation into elements with shorter half-life is investigated as a means to reduce the long-term hazards of these actinides. The effectiveness of the method is analysed by applying it to fission product solutions from the first extraction cycle of fuel reprocessing plants. Basic principles, separation techniques and transmutation efficiencies are studied and discussed in detail

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

    Science.gov (United States)

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

    1967-01-01

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

  13. Thermodynamic analysis of volatile organometallic fission products

    International Nuclear Information System (INIS)

    Auxier II, J.D.; Hall, H.L.; Cressy, Derek

    2016-01-01

    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)

  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. Development of fission Mo-99 production technology

    International Nuclear Information System (INIS)

    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

  16. Minor actinide transmutation in accelerator driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Friess, Friederike [IANUS, TU Darmstadt (Germany)

    2015-07-01

    Transmutation of radioactive waste, the legacy of nuclear energy use, gains rising interest. This includes the development of facilities able to transmute minor actinides (MA) into stable or short-lived isotopes before final disposal. The most common proposal is to use a double-strata approach with accelerator-driven-systems (ADS) for the efficient transmutation of MA and power reactors to dispose plutonium. An ADS consists of a sub-critical core that reaches criticality with neutrons supplied by a spallation target. An MCNP model of the ADS system Multi Purpose Research Reactor for Hightech Applications will be presented. Depletion calculations have been performed for both standard MOX fuel and transmutation fuel with an increased content of minor actinides. The resulting transmutation rates for MAs are compared to published values. Special attention is given to selected fission products such as Tc-99 and I-129, which impact the radiation from the spent fuel significantly.

  17. Reprocessing free nuclear fuel production via fusion fission hybrids

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike, E-mail: mtk@mail.utexas.edu [Intitute for Fusion Studies, University of Texas at Austin (United States); Valanju, Prashant; Mahajan, Swadesh [Intitute for Fusion Studies, University of Texas at Austin (United States)

    2012-05-15

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively 'new' cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th{sup 232}-U{sup 233} conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO{sub 2} matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U{sup 235} fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  18. Reprocessing free nuclear fuel production via fusion fission hybrids

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Mahajan, Swadesh

    2012-01-01

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th 232 –U 233 conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO 2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U 235 fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  19. Decay and Transmutation of Nuclides

    CERN Document Server

    Aarnio, Pertti A

    1999-01-01

    We present a computer code DeTra which solves analytically the Bateman equations governing the decay, build-up and transmutation of radionuclides. The complexity of the chains and the number of nuclides are not limited. The nuclide production terms considered include transmutation of the nuclides inside the chain, external production, and fission. Time dependent calculations are possible since all the production terms can be re-defined for each irradiation step. The number of irradiation steps and output times is unlimited. DeTra is thus able to solve any decay and transmutation problem as long as the nuclear data i.e. decay data and production rates, or cross sections, are known.

  20. Fission-product release during accidents

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Cox, D.S.

    1991-09-01

    One of the aims when managing a reactor accident is to minimize the release of radioactive fission products. Release is dependent not only on the temperature, but also on the partial pressure of oxygen. Strongly oxidizing atmospheres, such as those that occurred during the Chernobyl accident, released semi-volatile elements like ruthenium, which has volatile oxides. At low temperatures, UO 2 oxidization to U 3 O 8 can result in extensive breakup of the fuel, resulting in the release of non-volatile fission products as aerosols. Under less oxidizing conditions, when hydrogen accumulates from the zirconium-water reaction, the resulting low oxygen partial pressure can significantly reduce these reactions. At TMI-2, only the noble gases and volatile fission products were released in significant quantities. A knowledge of the effect of atmosphere as well as temperature on the release of fission products from damaged reactor cores is therefore a useful, if not necessary, component of information required for accident management

  1. Vitrification processes for fission product solutions

    International Nuclear Information System (INIS)

    Bonniaud, R.; Jouan, A.; Moncouyoux, J.P.; Sombret, C.

    1982-10-01

    The different processes for fission product vitrification in the world are reviewed. Continuous or discontinuous processes, induction or arc heating, in can melting or casting, tests with radioactive or simulated wastes and industrial realizations are described [fr

  2. Systematics of Fission-Product Yields

    International Nuclear Information System (INIS)

    Wahl, A.C.

    2002-01-01

    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 Z F = 90 thru 98, mass number A F = 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

  3. Systematics of Fission-Product Yields

    Energy Technology Data Exchange (ETDEWEB)

    A.C. Wahl

    2002-05-01

    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 Z{sub F} = 90 thru 98, mass number A{sub F} = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru {approx}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 {approx} 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 ({approx} fission spectrum) induced fission reactions.

  4. Transport properties of fission product vapors

    International Nuclear Information System (INIS)

    Im, K.H.; Ahluwalia, R.K.

    1983-07-01

    Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors

  5. Modeling Fission Product Sorption in Graphite Structures

    International Nuclear Information System (INIS)

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-01-01

    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 products

  6. Nuclear waste transmutation

    International Nuclear Information System (INIS)

    Salvatores, M.; Girard, C.; Delpech, M.; Slessarev, I.; Tommasi, J.

    1994-01-01

    Waste management strategies foresee the use of a deep geological repository either for final disposal of irradiated fuel or, after reprocessing and reuse of U and Pu for final disposal of long-lived radio-active materials. In the second case, partitioning and transmutation of these materials can be considered to reduce the impact of radiation on man due to the storage. On the basis of the SPIN programme developed by CEA in this field, the main features of transmutation is presented. The goal to achieve and the criteria to use are quite difficult to establish. The rights para-meters to characterize the risk are the potential radiotoxicity in the the repository and the residual radiotoxicity at the outlet. Transmutation studies in CEA used the potential radiotoxicity which is based on well-known parameters and less precise hazardous factors. The second point to appreciate the trans- mutation interest is to dispose of a criteria for the radio-radiotoxicity reduction. As there is no general agreement, we try to have a toxicity as low as possible within reasonable technical limits. To reduce the long term radio- toxicity, Pu, minor actinides and some long-lived fission products have to be transmuted. To assess the feasibility of such trans-mutation in reactors or advanced systems, one has to consider constraints on neutronic balance, safety, fuel cycle, technology , economy. Taking in account the main conclusions of this analysis, parametric studies of homogeneous and heterogenous transmutation permit a choice of promising solutions. Goals are to use every long-lived element with a minimized production of other long- lived elements in order to obtain an appreciable radiotoxicity reduction. It implies multi recycling of Pu which favours fast neutron reactors and different strategies of multi recycling for Np, Am, Cm. Multi recycling makes the results strongly dependant of losses. Researches to obtain the high partitioning efficiency needed are in progress. Calculations

  7. Recovery and use of fission product noble metals

    International Nuclear Information System (INIS)

    Jensen, G.A.; Rohmann, C.A.; Perrigo, L.D.

    1980-06-01

    Noble metals in fission products are of strategic value. Market prices for noble metals are rising more rapidly than recovery costs. A promising concept has been developed for recovery of noble metals from fission product waste. Although the assessment was made only for the three noble metal fission products (Rh, Pd, Ru), there are other fission products and actinides which have potential value

  8. Fission products in glasses. Pt. 2

    International Nuclear Information System (INIS)

    De, A.K.; Luckscheiter, B.; Malow, G.; Schiewer, E.

    1977-09-01

    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) [de

  9. Dual neutral particle induced transmutation in CINDER2008

    Energy Technology Data Exchange (ETDEWEB)

    Martin, W.J., E-mail: wjmarti@sandia.gov [Sandia National Laboratories, Albuquerque, NM 87185 (United States); University of New Mexico, Albuquerque, NM 87131 (United States); Oliveira, C.R.E. de; Hecht, A.A. [University of New Mexico, Albuquerque, NM 87131 (United States)

    2014-12-11

    Although nuclear transmutation methods for fission have existed for decades, the focus has been on neutron-induced reactions. Recent novel concepts have sought to use both neutrons and photons for purposes such as active interrogation of cargo to detect the smuggling of highly enriched uranium, a concept that would require modeling the transmutation caused by both incident particles. As photonuclear transmutation has yet to be modeled alongside neutron-induced transmutation in a production code, new methods need to be developed. The CINDER2008 nuclear transmutation code from Los Alamos National Laboratory is extended from neutron applications to dual neutral particle applications, allowing both neutron- and photon-induced reactions for this modeling with a focus on fission. Following standard reaction modeling, the induced fission reaction is understood as a two-part reaction, with an entrance channel to the excited compound nucleus, and an exit channel from the excited compound nucleus to the fission fragmentation. Because photofission yield data—the exit channel from the compound nucleus—are sparse, neutron fission yield data are used in this work. With a different compound nucleus and excitation, the translation to the excited compound state is modified, as appropriate. A verification and validation of these methods and data has been performed. This has shown that the translation of neutron-induced fission product yield sets, and their use in photonuclear applications, is appropriate, and that the code has been extended correctly. - Highlights: • The CINDER2008 transmutation code was modified to include photon-induced transmutation tracking. • A photonuclear interaction library was created to allow CINDER2008 to track photonuclear interactions. • Photofission product yield data sets were created using fission physics similarities with neutron-induced fission.

  10. JNDC nuclear data library of fission products

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Ihara, Hitoshi; Akiyama, Masatsugu; Yoshida, Tadashi; Matumoto, Zyun-itiro; Nakasima, Ryuzo

    1983-10-01

    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 10 3 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 10 3 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)

  11. Separation of short-lived fission products

    International Nuclear Information System (INIS)

    Tamai, Tadaharu; Ohyoshi, Emiko; Ohyoshi, Akira; Kiso, Yoshiyuki; Shinagawa, Mutsuaki.

    1976-01-01

    A rbief review is presented on the various methods of separation available for both gaseous and liquid states, for the separation of short-lived fission products formed by binary fission of neutron irradiated uranium. The means available for gaseous state are the hot atom reaction, the hydride method and on-line mass separation. For liquid state, use can be made of precipitation, ionic or atomic exchange, solvent extraction and paper electrophoresis. Particular reference is made to electrophoretic separation of ions produced by fission in aqueous solution of uranium. The principle of electrophoretic separation and the procedures for separating the element of interest from the other fission products are outlined, with reference made to the results obtained with the method by the present authors. The elements in question are alkalines, alkaline earths, rare earths, halogens, selenium and

  12. Correlation of recent fission product release data

    International Nuclear Information System (INIS)

    Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

    1989-01-01

    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

  13. 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

  14. Fuel morphology effects on fission product release

    International Nuclear Information System (INIS)

    Osetek, D.J.; Hartwell, J.K.; Cronenberg, A.W.

    1986-01-01

    Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO 2 fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid and slow cooldown, low and high (36 GWd/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product species and release rates of noble gas species are reported. Significant differences in release behavior are discussed between heatup and cooldown periods, low and high burnup fuel and long- and short-lived fission products. Explanations for the observed differences are offered that relate fuel morphology changes to the releases

  15. Fission product release mechanisms and groupings

    International Nuclear Information System (INIS)

    Iglesia, F.C.; Brito, A.C.; Liu, Y.

    1995-01-01

    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)

  16. Fission product release mechanisms and groupings

    Energy Technology Data Exchange (ETDEWEB)

    Iglesia, F C; Brito, A C; Liu, Y [Ontario Hydro, Toronto, ON (Canada); and others

    1996-12-31

    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) 92 refs., 6 tabs.

  17. Chemistry of fission products for accident analysis

    International Nuclear Information System (INIS)

    Potter, P.E.

    1985-01-01

    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

  18. Chemistry of actinides and fission products

    International Nuclear Information System (INIS)

    Pruett, D.J.; Sherrow, S.A.; Toth, L.M.

    1988-01-01

    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

  19. Regulatory simplification of fission product chemistry

    International Nuclear Information System (INIS)

    Read, J.B.J.; Soffer, L.

    1986-01-01

    The requirements for design provisions intended to limit fission product escape during reactor accidents have been based since 1962 upon a small number of simply-stated assumptions. These assumptions permeate current reactor regulation, but are too simple to deal with the complex processes that can reasonably be expected to occur during real accidents. Potential chemical processes of fission products in severe accidents are compared with existing plant safety features designed to minimize off-site consequences, and the possibility of a new set of simply-stated assumptions to replace the 1982 set is discussed

  20. Waste transmutation: perspectives

    International Nuclear Information System (INIS)

    Leray, S.

    1997-01-01

    After the introduction on the source and nature of nuclear waste, this lecture analyzes the different methods proposed to transmute long-lived isotopes into stable or short-lived isotopes. It is shown that direct methods (photonuclear reactions, spallation, muon catalyzed fusion) do not lead to a sufficient transmutation rate within a reasonable cost. Only the use of hybrid systems, fusion-fission or spallation-fission, can be foreseen. (author)

  1. Minor actinides transmutation performance in a fast reactor

    International Nuclear Information System (INIS)

    Takeda, Toshikazu

    2016-01-01

    Highlights: • A method for calculating MA transmutation for individual nuclides has been proposed by introducing two formulas of the MA transmutation. One corresponds to the difference of MA amounts, and the other corresponds to the sum of the fission amounts and the plutonium production amounts. • Using the method the MA transmutation was calculated for Np-237 and Am-241 in a fast reactor. The burnup period was changed from 1 year to 12 year. • For the 1 year burnup a large amount of Am-242m, Cm-242 are produced from Am-241. The total MA transmutation amount increases with burnup time, but its gradient with respect to burnup time decreases after 9 years, and the transmutation amount by overall fission increases almost linearly with burnup time. • However, after the 6 year burnup the fission contribution became large because of the large production of Pu isotopes from the original Am-241. • In addition to the homogeneous loading of the MA nuclides into the cores, a heterogeneous loading of Am-241 to the blanket region was considered. - Abstract: Results obtained in the project named “Study on Minor Actinides Transmutation using Monju Data”, which has been sponsored by the Ministry of Education, Culture, Sports, Science and Technology in Japan (MEXT) are described. In order to physically understand transmutation of individual MA nuclides in fast reactors, a new method was developed in which the MAs transmutation is interpreted by two formulas. One corresponds to the difference of individual MA nuclides amounts before and after a burnup period, and the other is the sum of amount of fission of a relevant MA nuclide and the net plutonium production from the MA nuclide during a burnup period. The method has been applied to two fast reactors with MA fuels loaded in cores homogeneously and in a blanket region heterogeneously. Numerical results of MA transmutation for the two reactors are shown.

  2. Safety characteristics of potential waste transmutation systems

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.

    1993-01-01

    For nuclear waste transmutation to alter significantly the need for geologic disposal of spent fuel from US Light-water reactors (LWRs), about 1.4% of the spent fuel (by mass) must be separated and transmuted. This includes the plutonium, the minor actinides, and four fission products: iodine. technetium, cesium and strontium. Regarding the actinides, fissioning of the plutonium, neptunium, americium, and curium generates a great deal of heat, so much so that most of the plutonium should be used to produce power. However, these actinides have some undesirable neutronic characteristics, and their utilization in reactors or subcritical (proton-accelerator) targets requires either a fast neutronic spectrum or a very high thermal-neutron flux. Transmutation of the fission products is generally by neutron capture, although this is difficult in the case of cesium and strontium. In this paper, various proposed means of transmuting the actinides and fission products are discussed, with the main focus being on the safety characteristics of each approach

  3. Dosimetric measurement of the disintegration rate of fission products

    International Nuclear Information System (INIS)

    Solymosi, J.; Nagy, L.G.; Zagyvai, P.

    1992-01-01

    Investigations on the disintegration rate of fission products of 238 U and 239 Pu are presented. The intensity of the β-and γ-radiation of fission products were measured continously in an interval of 1-1300 hours following the fission, offering the possibility for determining the general and specific characteristics of the individual fission products. A universal measuring procedure was elaborated for the rapid in situ determination of the dosimetric features of fission products, which is suitable for the accurate evaluation and prediction of external absorbed dose even in case of fission products of various origin and unknown composition. (author) 6 refs.; 7 figs.; 1 tab

  4. Analytical evaluation of fission product sensitivities

    International Nuclear Information System (INIS)

    Sola, A.

    1977-01-01

    Evaluating the concentration of a fission product produced in a reactor requires the knowledge of a fairly large number of variables. Sensitivity studies were made to ascertain the important variables. Analytical formulae were developed sufficiently simple to allow numerical computations. Some simplified formulas are also given and they are applied to the following isotopes: 80 Se, 82 Se, 81 Br, 82 Br, 82 Kr, 83 Kr, 84 Kr, 85 Kr, 86 Kr. Their sensitivities to capture cross sections, fission yields, ratios of activation cross sections, half-lives (during and after irradiation), branching ratios, as well as to the neutron flux and to the time are considered

  5. Model for fission-product calculations

    International Nuclear Information System (INIS)

    Smith, A.B.

    1984-01-01

    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

  6. 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.

  7. Development of fission Mo-99 production technology

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    2009-01-01

    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

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

    International Nuclear Information System (INIS)

    Macdonald, L.D.; Lewis, B.J.; Iglesias, F.C.

    1989-05-01

    A microstructure-dependent fission-gas release and swelling model (FREEDOM) has been developed for UO 2 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

  10. Study on multi-recycle transmutation of LLFP in light water reactor

    International Nuclear Information System (INIS)

    Setiawan, M.B.; Kitamoto, A.

    2001-01-01

    The effectiveness of transmutation for long-lived fission products (LLFP) in light water reactors (LWR), i.e. both BWR and PWR, considering the large capture cross-section of FPs in thermal region was evaluated. Calculation results of iodine and technetium transmutation in BWR and PWR suggested an effective use of BWR as compared to PWR. To obtain transmutation fraction [TF] of 30 to 40%, the irradiation period needed for 99 Tc transmutation was estimated as 10 to 15 years, and the period for 129 I transmutation was estimated as 30 to 40 years, respectively. The evaluations bring a new concept of multi-recycle LLFP transmutation using LWR TR (LWR for transmutation)

  11. Thermochromatographic investigations of fission product transport and chemistry

    International Nuclear Information System (INIS)

    Growcock, F.B.; Aronson, S.; Friedlander, M.; Skalyo, J. Jr.; Hosseini, A.; Taylor, R.D.

    1978-01-01

    A thermochromatographic technique has been developed to investigate the chemical states of fission products from irradiated fuel as well as in fission product simulation studies. Some recent work on iodine transport and on release of fission products from irradiated fuel kernels will be discussed

  12. Investigation of the feasibility of a small scale transmutation device

    Science.gov (United States)

    Sit, Roger Carson

    This dissertation presents the design and feasibility of a small-scale, fusion-based transmutation device incorporating a commercially available neutron generator. It also presents the design features necessary to optimize the device and render it practical for the transmutation of selected long-lived fission products and actinides. Four conceptual designs of a transmutation device were used to study the transformation of seven radionuclides: long-lived fission products (Tc-99 and I-129), short-lived fission products (Cs-137 and Sr-90), and selective actinides (Am-241, Pu-238, and Pu-239). These radionuclides were chosen because they are major components of spent nuclear fuel and also because they exist as legacy sources that are being stored pending a decision regarding their ultimate disposition. The four designs include the use of two different devices; a Deuterium-Deuterium (D-D) neutron generator (for one design) and a Deuterium-Tritium (D-T) neutron generator (for three designs) in configurations which provide different neutron energy spectra for targeting the radionuclide for transmutation. Key parameters analyzed include total fluence and flux requirements; transmutation effectiveness measured as irradiation effective half-life; and activation products generated along with their characteristics: activity, dose rate, decay, and ingestion and inhalation radiotoxicity. From this investigation, conclusions were drawn about the feasibility of the device, the design and technology enhancements that would be required to make transmutation practical, the most beneficial design for each radionuclide, the consequence of the transmutation, and radiation protection issues that are important for the conceptual design of the transmutation device. Key conclusions from this investigation include: (1) the transmutation of long-lived fission products and select actinides can be practical using a small-scale, fusion driven transmutation device; (2) the transmutation of long

  13. Energy Production and Transmutation of Nuclear Waste by Accelerator Driven Systems

    Science.gov (United States)

    Zhivkov, P. K.

    2018-05-01

    There is a significant amount of highly radiotoxic long-life nuclear waste (NW) produced by NPP (Nuclear Power Plants). Transmutation is a process which transforms NW into less radiotoxic nuclides with a shorter period of half-life by spallation neutrons or radiative capture of neutrons produced by ADS (Accelerator Driven System). In the processes of transmutation new radioactive nuclides are produced. ADS is big energy consumer equipment. It is a method for production of a high-flux and high-energy neutron field. All these processes occur in ADS simultaneously. ADS is able to transmute actinides and produce energy simultaneously. The article considers the energy production problems in ADS. Several ideas are developed regarding the solution of the global energy supply.

  14. Transmutation of high-level radioactive waste - Perspectives

    CERN Document Server

    Junghans, Arnd; Grosse, Eckart; Hannaske, Roland; Kögler, Toni; Massarczyk, Ralf; Schwengner, Ronald; Wagner, Andreas

    2014-01-01

    In a fast neutron spectrum essentially all long-lived actinides (e.g. Plutonium) undergo fission and thus can be transmuted into generally short lived fission products. Innovative nuclear reactor concepts e.g. accelerator driven systems (ADS) are currently in development that foresee a closed fuel cycle. The majority of the fissile nuclides (uranium, plutonium) shall be used for power generation and only fission products will be put into final disposal that needs to last for a historical time scale of only 1000 years. For the transmutation of high-level radioactive waste a lot of research and development is still required. One aspect is the precise knowledge of nuclear data for reactions with fast neutrons. Nuclear reactions relevant for transmutation are being investigated in the framework of the european project ERINDA. First results from the new neutron time-of-flight facility nELBE at Helmholtz-Zentrum Dresden-Rossendorf will be presented.

  15. Preparative electrophoresis of industrial fission product solutions

    International Nuclear Information System (INIS)

    Tret, Joel

    1971-07-01

    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): 137 Cs, 90 Sr, 141+144 Ce, 91 Y, 95 Nb, 95 Zr, 103+106 Ru. (author) [fr

  16. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

    1994-12-31

    Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

  17. Nuclear fission and fission-product spectroscopy: 3. International workshop on nuclear fission and fission-product spectroscopy

    International Nuclear Information System (INIS)

    Goutte, Heloise; Fioni, Gabriele; Faust, Herbert; Goutte, Dominique

    2005-01-01

    The present book contains the proceedings of the third workshop in a series of workshops previously held in Seyssins in 1994 and 1998. The meeting was jointly organized by different divisions of CEA and two major international laboratories. In the opening address, Prof. B. Bigot, the French High Commissioner for Atomic Energy, outlined France's energy policy for the next few decades. He emphasized the continuing progress of nuclear fission in both technical and economic terms, allowing it to contribute to the energy needs of the planet even more in the future than it does today. Such progress implies a very strong link between fundamental and applied research based on experimental and theoretical approaches. The workshop gathered the different nuclear communities studying the fission process, including topics as the following: - nuclear fission experiments, - spectroscopy of neutron rich nuclei, - fission data evaluation, - theoretical aspects of nuclear fission, - and innovative nuclear systems and new facilities. The scientific program was suggested by an International Advisory Committee. About 100 scientists from 13 different countries attended the conference in the friendly working atmosphere of the Castle of Cadarache in the heart of the Provence. The proceedings of the workshop were divided into 11 sections addressing the following subject matters: 1. Cross sections and resonances (5 papers); 2. Fission at higher energies - I (5 papers); 3. Fission: mass and charge yields (4 papers); 4. Light particles and cluster emission (4 papers); 5. Spectroscopy of neutron rich nuclei (5 papers); 6. Resonances, barriers, and fission times (5 papers); 7. Fragment excitation and neutron emission (4 papers); 8. Mass and energy distributions (4 papers); 9. Needs for nuclear data and new facilities - I (4 papers); 10. Angular momenta and fission at higher Energies - II (3 papers); 11. New facilities - II (2 papers). A poster session of 8 presentations completed the workshop

  18. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

  19. Release of fission products in transients

    International Nuclear Information System (INIS)

    Christensen, H.; Lundqwist, R.

    1979-07-01

    A station for automatic sampling of coolant has been put in operation at the Oskarshamn-1 reactor. The release of 131 J and other fission products in spikes in connection with reactor trips and scheduled shutdowns has been measured. A model developed at General Electric has been used to predict the spike release in Oskarshamn-1 and the predicted values have been compared with experimental values. Literature data of iodine spikes in BWR and PWR have been reviewed. (author)

  20. Chemical immobilization of fission products reactive with nuclear reactor components

    International Nuclear Information System (INIS)

    Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

    1975-01-01

    This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

  1. Investigation of short-living fission products from the spontaneous fission of Cf-252

    International Nuclear Information System (INIS)

    Klonk, H.

    1976-01-01

    In this paper, a method of separating and measuring fission products of Cf-252 is presented. The measurement was achieved by means of γ-spectrometry and thus provides a quantitative analysis with a good separation of the fission products with respect to both atomic number Z and mass number A. The separation of the fission products from the fission source was achieved by means of solid traps. An automatic changing apparatus made it possible to keep irradiation and measuring times short, so even very short-lived fission products could be registered. The quantitative evaluation of primary fission products was made possible by correction according to Bateman equations. With that, the yields of single nuclides and the dispersion of charge can be determined. (orig./WL) [de

  2. Gas-phase transport of fission products

    International Nuclear Information System (INIS)

    Tang, I.N.; Munkelwitz, H.R.

    1982-01-01

    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

  3. Development of fission Mo production technology

    International Nuclear Information System (INIS)

    Kim, B. K.; Park, K. B.; Jun, B. J.; Park, J. H.; Choung, W. M.; Lee, K. I.; Woo, M. S.; Whang, D. S.; Kim, Y. K.; Yoo, J. H.; Sohn, D. S.; Lee, Y. W.; Na, S. H.; Koo, Y. H.; Hwang, D. H.; Joo, P. K.

    1997-08-01

    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

  4. Proposed partitioning and transmutation of long-lived nuclear wastes

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Rawlins, J.A.

    1991-01-01

    A means of transmuting key long-lived nuclear wastes, primarily the minor actinides (Np, Am, Cm) and iodine, using a hybrid proton accelerator and sub-critical lattice, is proposed. By partitioning light water reactor (LWR) spent fuel and by transmuting key elements, such as the plutonium, the minor actinides, and a few of the long-lived fission products, some of the most significant challenges in building a waste repository can be substantially reduced. The proposed machine would transmute the minor actinides and the iodine produced by 75 LWRs, and would generate usable electricity (beyond that required to run the large accelerator) of 850 MW e . 14 refs., 10 figs

  5. Tokamak transmutation of (nuclear) waste (TTW): Parametric studies

    International Nuclear Information System (INIS)

    Cheng, E.T.; Krakowski, R.A.; Peng, Y.K.M.

    1994-01-01

    Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low-aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses

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

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James

    2012-12-19

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

  7. Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium-Uranium-Plutonium Fuel Cycle

    Science.gov (United States)

    Marshalkin, V. Ye.; Povyshev, V. M.

    2017-12-01

    It is shown for a closed thorium-uranium-plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain 54 kg of fission products, 0.8 kg of thorium, 0.10 kg of uranium isotopes, 0.005 kg of plutonium isotopes, 0.002 kg of neptunium, and "trace" amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

  8. Fission product induced swelling of U–Mo alloy fuel

    International Nuclear Information System (INIS)

    Kim, Yeon Soo; Hofman, G.L.

    2011-01-01

    Highlights: ► We measured fuel swelling of U–Mo alloy by fission products at temperatures below 250 °C. ► We quantified the swelling portion of U–Mo by fission gas bubbles. ► We developed an empirical model as a function of fission density. - Abstract: Fuel swelling of U–Mo alloy was modeled using the measured data from samples irradiated up to a fission density of ∼7 × 10 27 fissions/m 3 at temperatures below ∼250 °C. The overall fuel swelling was measured from U–Mo foils with as-fabricated thickness of 250 μm. Volume fractions occupied by fission gas bubbles were measured and fuel swelling caused by the fission gas bubbles was quantified. The portion of fuel swelling by solid fission products including solid and liquid fission products as well as fission gas atoms not enclosed in the fission gas bubbles is estimated by subtracting the portion of fuel swelling by gas bubbles from the overall fuel swelling. Empirical correlations for overall fuel swelling, swelling by gas bubbles, and swelling by solid fission products were obtained in terms of fission density.

  9. Fuels and targets for the transmutation of high activity long lived radioactive wastes

    International Nuclear Information System (INIS)

    Pillon, S.; Warin, D.

    2010-01-01

    The authors present and comment the different strategies which can be adopted to transmute minor actinides (concerned reactors, in fast breeder reactors, in accelerator driven systems or ADS), and the chemical composition of transmutation fuels (actinide compounds, inert matrices, fuels and targets). They describe the behaviour of refractory ceramic fuels during their service life under irradiation with their different damage origins (neutrons, fission by-products, alpha particles), the fabrication of transmutation fuels and targets through different processes (metallurgical, co-precipitate, sol-gel, wax, infiltration of radioactive materials, VIPAC/SPHEREPAC) and the reprocessing or recycling of these transmutation fuels and targets

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

    International Nuclear Information System (INIS)

    Lammer, M.

    1990-11-01

    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

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

    International Nuclear Information System (INIS)

    Lammer, M.

    1994-06-01

    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

  12. Partitioning and transmutation (P and T) 1997. Status report

    International Nuclear Information System (INIS)

    Enarsson, Aasa; Landgren, A.; Liljenzin, J.O.; Skaalberg, M.; Spjuth, L.; Gudowski, W.; Wallenius, J.

    1998-05-01

    Research on and the evaluation of partitioning and transmutation are currently in progress in many industrial countries due to its potential as a long-term, sustainable energy source with low environmental impact and due to its ability to destroy many long-lived nuclides. The cost of the research and development work on partitioning and transmutation is considered to be so great that international co-operation is required. With respect to Sweden, we recommend a balanced research work on both partitioning and transmutation technology. Within the area of partitioning, it is above all a question of locating new reagents which can be used to simplify the necessary partitioning processes and minimize the losses. The requirements with respect to high selectivity and minor losses will be significantly higher in a recirculating system based on transmutation than in the reprocessing facilities of today where only uranium and plutonium are recovered. If the utilized reagents can be easily destroyed, by dry or wet incineration and conversion into non-complex gaseous chemical compounds, this will open up good opportunities for the recovery of the radionuclides. From a purely technical standpoint, it would seem that a combination of different types of reactor systems would give the best possible transmutation efficiency. While existing light water reactors can be utilized for increased plutonium incineration, there is currently consensus about the view that reactors with high-energy neutrons are necessary to achieve a sufficiently high transmutation efficiency for neptunium, americium, curium and certain fission products. By allowing an accelerator-based neutron source to drive a subcritical heavy metal-cooled reactor, the potential for transmutation of fission products is increased, at the same time that satisfactory safety margins are achieved for certain fuel types with a low share of delayed neutrons and a high heat conductivity. Regardless of what types of systems are

  13. Partitioning and transmutation (P and T) 1997. Status report

    Energy Technology Data Exchange (ETDEWEB)

    Enarsson, Aasa; Landgren, A.; Liljenzin, J.O.; Skaalberg, M.; Spjuth, L. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Chemistry; Gudowski, W.; Wallenius, J. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    1998-05-01

    Research on and the evaluation of partitioning and transmutation are currently in progress in many industrial countries due to its potential as a long-term, sustainable energy source with low environmental impact and due to its ability to destroy many long-lived nuclides. The cost of the research and development work on partitioning and transmutation is considered to be so great that international co-operation is required. With respect to Sweden, we recommend a balanced research work on both partitioning and transmutation technology. Within the area of partitioning, it is above all a question of locating new reagents which can be used to simplify the necessary partitioning processes and minimize the losses. The requirements with respect to high selectivity and minor losses will be significantly higher in a recirculating system based on transmutation than in the reprocessing facilities of today where only uranium and plutonium are recovered. If the utilized reagents can be easily destroyed, by dry or wet incineration and conversion into non-complex gaseous chemical compounds, this will open up good opportunities for the recovery of the radionuclides. From a purely technical standpoint, it would seem that a combination of different types of reactor systems would give the best possible transmutation efficiency. While existing light water reactors can be utilized for increased plutonium incineration, there is currently consensus about the view that reactors with high-energy neutrons are necessary to achieve a sufficiently high transmutation efficiency for neptunium, americium, curium and certain fission products. By allowing an accelerator-based neutron source to drive a subcritical heavy metal-cooled reactor, the potential for transmutation of fission products is increased, at the same time that satisfactory safety margins are achieved for certain fuel types with a low share of delayed neutrons and a high heat conductivity. Regardless of what types of systems are

  14. Recovery of noble metals from fission products

    International Nuclear Information System (INIS)

    Jenson, G.A.; Platt, A.M.; Mellinger, G.B.; Bjorklund, W.J.

    1982-11-01

    Scoping studies were started in 1979 to develop a cost-effective, waste-management-compatible process to extract noble metals from fission products. The process, involving the reaction with glassmelting chemicals, a metal oxide (PbO), and a reducing agent (charcoal), was demonstrated for recovering noble metals from simulated high-level waste oxides. The process has now been demonstrated on a laboratory scale (100 g) using irradiated fuels. Recoveries in the recovered lead averaged 80% for Pd, 60% for Rh, and 14% Ru. The resulting glass product was homogeneous in appearance, and the chemical durability was comparable to other waste oxides

  15. Detector for gaseous nuclear fission products

    International Nuclear Information System (INIS)

    Kobayashi, Yoshihiro; Kubo, Katsumi.

    1979-01-01

    Purpose: To facilitate the fabrication of a precipitator type detector, as well as improve the reliability. Constitution: Gas to be measured flown in an anode is stored in a gas processing system. By applying a voltage between the anode and the cathode, if positively charged Rb or Cs which is the daughter products of gaseous fission products are present in the gas to be measured, the daughter products are successively deposited electrostatically to the cathode. The daughter products issue beta-rays and gamma-rays to ionize the argon gas at the anode, whereby ionizing current flows between both of the electrodes. Pulses are generated from the ionizing current, and presence or absence, as well as the amount of the gaseous fission products are determined by the value recorded for the number of the pulses to thereby detect failures in the nuclear fuel elements. After the completion of the detection, the inside of the anode is evacuated and the cathode is heated to evaporate and discharge the daughter products externally. This eliminates the effects of the former detection to the succeeding detection. (Moriyama, K.)

  16. Burn-up physics in a coupled Hammer-Technion/Cinder-2 system and ENDF/B-V aggregate fission product thermal cross section validation

    International Nuclear Information System (INIS)

    Santos, A. dos.

    1990-01-01

    The new methodology developed in this work has the following purposes: a) to implement a burnup capability into the HAMMER-TECHNION/9/computer code by using the CINDER-2/10/computer code to perform the transmutation analysis for the actinides and fission products; b) to implement a reduced version of the CINDER-2 fission product chain structure to treat explicity nearly 99% of all original CINDER-2 fission product absorption in a typical PWR unit cell; c) to treat the effect of the fission product neutron absorption in an unit cell in a multigroup basis; d) to develop a tentative validation procedure for the ENOF/C-V stable and long-lived fission product nuclear data based on the available experimental data/11-14/. The analysis will be performed by using the reduce chain in the coupled system CINDER-2 to generate the time dependent effective four group cross sections for actinides and fission products and CINDER-2 to perform the complete transmutation analysis with its built-in chain structure. (author)

  17. Separation of fission products using inorganic exchangers

    International Nuclear Information System (INIS)

    Murthy, T.S.; Balasubramanian, K.R.; Rao, K.L.N.; Venkatachalam, R.; Varma, R.N.

    1981-01-01

    This paper describes the separation of long lived fission products like caesium-137, strontium-90 using inorganic exchangers ammonium phosphomolybdate and zirconium antimonate. A revised flow sheet is proposed for the sequential separation of these isotopes using the above two compounds. This is a modification of the earlier scheme developed which involved the use of four inorganic exchangers namely ammonium phosphomolybdate, manganese dioxide, zirconium antimonate and polyantimonic acid. The elution of the adsorbed elements like cerium, strontium, and sodium has been studied and it has been possible to elute these using different eluting agents. (author)

  18. Towards cleaner methods for the production of Mo-99 using refractory ceramics and its relevance to actinide partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Luca, V.; Dos Santos, L.; Vaccaro, J. [Comision Nacional de Energia Atomica, Centro Atomico Constituyentes, Av. General Paz 1499, 1650 San Martin, Buenos Aires (Argentina)

    2016-07-01

    Mo-99 is the most utilized isotope in nuclear medicine accounting for over 30 million medical diagnostic procedures annually worldwide. The process for the production of Mo-99 through fission of U-235 normally involves the irradiation of UAl{sub x} dispersion plate fuel in a research reactor, the subsequent dissolution of the fuel plate, the selective separation of the Mo-99 from all of the other fission products and possibly also the recovery of U-235 for future reuse. Compared to the amount of product recovered, copious radioactive waste is generated during the Mo-99 production process. Gaseous wastes are produced at the head-end during the plate dissolution and several liquid wastes are produced during the recovering of Mo-99 using solid extractants, typically polymeric ion exchange resins, which themselves constitute an additional waste stream. It would be extremely advantageous to devise a new process that generates little or no waste. We have been working on a new strategy for the production of fission Mo-99 that involves replacing the dispersion plate targets that are used in the traditional process with inert or active matrix fuel particles that do not need to be dissolved. In one embodiment of the strategy the preparation of new highly porous ZrC{sub x} and graphite-ZrC{sub x} composite target kernels are used that are prepared through polymer templating. The surface properties of these porous materials have been studied and are such that they can be easily loaded with uranium, or for that matter, with any other actinide. In our work we are exploring the possibility of selectively extracting the Mo-99 from the irradiated target kernels by either solution or gas-phase methods and then easily recover the uranium. The fission product-containing kernels can be oxidized in air to generate ZrO{sub 2} that can act as a stable host material either alone or as part of a multiphase ceramic matrix or possibly even as an actinide transmutation host. At the conceptual

  19. Evaluation on transmutation performance of minor actinides with high-flux BWR

    International Nuclear Information System (INIS)

    Setiawan, M.B.; Kitamoto, A.; Taniguchi, A.

    2001-01-01

    The performance of high-flux BWR (HFBWR) for burning and/or transmutation (B/T) treatment of minor actinides (MA) and long-lived fission products (LLFP) was discussed herein for estimating an advanced waste disposal with partitioning and transmutation (P and T). The concept of high-flux B/T reactor was based on a current 33 GWt-BWR, to transmute the mass of long-lived transuranium (TRU) to short-lived fission products (SLFP). The nuclide selected for B/T treatment was MA (Np-237, Am-241, and Am-243) included in the discharged fuel of LWR. The performance of B/T treatment of MA was evaluated by a new function, i.e. [F/T ratio], defined by the ratio of the fission rate to the transmutation rate in the core, at an arbitrary burn-up, due to all MA nuclides. According to the results, HFBWR could burn and/or transmute MA nuclides with higher fission rate than BWR, but the fission rate did not increase proportionally to the flux increment, due to the higher rate of neutron adsorption. The higher B/T fraction of MA would result in the higher B/T capacity, and will reduce the units of HFBWR needed for the treatment of a constant mass of MA. In addition, HFBWR had a merit of higher mass transmutation compared to the reference BWR, under the same mass loading of MA

  20. Energy production using fission fragment rockets

    International Nuclear Information System (INIS)

    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 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

  1. The role of fission products in whole core accidents

    Energy Technology Data Exchange (ETDEWEB)

    Baker, A R [FRSD, UKAEA, RNPDE, Risley, Warrington (United Kingdom); Teague, H J [SRD, UKAEA, Culcheth, Warrington (United Kingdom)

    1977-07-01

    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)

  2. RSAC, Gamma Doses, Inhalation and Ingestion Doses, Fission Products Inventory after Fission Products Release

    International Nuclear Information System (INIS)

    Richardson, L.C.

    1967-01-01

    1 - Description of problem or function: RSAC generates a fission product inventory from a given set of reactor operating conditions and then computes the external gamma dose, the deposition gamma dose, and the inhalation-ingestion dose to critical body organs as a result of exposure to these fission products. Program output includes reactor operating history, fission product inventory, dosages, and ingestion parameters. 2 - Method of solution: The fission product inventory generated by the reactor operating conditions and the inventory remaining at various times after release are computed using the equations of W. Rubinson in Journal of Chemical Physics, Vol. 17, pages 542-547, June 1949. The external gamma dose and the deposition gamma dose are calculated by determining disintegration rates as a function of space and time, then integrating using Hermite's numerical techniques for the spatial dependence. The inhalation-ingestion dose is determined by the type and quantity of activity inhaled and the biological rate of decay following inhalation. These quantities are integrated with respect to time to obtain the dosage. The ingestion dose is related to the inhalation dose by an input constant

  3. Migration of fission products in UO2. Final report

    International Nuclear Information System (INIS)

    Prussin, S.G.; Olander, D.R.

    1995-01-01

    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

  4. Impact of fuel chemistry on fission product behaviour

    International Nuclear Information System (INIS)

    Poortmans, C.; Van Uffelen, P.; Van den Berghe, S.

    1999-01-01

    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

  5. Partitioning and transmutation (P and D) 1995. A review of the current state of the art

    International Nuclear Information System (INIS)

    Skaalberg, M.; Landgren, A.; Spjuth, L.; Liljenzin, J.O.; Gudowski, W.

    1995-12-01

    The recent development in the field of partitioning and transmutation (P/T) is reviewed and evaluated. Current national and international R and D efforts are summarized. Nuclear transmutation with energy production is feasible in nuclear reactors where fast and thermal breeders are the most efficient for transmutation purposes. The operation of subcritical nuclear reactors by high current proton accelerators that generate neutrons in a spallation target is also an interesting option for transmutation and energy production, that has to be more carefully evaluated. These accelerator-driven systems are probably the only solution for the transmutation of long-lived fission products with small neutron capture cross sections and actinide isotopes with small fission cross sections. The requirements on the separation chemistry in the partitioning process depends on the transmutation strategy chosen. Recent developments in aqueous based separation chemistry opens some interesting possibilities to meet some of the requirements, such as separation of different actinides and some fission products and reduction of secondary waste streams. In the advanced accelerator-driven transmutation systems proposed, liquid fuels such as molten salts are considered. The partitioning processes that can be used for these types of fuel will, however, require a long term research program. The possibility to use centrifuge separation is an interesting partitioning option that recently has been proposed. 51 refs, 7 figs, 3 tabs

  6. 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

  7. Polyphase diffusion of fission products in graphite

    International Nuclear Information System (INIS)

    Dannert, V.

    1989-05-01

    The report attempts to give an introduction into the subject of fission product transport in nuclear graphite and results in an extended proposal of a transport-model. Beginning with a rough description of the graphite in question, an idea about the physical transport-phenomena in graphite is developed. Some of the basic experimental methods, especially techniques of porosimetry, determination of sorption-isotherms and of course several transport-experiments, are briefly described and their results are discussed. Some of the most frequent transport models are introduced and assessed with the criteria emphasized in this report. An extended model is proposed including the following main ideas: The transport of the fission-products is regarded as a two-phase-diffusion process through the open pores of the graphite. The two phases are: surface-diffusion and gas-diffusion. A time-dependent coupling of the two diffusion-phases by sorption-isotherms and a concentration-dependence of the surface diffusion coefficient, also related to the physical behaviour of the sorption-isotherms, are the basic properties of the proposed model. (orig./HP) [de

  8. Fuel and fission product release from sodium

    International Nuclear Information System (INIS)

    Sauter, H.

    1992-01-01

    The NALA program at Kernforschungszentrum Karlsruhe is concerned with the release of fuel and fission products from hot or boiling sodium pools (radiological secondary source term) in a liquid-metal fast breeder reactor accident scenario with tank failure. The main concern is to determine retention factors (RF), to uncover the most essential parameters that influence the RF values, and to describe the way they do it. In the framework of the last NALA series, NALA IIIc, the influence of sodium-concrete interaction was investigated, partly with subsequent sodium burning. In our experiments, ∼3 kg of sodium and added pieces of concrete reaching from 4 to 40 g was used. The composition of the concrete was suitable for shielding and construction as used in the SNR-300 reactor. Fuel was simulated by 20-μm particles of depleted UO 2 , and CeO 2 , NaI, and TeO 2 were used as fission products. Most experiments were performed in an inert argon gas atmosphere with monitored hydrogen development. In some cases, the preheated pool was allowed to come into contact with ambient air, which caused an ordinary sodium fire. For the latter case, we used the 220-m 3 FAUNA vessel as an outer containment and collected the fire aerosols by a trap and subsequent filters for analysis

  9. Resuspension of fission products from sump water

    International Nuclear Information System (INIS)

    Bunz, H.; Koyro, M.; Propheter, B.; Schoeck, W.; Wagner-Ambs, M.

    1992-11-01

    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

  10. Irradiation effects and behaviour of fission products in zirconia and spinel

    International Nuclear Information System (INIS)

    Gentils, A.

    2003-10-01

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

  11. Calculated apparent yields of rare gas fission products

    International Nuclear Information System (INIS)

    Delucchi, A.A.

    1975-01-01

    The apparent fission yield of the rare gas fission products from four mass chains is calculated as a function of separation time for six different fissioning systems. A plot of the calculated fission yield along with a one standard deviation error band is given for each rare gas fission product and for each fissioning system. Those parameters in the calculation that were major contributors to the calculated standard deviation at each separation time were identified and the results presented on a separate plot. To extend the usefulness of these calculations as new and better values for the input parameters become available, a third plot was generated for each system which shows how sensitive the derived fission yield is to a change in any given parameter used in the calculation. (U.S.)

  12. ELSA: A simplified code for fission product release calculations

    International Nuclear Information System (INIS)

    Manenc, H.; Notley, M.J.

    1996-01-01

    During a light water reactor severe accident, fission products are released from the overheated core as it progressively degrades. A new computer module named ELSA is being developed to calculate fission product release. The authors approach is to model the key phenomena, as opposed to more complete mechanistic approaches. Here they present the main features of the module. Different release mechanisms have been identified and are modeled in ELSA, depending on fission product volatility: diffusion seems to govern the release of the highly volatile species if fuel oxidation is properly accounted for, whereas mass transport governs that of lower volatility fission products and fuel volatilization that of the practically involatile species

  13. Simulation of Fission Product Liftoff Behavior During Depressurization Transients

    International Nuclear Information System (INIS)

    Tak, Nam-il; Yoon, Churl; Lee, Sung Nam

    2016-01-01

    As one of crucial technologies for the NHDD project, the development of the GAMMA-FP code is on-going. The GAMMA-FP code is targeted for fission product transport analysis under accident conditions. A well-known experiment named COMEDIE considered two important phenomena, i.e., fission product plateout and liftoff, for fission product transport within the primary circuit of a prismatic high temperature gas cooled reactor. The accumulated fission products on the structural material via the plateout can be liftoff during a blowdown phase after a pipe break accident. Since the fission product liftoff can increase a radioactivity risk, it is important to predict the amount of fission product liftoff during depressurization accidents. In this work, a model for fission product liftoff is implemented into the GAMMA-FP code and the GAMMA-FP code with the implemented model is validated using the COMEDIE blowdown test data. The results of GAMMA-FP show that the GAMMA-FP code can reliably simulate a pressure transient during blowdown phase after a pipe break accident. In addition, a reasonable amount of fission product liftoff was predicted by the GAMMA-FP code. The maximum difference between the measured and predicted liftoff fraction was less than a factor of 10. More in-depth study is required to increase the accuracy of prediction for a fission product liftoff

  14. Fission-product yields for thermal-neutron fission of curium-243

    International Nuclear Information System (INIS)

    Breederland, D.G.

    1982-01-01

    Cumulative fission yields for 25 gamma rays emitted during the decay of 23 fission products produced by thermal-neutron fission of 243 Cm have been determined. Using Ge(Li) spectroscopy, 33 successive pulse-height spectra of gamma rays emitted from a 77-ng sample of 243 Cm over a period of approximately two and one-half months were analyzed. Reduction of these spectra resulted in the identification and matching of gamma-ray energies and half-lives to specific radionuclides. Using these results, 23 cumulative fission-product yields were calculated. Only those radionuclides having half-lives between 6 hours and 65 days were observed. Prior to this experiment, no fission-product yields had been recorded for 243 Cm

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

    Science.gov (United States)

    Bhatia, C.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; 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.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

    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.

  16. Spallation reaction study for fission products in nuclear waste: Cross section measurements for {sup 137}Cs and {sup 90}Sr on proton and deuteron

    Energy Technology Data Exchange (ETDEWEB)

    Wang, H., E-mail: wanghe@ribf.riken.jp [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Otsu, H.; Sakurai, H.; Ahn, D.S. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Aikawa, M. [Faculty of Science, Hokkaido University, Sapporo 060-0810 (Japan); Doornenbal, P.; Fukuda, N.; Isobe, T. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Kawakami, S. [Department of Applied Physics, University of Miyazaki, Miyazaki 889-2192 (Japan); Koyama, S. [Department of Physics, University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-0033 (Japan); Kubo, T.; Kubono, S.; Lorusso, G. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Maeda, Y. [Department of Applied Physics, University of Miyazaki, Miyazaki 889-2192 (Japan); Makinaga, A. [Graduate School of Medicine, Hokkaido University, North-14, West-5, Kita-ku, Sapporo 060-8648 (Japan); Momiyama, S. [Department of Physics, University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-0033 (Japan); Nakano, K. [Department of Advanced Energy Engineering Science, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Niikura, M. [Department of Physics, University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-0033 (Japan); Shiga, Y. [Department of Physics, Rikkyo University, 3-34-1 Nishi-Ikebukuro, Toshima, Tokyo 171-8501 (Japan); RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Söderström, P.-A. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); and others

    2016-03-10

    We have studied spallation reactions for the fission products {sup 137}Cs and {sup 90}Sr for the purpose of nuclear waste transmutation. The spallation cross sections on the proton and deuteron were obtained in inverse kinematics for the first time using secondary beams of {sup 137}Cs and {sup 90}Sr at 185 MeV/nucleon at the RIKEN Radioactive Isotope Beam Factory. The target dependence has been investigated systematically, and the cross-section differences between the proton and deuteron are found to be larger for lighter spallation products. The experimental data are compared with the PHITS calculation, which includes cascade and evaporation processes. Our results suggest that both proton- and deuteron-induced spallation reactions are promising mechanisms for the transmutation of radioactive fission products.

  17. ENDF/B fission product decay data

    International Nuclear Information System (INIS)

    Rose, P.F.; Burrows, T.W.

    1976-08-01

    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 (BR) 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

  18. Transmutation of radioactive nuclear waste – present status and ...

    Indian Academy of Sciences (India)

    Transmutation of long-lived actinides and fission products becomes an important issue of the overall nuclear fuel cycle assessment, both for existing and future reactor systems. Reliable nuclear data are required for analysis of associated neutronics. The present paper gives a review of the status of nuclear data analysis ...

  19. ENDF/B-5. Fission Product Yields File

    International Nuclear Information System (INIS)

    Schwerer, O.

    1985-10-01

    The ENDF/B-5 Fission Product Yields File contains a complete set of independent and cumulative fission product yields, representing the final data from ENDF/B-5 as received at the IAEA Nuclear Data Section in June 1985. Yields for 11 fissioning nuclides at one or more neutron incident energies are included. The data are available costfree on magnetic tape from the IAEA Nuclear Data Section. (author). 4 refs

  20. Estimation of penetration depth of fission products in cladding Hull

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Jung, Yang Hong; Yoo, Byong Ok; Choo, Yong Sun; Hong, Kwon Pyo

    2005-01-01

    A disposal and a reprocessing for spent fuel rod with high burnup need de-cladding procedure. Pellet in this rod has been separated from a cladding hull to reduce a radioactivity of hull by chemical and mechanical methods. But fission products and actinides(U,Pu) still remain inside of cladding hull by chemical bonding and fission spike, which is called as 'contamination'. More specific removal of this contamination would have been considered. In this study, the sorts of fission products and penetration depth in hull were observed by EPMA test. To analyze this behavior, SRIM 2000 code was also used as energies of fission products and an oxide thickness of hull

  1. Determination of the fission products yields, lanthanide and yttrium, in the fission of 238U with neutrons of fission spectra

    International Nuclear Information System (INIS)

    Nicoli, I.G.

    1981-06-01

    A radiochemical investigation is performed to measure the cumulative fission product yields of several lantanides and yttrium nuclides in the 238 U by fission neutron spectra. Natural and depleted uranium are irradiated under the same experimental conditions in order to find a way to subtract the contribution of the 235 U fission. 235 U percentage in the natural uranium was 3.5 times higher than in the depleted uranium. Uranium oxides samples are irradiated inside the core of the Argonaut Reactor, at the Instituto de Engenharia Nuclear, and the lantanides and yttrium are chemically separated. The fission products gamma activities were detected, counted and analysed in a system constituted by a high resolution Ge(Li) detector, 4096 multichannel analyser and a PDP-11 computer. Cumulative yields for fission products with half-lives between 1 to 33 hours are measured: 93 Y, 141 La, 142 La, 143 Ce and 149 Nd. The chain total yields are calculated. The cumulative fission yields measured for 93 Y, 141 La, 142 La, 143 Ce and 149 Nd are 4,49%, 4,54%, 4,95%, 4,16% and 1,37% respectively and they are in good agreement with the values found in the literature. (Author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

    Gentils, A

    2003-10-01

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

  3. Fission products stability in uranium dioxide

    International Nuclear Information System (INIS)

    Brillant, G.; Gupta, F.; Pasturel, A.

    2011-01-01

    Fission product stability in nuclear fuels is investigated using density functional theory (DFT). In particular, incorporation and solution energies of He, Kr, Xe, I, Te, Ru, Sr and Ce in pre-existing trap sites of UO 2 (vacancies, interstitials, U-O divacancy, and Schottky trio defects) are calculated using the projector-augmented-wave method as implemented in the Vienna ab initio simulation package. Correlation effects are taken into account within the DFT+U approach. The stability of many binary and ternary compounds in comparison to soluted atoms is also explored. Finally the involvement of FP in the formation of metallic and oxide precipitates in oxide fuels is discussed in the light of experimental results.

  4. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

    Finck, P.J.; Stenberg, C.G.

    1998-01-01

    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

  5. Core degradation and fission product release

    International Nuclear Information System (INIS)

    Wright, R.W.; Hagen, S.J.L.

    1992-01-01

    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

  6. Fission product transport at Three Mile Island

    International Nuclear Information System (INIS)

    Owen, D.E.; Cox, T.E.; Broughton, J.M.

    1985-01-01

    Three Mile Island Unit 2 radionuclide analyses are reviewed and summarized in order to determine how fission products moved to various parts of the reactor system at the time of the accident. Despite high fuel temperatures and major core damage, the core retained a very large fraction of most radionuclides. Reactor coolant, either remaining in the primary system or released to various sumps and tanks, retained significant quantities of cesium and iodine. Noble gases were effectively retained within the containment building with the exception of small releases to the environment. Long-term deposition and retention on vessel, piping, and bulding surfaces were insignificant for all isotopes examined. The measured partitioning of radionuclides within these systems is tabulated and recommendations for additional analyses are presented

  7. Retention of fission products in air filters

    International Nuclear Information System (INIS)

    Sobnack, R.

    1986-01-01

    The plume from the Chernobyl nuclear reactor reached London in the morning of 1st May. Less than two weeks later, the Physics Department, University of Surrey, reported a measurable level of radioactivity in air filters. On 15th May air filters from within the air conditioning plant of the Radioisotope Department at the London Hospital were removed for radiation checks. Crude tests with a geiger counter gave readings of 5-10 times higher than background levels. Gamma-ray spectroscopy of the departmental air filters (AF1) using a 127 mm NaI detector revealed a pattern characteristic of emissions of fission products from a nuclear reactor. Another air filter (AF2), from the home of a member of staff, was much less active. Because of the complexity of the gamma-ray spectrum and the relatively high level of emission from the departmental air filter, a thorough investigation was carried out using a high purity germanium detector. (author)

  8. (Fuel, fission product, and graphite technology)

    Energy Technology Data Exchange (ETDEWEB)

    Stansfield, O.M.

    1990-07-25

    Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

  9. Actual point about fission products vitrification

    International Nuclear Information System (INIS)

    Bonniaud, R.

    1982-05-01

    The main characteristics concerning the continuous vitrification process for the confinement of fission product solutions operated at AVM are summarized. The general principle of a vitrification plant is described. The AVM plant efficiency as also its conception of consumable parts interchangeability are satisfying. The evolution of the process and its application developped in two ways: a more spaced installation conception and the improvement of the weak points remarked at AVM, as also the capacity of output. Two industrial units are designed at La Hague. The future evolution of the process aims at manufacturing glass at higher temperatures about 1400 degrees Celsius. Some problems remain to be resolved for the using of ceramic melters associated with a calcination unit. The studies provide for a satisfying behaviour for the material to long-term. The risks of damage by crystallisation, leaching and effects of alpha emission are analysed [fr

  10. Library of data for fission products

    International Nuclear Information System (INIS)

    Blachot, Jean; Devillers, Christian; Tourreil, Roland de; Nimal, Bernadette; Fiche, Charles; Noel, J.-P.

    1975-10-01

    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 [fr

  11. A study on transmutation of LLFPs using various types of HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Kora, Kazuki, E-mail: kora_k@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka-shi, Fukuoka-ken (Japan); Nakaya, Hiroyuki; Matsuura, Hideaki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka-shi, Fukuoka-ken (Japan); Goto, Minoru; Nakagawa, Shigeaki; Shimakawa, Satoshi [Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki, Ibaraki-ken (Japan)

    2016-04-15

    Highlights: • We propose utilization of a variety of HTGRs for LLFP transmutation and storage. • The transmutation performance of four types of HTGRs was examined and compared. • Some types of HTGRs show preferable characteristics for LLFP transmutation. - Abstract: In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system “GTHTR300”, which is the subject of our previous research, a small modular HTGR plant “HTR50S” and two types of plutonium burner HTGRs “Clean Burn with MA” and “Clean Burn without MA” were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.

  12. Apparatus for measuring the release of fission gases and other fission products by degassing

    Energy Technology Data Exchange (ETDEWEB)

    Stradal, Karl Alfred

    1970-10-15

    In gas-cooled high-temperature reactors, the fuel is, in general, inserted in the fuel elements in the form of small particles, which are, for example, coated with pyrolytic carbon. The purpose of this coating is to keep the fission products separate from the coolant gas. The further development of these coated particles makes it necessary to check the retention capacity. One possible method of doing this is the degassing test after irradiation in the reactor. An apparatus is described below, which was developed and installed in order to measure to a higher degree of sensitivity and in serial measurements the release of fission gases and sparingly volatile fission products.

  13. Chemical factors affecting fission product transport in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

  14. Immobilization of fission products in phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.

    1996-01-01

    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. Fission product source term research at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1985-01-01

    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

  16. REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

    Science.gov (United States)

    Chiotti, P.

    1964-02-01

    A process of regenerating a magnesium-thorium alloy contaminated with fission products, protactinium, and uranium is presented. A molten mixture of KCl--LiCl-MgCl/sub 2/ is added to the molten alloy whereby the alkali, alkaline parth, and rare earth fission products (including yttrium) and some of the thorium and uranium are chlorinated and

  17. On accelerator neutron/γ-ray incineration of long-lived fission products

    International Nuclear Information System (INIS)

    Nakamura, H.

    1995-01-01

    A methodology for evaluation and control of the incineration of Long-Lived Fission Products (LLFPs) by using the accelerator neutrons/γ-rays is presented. An arbitrary number of the auxiliary transmutation chains, each of which consists of a LLFP and its reaction precursors up to 144, are used for calculating the time-dependent depletion-production of the LLFP. In the energy range below 20 MeV, about 20 types of neutron reaction are energetically possible. The semi-empirical formulas and its parameter systematics are used for all the energy dependent reaction cross sections. A computer code TRANS-N.G based on the foregoing prescription for nuclear reactions could be applied to the LLFP incineration strategies under a large variety of situations. (author)

  18. Status of fission product yield data

    International Nuclear Information System (INIS)

    Cuninghame, J.G.

    1978-01-01

    The topics covered in this paper are: (a) cumulative yields in thermal neutron fission and in fast fission up to 14 MeV incident neutron energy, (b) dependence of the yields on incident neutron energy and spectrum, (c) independent yields, (d) charge dispersion and distribution, and (e) yields of light particles from ternary fission. The paper reviews information on these subjects for fission of actinides from 232 Th upwards with special emphasis on data published since the 1973 Bologna FPND Panel, compares data sets, and discusses the gaps still to be found in them. (author)

  19. Volatilization and reaction of fission products in flowing steam

    International Nuclear Information System (INIS)

    Johnson, I.; Steidl, D.V.; Johnson, C.E.

    1985-01-01

    The principal risk to the public from nuclear power plants derives from the highly radioactive atoms (fission products) generated as energy is produced in the nuclear fuel. The revolatilization of fission products from reactor system surfaces due to self-heating by radioactive decay has become a complicating factor in the source-term redefinition effort. It has had a major impact on calculations of fission product distributions in accident safety analyses. The focus of this research effort was to investigate the volatilization and transport of fission products and control rod materials in a flowing gaseous steam-hydrogen mixture. Fission product and control rod materials in various combinations were studied including CsI, CsOH, TeO 2 , SrO, Ag, In, Cd and Mn. The vaporization behavior of the deposits were characterized with respect to vaporization rates, chemical species and downstream transport behavior

  20. Fission product and aerosol behaviour within the containment

    International Nuclear Information System (INIS)

    Beard, A.M.; Benson, C.G.; Bowsher, B.R.; Dickinson, S.; Nichols, A.L.

    1990-04-01

    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

  1. Characteristics of fission product release from a molten pool

    International Nuclear Information System (INIS)

    Yun, J.I.; Suh, K.Y.; Kang, C.S.

    2001-01-01

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

  2. Behavior of solid fission products in irradiated fuel

    International Nuclear Information System (INIS)

    Song, Ung Sup; Jung, Yang Hong; Kim, Hee Moon; Yoo, Byun Gok; Kim, Do Sik; Choo, Yong Sun; Hong, Kwon Pyo

    2004-01-01

    Many fission products are generated by fission events in UO 2 fuel under irradiation in nuclear reactor. Concentration of each fission product is changed by conditions of neutron energy spectrum, fissile material, critical thermal power, irradiation period and cooling time. Volatile materials such as Cs and I, the fission products, degrade nuclear fuel rod by the decrease of thermal conductivity in pellet and the stress corrosion cracking in cladding. Metal fission products (white inclusion) make pellet be swelled and decrease volume of pellet by densification. It seems that metal fission products are filled in the pore in pellet and placed between UO 2 lattices as interstitial. In addition, metal oxide state may change structural lattice volume. Considering behavior of fission products mentioned above, concentration of them is important. Fission products could be classified as bellows; solid solution in matrix : Sr, Zr, Nb, Y, La, Ce, Pr, Nd, Pm, Sm - metal precipitates : Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sb, Te - oxide precipitates : Ba, Zr, Nb, Mo, (Rb, Cs, Te) - volatile and gases : Kr, Xe, Br, I, (Rb, Cs, Te)

  3. Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity [O1] - Fundamentals of P{sup 3} Mechanism and Methodology Development for Plutonium Categorization

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Masaki [Research Laboratory for Nuclear Reactor, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro, Tokyo, 1528550 (Japan)

    2009-06-15

    'Protected Plutonium Production (P{sup 3})' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MAs). Doping the small amount of MAs such as {sup 237}Np or {sup 241}Am with large neutron capture cross-section into the uranium fuel to enhance the production of {sup 238}Pu or {sup 242}Pu, which have high spontaneous fission neutron source or also high decay heat to makes the process of the nuclear weapon manufacture and maintenance technologically difficult, can be effective for improving the isotopic barrier of proliferation resistance of the plutonium in thermal reactors. Super weapon grade plutonium could be produced in the blanket of a conventional FBR. However, by increasing the {sup 238}Pu or {sup 242}Pu ratio in the total plutonium by MAs doping into the fresh blanket, the protected plutonium with high proliferation-resistance can be bred. A new evaluation function, 'attractiveness', defined as a ratio of potential of fission yield to the technological difficulties of nuclear explosive device, has been proposed to evaluate the proliferation resistance of Pu based on the nuclear material property for Plutonium Categorization. In the conference, the fundamentals of P{sup 3} mechanism by transmutation of MA, and the comparison of the 'attractiveness' of the Pu produced in advanced reactors based on P{sup 3} mechanism and in the conventional reactors will be presented. Instead of the geological disposal or just their burning of MAs by the fission reaction, they should be treated as valuable fertile materials to enhance the proliferation resistance of plutonium produced in the thermal and fast breeder reactors for peace and sustainable prosperity in future. Acknowledgement: Some parts of this work have been supported by the Ministry of Education, Culture, Sports, Science and Technology in Japan. (authors)

  4. Transmutation of high-level radioactive waste by a charged particle accelerator

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi.

    1993-01-01

    Transmutation of minor actinides and fission products using proton accelerators has many advantages over a transmutor operated in a critical condition. The energy required for this transmutation can be reduced by multiplying the spallation neutrons in a subcritical assembly surrounding the spallation target. The authors have studied the relation between the energy requirements and the multiplication factor, k, of the subcritical assembly, while varying the range of several parameters in the spallation target. A slightly subcritical reactor is superior to a reactor with large subcriticality in the context of the energy requirement of a small proton accelerator, the extent of radiation damage, and other safety problems. To transmute the fission products, the transmutor reactor must have a good neutron economy, which can be provided by a transmutor operated by a proton accelerator. The paper discusses the use of minor actinides to improve neutronics characteristics, such as a long fuel burn-up rather than simply transmuting this valuable material

  5. Transmutation of nuclear waste in nuclear reactors

    International Nuclear Information System (INIS)

    Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

    1996-03-01

    The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

  6. First results and future trends for the transmutation of long-lived radioactive wastes

    International Nuclear Information System (INIS)

    Prunier, C.; Salvatores, M.; Guerin, Y.; Zaetta, A.

    1993-01-01

    In the frame of the CEA SPIN program, a project has been set-up at the Direction of Nuclear Reactors of CEA, to study the transmutation of long-lived radioactive products (both minor actinides and fission products) resulting from the operation of current nuclear power plants. The program is focused on: transmutation in minor actinides (Np, Am) in fission reactors of known technology (both of the PWR or the fast reactor type), using the so-called ''homogeneous'' (mixed with Uranium or Uranium-Plutonium), and ''heterogeneous'' (mixed with inert matrices) recycling modes for both type of reactors. Transmutation studies in dedicated devices (both fission reactors with actinide/plutonium fuel or with high thermal flux, and particle accelerator-based systems). Fuel studies related to both homogeneous and heterogeneous recycling modes in fission reactors. For the homogeneous recycling mode, some experimental irradiations results are available from past PHENIX programs. For the heterogeneous mode, very limited experimental results are available, and new theoretical and experimental work is underway on the use of appropriate inert matrices. Basic data studies to assess the quality of existing nuclear data for fission reactor transmutation studies, future data needs of relevance, and model/data developments needed for accelerator-based systems. Strategy studies, to evaluate the consequences of the different transmutation options on the fuel cycle, according to different scenarios of nuclear power development. 7 refs., 3 figs., 5 tabs

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

    International Nuclear Information System (INIS)

    De, A.K.; Luckscheiter, B.; Lutze, W.; Malow, G.; Schiewer, E.

    1976-01-01

    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

  8. BIG-10 fission product generation and reaction rates

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1976-01-01

    Fission product generation rates for high quality fission foils and reaction rates of nonfission foils have been measured by gamma ray activation analyses. These foils were irradiated in the BIG-10 facility and the activities were measured by NaI counting techniques

  9. Efficiency Of Transuranium Nuclides Transmutation

    International Nuclear Information System (INIS)

    Kazansky, Yu.A.; Klinov, D.A.; Semenov, E.V.

    2002-01-01

    One of the ways to create a wasteless nuclear power is based on transmutation of spent fuel nuclides. In particular, it is considered that the radioactivity of the nuclear power wastes should be the same (or smaller), than radioactivity of the uranium and the thorium extracted from entrails of the Earth. The problem of fission fragments transmutation efficiency was considered in article, where, in particular, the concepts of transmutation factor and the ''generalised'' index of biological hazard of the radioactive nuclides were entered. The transmutation efficiency has appeared to be a function of time and, naturally, dependent on nuclear power activity scenario, from neutron flux, absorption cross-sections of the nuclides under transmutation and on the rate of their formation in reactors. In the present paper the efficiency of the transmutation of transuranium nuclides is considered

  10. An analysis of the additional fission product release phenomena

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Nagai, Hitoshi

    1978-09-01

    The additional fission product release behavior through a defect hole on the cladding of fuel rods has been studied qualitatively with a computer program CODAC-ARFP. The additional fission product release phenomena are described as qualitative evaluation. The additional fission product release behavior in coolant temperature and pressure fluctuations and in reactor start-up and shut-down depends on coolant water flow behavior into and from the free space of fuel rods through a defect hole. Based on the results of evaluations, the experimental results with an inpile water loop OWL-1 are described in detail. The estimation methods of fission product quantity in the free space and fission product release ratio (quantity released into the coolant/quantity in the free space before beginning of release) are necessary for analysis of the fission product release behavior; the estimation method of water flow through a defect hole is also necessary. In development of the above estimation methods, outpile and capsule experiments supporting the additional fission product release experiments are required. (author)

  11. Electrochemical separation of actinides and fission products in molten salt electrolyte

    Energy Technology Data Exchange (ETDEWEB)

    Gay, R.L.; Grantham, L.F.; Fusselman, S.P. [Rockwell International/Rocketdyne Division, Canoga Park, CA (United States)] [and others

    1995-10-01

    Molten salt electrochemical separation may be applied to accelerator-based conversion (ABC) and transmutation systems by dissolving the fluoride transport salt in LiCl-KCl eutectic solvent. The resulting fluoride-chloride mixture will contain small concentrations of fission product rare earths (La, Nd, Gd, Pr, Ce, Eu, Sm, and Y) and actinides (U, Np, Pu, Am, and Cm). The Gibbs free energies of formation of the metal chlorides are grouped advantageously such that the actinides can be deposited on a solid cathode with the majority of the rare earths remaining in the electrolyte. Thus, the actinides are recycled for further transmutation. Rockwell and its partners have measured the thermodynamic properties of the metal chlorides of interest (rare earths and actinides) and demonstrated separation of actinides from rare earths in laboratory studies. A model is being developed to predict the performance of a commercial electrochemical cell for separations starting with PUREX compositions. This model predicts excellent separation of plutonium and other actinides from the rare earths in metal-salt systems.

  12. Fission-product SiC reaction in HTGR fuel

    International Nuclear Information System (INIS)

    Montgomery, F.

    1981-01-01

    The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

  13. Status of decay data of fission products

    International Nuclear Information System (INIS)

    Blachot, J.

    1978-01-01

    Fission products (F.P.) are neutron rich isotopes ranging from Zn to Tm. The status of decay data of F.P. was described at the Bologna Panel 1973 by Rudstam. Since then, FPND have improved in general, but still much is valid of what Rudstam said about the accuracies of FPND. The lack of decay data for the short lived F.P. has been considerably reduced, and some of the short lived F.P. have now well studied decay data. The present status of decay data is given in this review, which is composed of six sections. In the first one, the principal new facilities used in decay data measurements are reviewed. The second part is devoted to the total decay energy (Q). In the third Section, the half lives are treated. In the fourth and fifth Sections, beta and gamma energies and intensities, and also average values are discussed. Finally, the last Section considers the different files and compilations devoted to the decay of F.P

  14. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    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 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  15. Chemistry of fission product iodine under nuclear reactor accident conditions

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

  16. RSAC-6, Gamma doses, inhalation and ingestion doses, fission products inventory after fission products release

    International Nuclear Information System (INIS)

    Wenzel, Douglas R.; Schrader, Brad J.

    2007-01-01

    1 - Description of program or function: RSAC-6 is the latest version of the program RSAC (Radiological Safety Analysis Computer Program). It calculates the consequences of a release of radionuclides to the atmosphere. Using a personal computer, a user can generate a fission product inventory; decay and in-grow the inventory during transport through processes, facilities, and the environment; model the downwind dispersion of the activity; and calculate doses to downwind individuals. Internal dose from the inhalation and ingestion pathways is calculated. External dose from ground surface and plume gamma pathways is calculated. New and exciting updates to the program include the ability to evaluate a release to an enclosed room, resuspension of deposited activity and evaluation of a release up to 1 meter from the release point. Enhanced tools are included for dry deposition, building wake, occupancy factors, respirable fraction, AMAD adjustment, updated and enhanced radionuclide inventory and inclusion of the dose-conversion factors from FOR 11 and 12. 2 - Methods: RSAC6 calculates meteorological dispersion in the atmosphere using Gaussian plume diffusion for Pasquill-Gifford, Hilmeier-Gifford and Markee models. A unique capability is the ability to model Class F fumigation conditions, the meteorological condition that causes the highest ground level concentrations from an elevated release. Doses may be calculated for various pathways including inhalation, ingestion, ground surface, air immersion, water immersion pathways. Dose calculations may be made for either acute or chronic releases. Internal doses (inhalation and ingestion) are calculated using the ICRP-30 model with dose conversion factors from FOR 11. External factors are calculated using FOR 12. 3 - Unusual Features: RSAC6 calculates complete progeny in-growth and decay during all accident phases. The calculation of fission product inventories in particularly useful in the analysis of accidents where the

  17. Separation of caesium-137 from fission products using phosphotungstic acid

    International Nuclear Information System (INIS)

    Murthy, T.S.; Balasubramaniam, K.R.; Ananthakrishnan, M.; Varma, R.N.

    1977-01-01

    Separation of caesium 137 from fission products using phosphotungstic acid is reported. Phosphotungstate caesium is precipitated as caesium from fission product waste solution in acid medium and subsequently purified. Separation of phosphate and tungstate ions has been done using a typical hydrous oxide like alumina. The exchange capacity of alumina for phosphate and tungstate ions, and the purity of the product are determined. Results are discussed. Based on the findings a procedure is recommended for caesium 137 separation. (A.K.)

  18. Fission product yield measurements using monoenergetic photon beams

    Science.gov (United States)

    Krishichayan; Bhike, M.; Tonchev, A. P.; Tornow, W.

    2017-09-01

    Measurements of fission products yields (FPYs) are an important source of information on the fission process. During the past couple of years, a TUNL-LANL-LLNL collaboration has provided data on the FPYs from quasi monoenergetic neutron-induced fission on 235U, 238U, and 239Pu and has revealed an unexpected energy dependence of both asymmetric fission fragments at energies below 4 MeV. This peculiar FPY energy dependence was more pronounced in neutron-induced fission of 239Pu. In an effort to understand and compare the effect of the incoming probe on the FPY distribution, we have carried out monoenergetic photon-induced fission experiments on the same 235U, 238U, and 239Pu targets. Monoenergetic photon beams of Eγ = 13.0 MeV were provided by the HIγS facility, the world's most intense γ-ray source. In order to determine the total number of fission events, a dual-fission chamber was used during the irradiation. These irradiated samples were counted at the TUNL's low-background γ-ray counting facility using high efficient HPGe detectors over a period of 10 weeks. Here we report on our first ever photofission product yield measurements obtained with monoenegetic photon beams. These results are compared with neutron-induced FPY data.

  19. Fission product yield measurements using monoenergetic photon beams

    Directory of Open Access Journals (Sweden)

    Krishichayan

    2017-01-01

    Full Text Available Measurements of fission products yields (FPYs are an important source of information on the fission process. During the past couple of years, a TUNL-LANL-LLNL collaboration has provided data on the FPYs from quasi monoenergetic neutron-induced fission on 235U, 238U, and 239Pu and has revealed an unexpected energy dependence of both asymmetric fission fragments at energies below 4 MeV. This peculiar FPY energy dependence was more pronounced in neutron-induced fission of 239Pu. In an effort to understand and compare the effect of the incoming probe on the FPY distribution, we have carried out monoenergetic photon-induced fission experiments on the same 235U, 238U, and 239Pu targets. Monoenergetic photon beams of Eγ = 13.0 MeV were provided by the HIγS facility, the world's most intense γ-ray source. In order to determine the total number of fission events, a dual-fission chamber was used during the irradiation. These irradiated samples were counted at the TUNL's low-background γ-ray counting facility using high efficient HPGe detectors over a period of 10 weeks. Here we report on our first ever photofission product yield measurements obtained with monoenegetic photon beams. These results are compared with neutron-induced FPY data.

  20. Repeated radiation injuries by fission products

    International Nuclear Information System (INIS)

    Vasilenko, I.Ya.

    1986-01-01

    Attention is given to repeated radiation injuries during internal irradiation of theoretical and practical interest, particularly in case of the intake into organism of young products of nuclear fission (PNF). The results of experiments with dogs with repeated radioactive iodine injury the isotopes of which (131-135sub(I)) constitute a considerable part of PNF activity are discussed. The blood reaction and protein metabolism state have been studied. Observations for dogs have been continued for about 4 years. The doses for thyroid, gastrointestinal tract and liver subjected to the most intensive irradiation consituted in the first series of experiments after the first intake about 3;0.3;0.05 Gy, after the second - 5;0.5;0.08 Gy and in the second series of experiments - 3;0.3;0.05 Gy and 0.6;0.06;0.01 Gy, respectively. Hematologic factors,thyroid function, changes in exchange and immunologic reactivity have been studied. The dogs have been under observation for 5 years. It is shown in case of repeated intake of Isup(131) PNF into animals organism in quantity which does not cause during the acute period a clinically outlined sickness, substantial differences in the organism reaction as compared with the first intake of radionuclides have not been found. The presence of residual radiation injuries did not cause charging action during the acute period during PNF and repeated intake which in the author's opinion testifies to perfection of compensator mechanisms in case of intake of such quantities of radioactive products. At the remote periods blastomogenic action manifested which is estimated as a result of general biological action of radionuclides administered to the organism. The necessity in subsequent investigations for obtaining the data on organism reactivity, clinic and pathogenesis with the aim of prophylaxis and treatment of such injuries is indicated

  1. A low power ADS for transmutation studies in fast systems

    Science.gov (United States)

    Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

    2017-12-01

    In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

  2. New infrastructure for studies of transmutation and fast systems concepts

    Science.gov (United States)

    Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

    2017-09-01

    In this work we report initial studies on a low power Accelerator-Driven System as a possible experimental facility for the measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

  3. Evaluation of alternative partitioning/transmutation scenarios using transmutation in light-water reactors (LWRs)

    International Nuclear Information System (INIS)

    Collins, E.D.; Renier, J.P.; Del Cul, B.; Spencer, B.

    2005-01-01

    Previous Advanced Fuel Cycle Initiative (AFCI) studies were made to assess the effects of the existing accumulation of LWR spent fuel in the United States on the capability to partition/transmute actinides using existing and advanced LWR. The concept of treating the oldest fuel first indicated that significant advantages could be gained in both partitioning, transmutation, and in overall cost reduction. The processing scenarios previously evaluated assumed that (1) 2000 MT/year of spent fuel, irradiated to 45 GWd/MT and decayed for 30 years is processed; (2) recovered plutonium and 90% of the neptunium are transmuted in LWR MOX fuel; and (3) minor actinides, consisting of americium, curium, and 10% of the neptunium are transmuted in burnable poison type targets. Results of the previous study showed that significant benefits could be obtained, including (1) lower costs for partitioning and transmutation and for storage of spent fuel, (2) maintenance of proliferation resistance for the fissile plutonium in spent fuels, and (3) extended lifetime for the repository. The lower costs would be achieved primarily because no capital investment for a special transmuter reactor (fast reactor, accelerator-driven system, etc.) would be required. Instead, only existing and new LWRs would be utilized. Moreover, no new storage capacity could be needed for spent fuels and irradiated targets because the number of spent fuel assemblies would remain the same after the scenario is begun. Even though the total inventory of plutonium would rise during the early cycles, ∼98% of the plutonium would be contained in stored spent fuel and would be protected by high radiation (the Spent Fuel Standard). This is because the spent fuel would be reprocessed and re-irradiated at intervals within which the fission products, 137 Cs and 90 Sr, both with half-lives of ∼ 30 years, exist in significantly high concentrations.The lifetime of the repository would be extended significantly because all of

  4. The possible transmutation of radioactive waste from nuclear reactors

    International Nuclear Information System (INIS)

    Harries, J.R.

    1974-01-01

    A nuclear reactor power program produces high level and long lived radioactive wastes. The high level activity is associated with fission products, but beyond 400 years the principal waste hazard is from transuranic elements produced in the reactor. Several schemes have been proposed for the transmutation of the problem isotopes into more easily handled isotopes. The neutron flux in a thermal reactor is not high enough to significantly reduce the longer lived fission product isotopes 90 Sr and 132 Gs, but the transuranic elements can be reduced by recycling through power reactors. The limitation on recycling of the transuranic elements is the separation process to remove trace quantities from the waste stream. In fast reactors the transuranic elements are the principal fuel and fast reactor waste contains only half as much 90 Sr as thermal reactors. However, the overall waste hazard is similar to thermal reactors. A sufficiently intense neutron flux for fission product transmutation could perhaps be produced by a spallation reactor driven by a proton linear accelerator or a controlled thermonuclear reactor. However, both concepts are still some years in the future. Transmutation by accelerator sources of protons, electrons of gammas tend to require more energy than neutron transmutation. (author)

  5. Transmutation of technetium into stable ruthenium in high flux conceptual research reactor

    International Nuclear Information System (INIS)

    Amrani, N.; Boucenna, A.

    2007-01-01

    The effectiveness of transmutation for the long lived fission product technetium-99 in high flux research reactor, considering its large capture cross section in thermal and epithermal region is evaluated. The calculation of Ruthenium concentration evolution under irradiation was performed using Chain Solver 2.20 code. The approximation used for the transmutation calculation is the assumption that the influence of change in irradiated materials structures on the reactor operator mode characteristics is insignificant. The results on Technetium transmutation in high flux research reactor suggested an effective use of this kind of research reactors. The evaluation brings a new concept of multi-recycle Technetium transmutation using HFR T RAN (High Flux Research Reactor for Transmutation)

  6. Feasibility of waste transmutation using accelerator-driven IRIS subcritical system

    International Nuclear Information System (INIS)

    Petroviae, B.; Carelli, M.; Paramonov, D.

    2001-01-01

    Waste transmutation is considered for reducing radio-toxicity of nuclear waste generated in power reactors. Accelerator driven subcritical systems (ADS) offer certain advantages over the use of nuclear reactors. Transmutation of fission products (e.g. 99 Tc) generally requires thermal neutron spectrum, while for actinides fast spectrum provides better performance. Proposed solutions to this problem include a multi-strata approach as well as a multi-zone (thermal/fast-spectrum) single systems. In this paper we examine the feasibility of employing a dual-spectrum two-zone accelerator-driven IRIS subcritical for waste transmutation. (author)

  7. Conceptual study on high performance blanket in a spherical tokamak fusion-driven transmuter

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Yican

    2000-01-01

    A preliminary conceptual design on high performance dual-cooled blanket of fusion-driven transmuter is presented based on neutronic calculation. The dual-cooled system has some attractive advantages when utilized in transmutation of HLW (High Level Wastes). The calculation results show that this kind of blanket could safely transmute about 6 ton minor actinides (produced by 170 GW(e) Year PWRs approximately) and 0.4 ton fission products per year, and output 12 GW thermal power. In addition, the variation of power and critical factor of this blanket is relatively little during its 1-year operation period. This blanket is also tritium self-sustainable

  8. RF system considerations for accelerator production of tritium and the transmutation of nuclear waste

    International Nuclear Information System (INIS)

    Tallerico, P.J.; Lynch, M.T.

    1993-01-01

    RF driven proton accelerators for the transmutation of nuclear waste (ATW) or for the production of tritium (APT) require unprecedented amounts of CW RF power at UHF frequencies. For both systems, the baseline design is for 246 MW at 700 MHz and 8,5 MW at 350 MHz. The main technical challenges are how to design and build such a large system so that it has excellent reliability, high efficiency, and reasonable capital cost. The issues associated with the selection of the RF amplifier and the sizes of the power supplies are emphasized in this paper

  9. Transport of fission products in matrix and graphite

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1983-06-01

    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)

  10. Analytical measurements of fission products during a severe nuclear accident

    Science.gov (United States)

    Doizi, D.; Reymond la Ruinaz, S.; Haykal, I.; Manceron, L.; Perrin, A.; Boudon, V.; Vander Auwera, J.; tchana, F. Kwabia; Faye, M.

    2018-01-01

    The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d'Investissement d'Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

  11. Analytical measurements of fission products during a severe nuclear accident

    Directory of Open Access Journals (Sweden)

    Doizi D.

    2018-01-01

    Full Text Available The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d’Investissement d’Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

  12. Interactions of fission product vapours with aerosols

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  13. Interactions of fission product vapours with aerosols

    International Nuclear Information System (INIS)

    Benson, C.G.; Newland, M.S.

    1996-01-01

    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 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

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

    International Nuclear Information System (INIS)

    Besson, A.; Prigent, Y.; Van-Kote, F.

    1969-01-01

    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) [fr

  15. Fission product detection by means of photovoltaic cells

    International Nuclear Information System (INIS)

    Liatard, E.; Akrouf, S.; Bruandet, J.F.; Fontenille, A.; Glasser, F.; Stassi, P.; Tsan Ung Chan

    1988-01-01

    The response of photovoltaic cells to heavy ions and fission products have been tested in-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 12 ns (FWHM) has been measured between two cells. (orig.)

  16. ENDF/B-5 Fission Products Library 1979

    International Nuclear Information System (INIS)

    Schwerer, O.; Lemmel, H.D.

    1981-10-01

    This document summarizes contents and documentation of the 1979 version of the Fission Products File of the ENDF/B Library maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

  17. Spray removal of fission products in PWR containments

    International Nuclear Information System (INIS)

    Grist, D.R.

    1982-11-01

    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)

  18. ENDF/B-5 Fission Products Library. Rev. 2

    International Nuclear Information System (INIS)

    Schwerer, O.; Pronyaev, V.G.; Lemmel, H.D.

    1984-07-01

    This document summarizes contents and documentation of the 1984 version of the Fission Products Nuclear Data File of the ENDF/B-5 Library (Rev. 2) maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

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

    International Nuclear Information System (INIS)

    Kodaira, Hideki

    1990-01-01

    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

  20. Potential of cyclotron based accelerators for energy production and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Stammbach, T.; Adam, S.; Fitze, H.R. [Paul Scherrer Institute, Villigen (Switzerland)] [and others

    1995-10-01

    PSI operates a 590 MeV-cyclotron facility for high intensity proton beams for the production of intense beams of pions and muons. The facility, commissioned in 1974, has been partially upgraded and is now operated routinely at a beam current of 1 mA, which corresponds to a beam power of 0.6 MW. At this current, the beam losses in the cyclotron are about 0.02%. By the end of 1995 the authors expect to have 1.5 mA of protons. Extensive theoretical investigations on beam current limitations in isochronous cyclotrons were undertaken. They show that the longitudinal space charge effects dominate. Based on their experience the authors present a preliminary design of a cyclotron scheme that could produce a 10 MW beam as a driver for an {open_quotes}energy amplifier{close_quotes} as proposed by C. Rubbia and his collaborators. The expected efficiency for the conversion of AC into beam power would be about 50% (for the RF-systems only). The beam losses in the cyclotron are expected to be a few {mu}A, leading to a tolerable activation level.

  1. Yields of fission products produced by thermal-neutron fission of 229Th

    International Nuclear Information System (INIS)

    Dickens, J.K.; McConnell, J.W.

    1983-01-01

    Absolute yields have been determined for 47 gamma rays emitted in the decay of 37 fission products representing 25 mass chains created during thermal-neutron fission of 229 Th. Using a Ge(Li) detector, spectra were obtained of gamma rays emitted between 15 min and 0.4 yr after very short irradiations by thermal neutrons of a 15-μg sample of 229 Th. On the basis of measured gamma-ray yields and known nuclear data, yields for cumulative production of 37 fission products were deduced. The absolute overall normalization uncertainty is 235 U, we postulate a simple functional dependence sigma = sigma(Z/sub p/), and using this dependence obtain values of Z/sub p/(A) for 15 mass chains created during fission of 229 Th. Values of Z/sub p/(A) were estimated for other mass chains based upon results of a recent study of Z/sub p/(A). Charge distributions determined using the deduced mass distribution and the deduced sets of Z/sub p/(A) and sigma(Z/sub p/) are in very good agreement with recent measurements, exhibiting a pronounced even-odd effect in elemental yields. These results may be used to predict unmeasured yields for 229 Th fission

  2. Calculation of displacement, gas, and transmutation production in stainless steel irradiated with spallation neutrons

    International Nuclear Information System (INIS)

    Wechsler, M.A.; Ramavarapu, R.; Daugherty, E.L.; Palmer, R.C.; Bullen, D.B.; Sommer, W.F.

    1993-01-01

    Calculations using the high-energy transport code LAHET have been made for the production of displacements, helium gas, and transmuted atoms for stainless steel (Fe-18 wt % Cr-10 wt % Ni) irradiated with spallation neutrons at energies of 100 to 1600 MeV. The damage energy cross section increased from about 250 to 350 b keV for increasing neutron energies from 100 to 1600 MeV with a spallation spectrum average of 281 barns-keV. For a displacement threshold energy of 33 eV, the corresponding spectrum-average displacement cross section is 3400 barns. The PKA spectrum was found to be fairly independent of the incident neutron energy, with an average damage energy of 0.25--0.30 MeV. The helium production cross section increased monotonically with increasing neutron energy, with a spectrum average of 0.32 barns. The maximum transmutation yield was observed near manganese (Z = 25), corresponding to a production cross section of about 0.2 barns. Relevance to fusion materials is discussed

  3. Attachment behavior of fission products to solution aerosol

    Energy Technology Data Exchange (ETDEWEB)

    Takamiya, Koichi; Tanaka, Toru; Nitta, Shinnosuke; Itosu, Satoshi; Sekimoto, Shun; Oki, Yuichi; Ohtsuki, Tsutomu [Research Reactor Institute, Kyoto University, Osaka (Japan)

    2016-12-15

    Various characteristics such as size distribution, chemical component and radioactivity have been analyzed for radioactive aerosols released from Fukushima Daiichi Nuclear Power Plant. Measured results for radioactive aerosols suggest that the potential transport medium for radioactive cesium was non-sea-salt sulfate. This result indicates that cesium isotopes would preferentially attach with sulfate compounds. In the present work the attachment behavior of fission products to aqueous solution aerosols of sodium salts has been studied using a generation system of solution aerosols and spontaneous fission source of {sup 248}Cm. Attachment ratios of fission products to the solution aerosols were compared among the aerosols generated by different solutions of sodium salt. A significant difference according as a solute of solution aerosols was found in the attachment behavior. The present results suggest the existence of chemical effects in the attachment behavior of fission products to solution aerosols.

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

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.; Klisinska, M.

    1997-01-01

    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 UO 2 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)

  5. Chemical separations schemes for partitioning and transmutation systems

    International Nuclear Information System (INIS)

    Laidler, J.

    2002-01-01

    In the initial phase of the U.S. Accelerator Transmutation of Waste (ATW) program, a single-tier system was foreseen in which the transuranics and long-lived fission products (specifically, 99 Tc and 129 I) recovered from spent LWR oxide fuel would be sent directly to an accelerator-driven transmuter reactor [1]. Because the quantity of fuel to be processed annually was so large (almost 1,500 tons per year), an aqueous solvent extraction process was chosen for LWR fuel processing. Without the need to separate transuranics from one another for feed to the transmuter, it became appropriate to develop an advanced aqueous separations method that became known as UREX. The UREX process employs an added reagent (acetohydroxamic acid) that suppresses the extraction of plutonium and promotes the extraction of technetium together with uranium. Technetium can then be efficiently removed from the uranium; the recovered uranium, being highly decontaminated, can be disposed of as a low-level waste or stored in an unshielded facility for future use. Plutonium and the other transuranic elements, plus the remaining fission products, are directed to the liquid waste stream. This stream is calcined, converting the transuranics and fission products to their oxides. The resulting oxide powder, now representing only about four percent of the original mass of the spent fuel, is reduced to metallic form by means of a pyrometallurgical process. Subsequently, the transuranics are separated from the fission products in another pyro-metallurgical step involving molten salt electrorefining

  6. Transmutation of LWR waste actinides in thermal reactors

    International Nuclear Information System (INIS)

    Gorrell, T.C.

    1979-01-01

    Recycle of actinides to a reactor for transmutation to fission products is being considered as a possible means of waste disposal. Actinide transmutation calculations were made for two irradiation options in a thermal (LWR) reactor. The cases considered were: all actinides recycled in regular uranium fuel assemblies, and transuranic actinides recycled in separate mixed oxide (MOX) assemblies. When all actinides were recycled in a uranium lattice, a reduction of 62% in the transuranic inventory was achieved after 10 recycles, compared to the inventory accumulated without recycle. When the transuranics from 2 regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after 5 recycles

  7. Transmutation potential of current and innovative nuclear power systems

    International Nuclear Information System (INIS)

    Slessarev, I.; Salvatores, M.; Uematsu, M.

    1993-01-01

    In the present paper we have investigated the transmutation potential of different nuclear systems from a physical point of view. Transuranium (TRU) elements have been considered, but also long lived fission products (LLFP). The potential for transmutation has to take into account not only the consumption of a specific nucleus (or of a specific 'family' of nuclei), but also the reproduction of other nuclei of higher masses. The present study allows an intercomparison taking into account both aspects. Technological, safety and design constraints were not considered at this stage. However strategic indications for future studies have been obtained. 3 refs., 3 tabs

  8. Development of Commercial-scale Fission Mo-99 Production System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung-Kon; Lee, Suseung; Hong, Soon-Bog; Jang, Kyung-Duk; Park, Ul Jael; Lee, Jun Sig [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    These days, worldwide {sup 99} Mo supply is not only insufficient but also unstable. Because, most of the main {sup 99}Mo production reactors are more than years old and suffered from frequent and unscheduled shutdown. Therefore, movement to replace old reactors to keep stable supply is now active. Under these conditions, KAERI (Korea Atomic Energy Research Institute) is developing LEU-based fission {sup 99}Mo production process which is connected to the new research reactor (Kijang New Research Reactor, KJRR), which is being constructed in Gijang, Busan, Korea. Historically, the most fission {sup 99}Mo producers have been used highly enriched uranium (HEU) targets so far. However, to reduce the use of HEU in private sector for non-proliferation, {sup 99}Mo producers are forced to convert their HEU-based process to use low enriched uranium (LEU) targets. Economic impact of a target conversion from HEU to LEU is significant. In this study, fission {sup 99}Mo process with non-irradiated LEU targets was presented except separation and purification steps. Pre- and post-irradiation tests of the fission {sup 99}Mo target will be done in 4th quarter of 2016. For the fission Mo production process development, hot experiments with irradiated LEU targets will be done in 4th quarter of 2016. Then, verification of the production process with quality control will be followed until the commercial production of fission {sup 99}Mo scheduled in 2019.

  9. Neutronic Parametric Study on a Conceptual Design for a Transmutation Fusion Blanket

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2011-01-01

    Fusion energy may be the one of options of future energy. In all over the world, researchers are putting their efforts for its commercial and economical availability. Fusion-fission hybrid reactors have been studied for various applications in China. First milestone of fusion energy is expected to be the fusion fission hybrid reactors. In fusion-fission hybrid reactor the blanket design is of second prime importance after fusion source. In this study conceptual design of a fusion blanket is initiated for calculation of tritium production, transmutation of minor actinides (MA) and fission products (FP) and energy multiplication calculations

  10. Transmutation of radioactive waste: Effect on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Rasmussen, N.C.; Pigford, T.H.

    1997-01-01

    A committee of the National Research Council reviewed three concepts for transmuting radionuclides recovered from the chemical reprocessing of commercial light-water-reactor (LWR) fuel: LWR transmutation reactors fueled with recycled actinides, advanced liquid-metal reactors (ALMRs), and accelerator-driven subcritical reactors for transmutation of waste (ATW). The concepts were evaluated in terms of: (1) the extent to which waste disposal would benefit from transmutation, (2) time required to reduce the total inventory of radionuclides in the waste and fuel cycle, (3) the complexity of the overall transmutation system, (4) the extent of new development required, and (5) institutional and economic problems of operating such systems. Transmutation could affect geologic disposal of waste by reducing the inventory of transuranics (TRUs), fission products, and other radionuclides in the waste. Reducing the inventory of transuranics does not necessarily affect radiation doses to people who use contaminated ground water if the dissolution rate of transuranics in waste is controlled by elemental solubilities. However, reducing inventories of Am and Pu would decrease potential hazards from human intrusion. The likelihood for underground nuclear criticality would also be reduced. The long-lived fission products Tc-99, I-129, Cs-135 and others typically contribute most to the long-term radiation doses to future populations who use contaminated water from the repository. Their transmutation requires thermal or epithermal neutrons, readily available in LWR and ATW transmutors. ALMR and LWR transmutors would require several hundred years to reduce the total transuranic inventory by even a factor of 10 at constant electric power, and thousands of years for a hundred-fold reduction. For the same electrical power, the ATW could reduce total transuranic inventory about tenfold more rapidly, because of its very high thermal-neutron flux. However, extremely low process losses would be

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  12. Partitioning and transmutation: Radioactive waste management option

    International Nuclear Information System (INIS)

    Stanculescu, A.

    2005-01-01

    Growing world population with increasing energy needs, especially in the developing countries, Threat of global warming due to CO 2 emissions demands non-fossil electricity production. Nuclear will have to be part of a sustainable mix of energy production options Figures show that 350 GWe worldwide capacity is 'nuclear'. Present worldwide spent fuel (containing high Pu inventory) and HLW would need large repositories. In view of the previous facts this lecture deals Partitioning and transmutation as radioactive waste management option. Partitioning and transmutation (P and T) is a complex technology i.e. advanced reprocessing, and demand transuranics fuel fabrication plants, as well as innovative and/or dedicated transmutation reactors. In addition to U, Pu, and 129 I, 'partitioning' extracts from the liquid high level waste the minor actinides (MA) and the long-lived fission products (LLFP) 99-Tc, 93-Zr, 135-Cs, 107-Pd, and 79-Se). 'Transmutation' requires fully new fuel fabrication plants and reactor technologies to be developed and implemented on industrial scale. Present LWRs are not suited for MA and LLFP transmutation (safety consideration, plant operation, poor incineration capability). Only specially licensed LWRs can cope with MOX fuel; for increased Pu loadings (up to 100%), special reactor designs (e.g., ABB80+) are required; a combination of these reactor types could allow Pu inventory stabilization. Long-term waste radiotoxicity can be effectively reduced only if transuranics are 'incinerated' through fission with very hard neutron spectra. New reactor concepts (dedicated fast reactors, Accelerator Driven Systems (ADS), fusion/fission hybrid reactors) have been proposed as transmuters/incinerators. Significant Pu+MAs incineration rates can be achieved in symbiotic scenarios: LWR-MOX and dedicated fast reactors; fast neutron spectrum ADS mainly for MA incineration; very high thermal flux ADS concepts could also provide a significant transuranics

  13. Radioanalytical prediction of radiative capture in 99Mo production via transmutation adiabatic resonance crossing by cyclotron

    CERN Document Server

    Khorshidi, Abdollah; Pazirandeh, Ali; Tenreiro, Claudio; Kadi, Yacine

    2014-01-01

    In this study, the transmutation adiabatic resonance crossing (TARC) concept was estimated in Mo-99 radioisotope production via radiative capture reaction in two designs. The TARC method was composed of moderating neutrons in lead or a composition of lead and water. Additionally, the target was surrounded by a moderator assembly and a graphite reflector district. Produced neutrons were investigated by (p,xn) interactions with 30 MeV and 300 mu A proton beam on tungsten, beryllium, and tantalum targets. The Mo-99 production yield was related to the moderator property, cross section, and sample positioning inside the distinct region of neutron storage as must be proper to achieve gains. Gathered thermal flux of neutrons can contribute to molybdenum isotope production. Moreover, the sample positioning to gain higher production yield was dependent on a greater flux in the length of thermal neutrons and region materials inside the moderator or reflector. When the sample radial distance from Be was 38 cm inside the...

  14. Partitioning and Transmutation of minor actinides

    International Nuclear Information System (INIS)

    Koch, L.; Wellum, R.

    1991-01-01

    The partitioning of minor actinides from spent fuels and their transmutation into short-lived fission products has been the topic of two dedicated meetings organized jointly by the European Commission and the OECD. The conclusion of the last meeting in 1980, in short, was that partitioning and transmutation of minor actinides, especially in fast reactors, seemed possible. However, the incentive, which would be a reduction of the radiological hazard to the public, was too small if long-lived fission products were not included. Furthermore this meeting showed that minor actinide targets or possible nuclear fuels containing minor actinides for transmutation had not yet been developed. The European Institute for Transuranium Elements took up this task and has carried it out as a small activity for several years. Interests expressed recently by an expert meeting of the OECD/NEA (Paris, 25 April 1989), which was initiated by the proposed Japanese project Omega, led us to the conclusion that the present state of knowledge should be looked at in a workshop environment. Since the Japanese proposal within the project Omega is based on a broader approach we needed this evaluation to assess the relevance of our present activity and wanted to identifiy additional studies which might be needed to cover possible future demands from the public. This workshop was therefore organized, and participants active in the field from EC countries, the USA and Japan were invited

  15. Recent progress in fission product separation

    International Nuclear Information System (INIS)

    Raggenbass, A.

    1964-01-01

    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

  16. Fission-product releases from a PHWR terminal debris bed

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Bailey, D.G., E-mail: morgan.brown@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model. (author)

  17. Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system

    International Nuclear Information System (INIS)

    Davidson, J.W.; Battat, M.E.

    1993-01-01

    Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies

  18. Transmutation of radioactive wastes from nuclear power plants. A contribution to the reduction of the final repository problem; Transmutation radioaktiver Reststoffe aus Kernkraftwerken. Ein Beitrag zur Verringerung der Endlagerproblematik

    Energy Technology Data Exchange (ETDEWEB)

    Mach, Manfred [Technische Univ. Berlin (Germany). Inst. fuer Technologie und Management

    2015-07-01

    The brochure on transmutation of radioactive wastes from nuclear power plants - a contribution to the reduction of the final repository problem covers the following issues: What is transmutation? Nuclear power in Germany; energy density of fuels; time span of energy resources; CO{sub 2} emissions from different energy sources; types of nuclear power plants in Germany; cost of German electricity generation plants; nuclear power plants worldwide; wastes from nuclear electricity production; radiation from fission products; radiation effects on humans, the nuclear fuel cycle, direct final disposal of radioactive wastes; risk assessment of the direct final disposal; partitioning of actinides; transmutation of actinides.

  19. FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

    International Nuclear Information System (INIS)

    Pierce, V.H.

    2005-01-01

    1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

  20. Evaluation of systems incorporating transmutation for the reduction of the long term toxicity of high-level waste

    International Nuclear Information System (INIS)

    Davidson, J.W.

    1979-01-01

    One of the alternative high-level nuclear waste (HLW) management/disposal concepts proposed involves the separation from HLW of the elements with isotopes which dominate the radiotoxicity and the transmutation of these nuclides to shortlived or stable products. The waste management system required for transmutation employs chemical processing of HLW to recover waste nuclides for irradiation with neutrons in a transmutation device. The transmuter periodically requires replenishment of the target nuclides and chemical processing to remove the transmutation products. The waste streams from HLW processing and product recovery together comprise the discharge from the system. An imploding liner fusion reactor (ILFR) is assumed for the transmuter with the waste nuclides dissolved in a molten lead-lithium alloy blanket. The potential transmutation candidates are defined as the elements with toxicities per unit volume (toxicity indexes) in solidified HLW at 1000 years which are greater than that for 0.2% uranium ore (carnotite). The candidates which require separation for transmutation are the actinides; Np, Pu, Am, and Cu and the fission products; I and Tc. Certain assumptions were made for the parameters for the ILFR and its operating conditions, and a system evaluation was done. System evaluations indicate that blanket waste loadings on the order of several percent of the total concentration result in attractive performance in terms of high transmutation capacities and low blanket processing requirments. It appears that transmutation system goals in terms of toxicity reduction are achievable with a modest number of transmuters. In addition, requirements for transmuter performance, chemical processing capacity and chemical separation efficiency appear to be within projected values for this technology

  1. Transmutation of radioactive nuclear waste

    International Nuclear Information System (INIS)

    Toor, A; Buck, R

    2000-01-01

    Lack of a safe disposal method for radioactive nuclear waste (RNW) is a problem of staggering proportion and impact. A typical LWR fission reactor will produce the following RNW in one year: minor actinides (i.e. 237 Np, 242-243 Am, 243-245 Cm) ∼40 kg, long-lived fission products (i.e, 99 Tc, 93 Zr, 129 I, 135 Cs) ∼80 kg, short lived fission products (e.g. 137 Cs, 90 Sr) ∼50kg and plutonium ∼280 kg. The total RNW produced by France and Canada amounts to hundreds of metric tonnes per year. Obtaining a uniform policy dealing with RNW has been blocked by the desire on one hand to harvest the energy stored in plutonium to benefit society and on the other hand the need to assure that the stockpile of plutonium will not be channeled into future nuclear weapons. In the meantime, the quantity and handling of these materials represents a potential health hazard to the world's population and particularly to people in the vicinity of temporary storage facilities. In the U.S., societal awareness of the hazards associated with RNW has effectively delayed development of U.S. nuclear fission reactors during the past decade. As a result the U.S. does not benefit from the large investment of resources in this industry. Reluctance to employ nuclear energy has compelled our society to rely increasingly on non-reusable alternative energy sources; coal, oil, and natural gas. That decision has compounded other unresolved global problems such as air pollution, acid rain, and global warming. Relying on these energy sources to meet our increasing energy demands has led the U.S. to increase its reliance on foreign oil; a policy that is disadvantageous to our economy and our national security. RNW can be simplistically thought of as being composed of two principal components: (1) actinides with half lives up to 10 6 years and (2) the broad class of fission fragments with typical half lives of a few hundred years. One approach to the RNW storage problem has been to transmute the

  2. Separation of fission Molybdenum for production of technetium generator

    International Nuclear Information System (INIS)

    Bayat, L.; Shaham, V.; Davarkha, R.

    2002-01-01

    There are two basically different methods for Mo-99 productions: Activation of Mo-99 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 activity of the product. Idolisation of fission molybdenum from irradiated nuclear fuel targets which have undergone short-term cooling. Maximum fission yield 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 consist in the retention of the fission gases the recycling of non-consumed fuel and the treatment of the waste streams arising. This publication will deal with the individual steps in the process

  3. Trapping technology for gaseous fission products from voloxidation process

    International Nuclear Information System (INIS)

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

    2005-05-01

    The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, 14 C, Kr, Xe, I and 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 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 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

  4. 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.

  5. An Evaluation of a Fission Product Inventory for CANDU Fuels

    International Nuclear Information System (INIS)

    Jung, Jong Yeob; Park, Joo Hwan

    2007-01-01

    Fission products are released by two processes when a single channel accident occurs. One is a 'prompt release' and the other is a 'delayed release'. Prompt release assumes that the gap inventory of the fuel elements is released by a fuel element failure at the time of an accident. Delayed release assumes that the inventories within the grain or at the grain boundary are released after a accident due to a diffusion through grains, an oxidation of the fuel and an interaction between the fuel and the Zircaloy sheath. Therefore, the calculation of a fission product inventory and its distribution in a fuel during a normal operating is the starting point for the assessment of a fission product release for single channel accidents. In this report, the fission product inventories and their distributions within s fuel under a normal operating condition are evaluated for three types of CANDU fuels such as the 37 element fuel, CANFLEX-NU and CANFLEX-RU fuel bundles in the 'limiting channel'. To accomplish the above mentioned purposes, the basic power histories for each type of CANDU fuel were produced and the fission product inventories were calculated by using the ELESTRES code

  6. Evaluation and compilation of fission product yields 1993

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  7. 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.

  8. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

    1993-03-01

    The management of radioactive waste is one of the key issues in today's discussions on nuclear energy, especially the long term disposal of high level radioactive wastes. The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) would allow 'burning' of the associated extremely long life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. The International Working Group on Fast Reactors (IWGFR) decided to include the topic of actinide transmutation in liquid metal fast breeder reactors in its programme. The IAEA organized the Specialists Meeting on Use of Fast Breeder Reactors for Actinide Transmutation in Obninsk, Russian Federation, from 22 to 24 September 1992. The specialists agree that future progress in solving transmutation problems could be achieved by improvements in: Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes); technological research and development on the design, fabrication and irradiation of the minor actinides (MAs) containing fuels; nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners; demonstration of the feasibility of the safe and economic MA burner cores; knowledge of the impact of maximum tolerable amount of rare earths in americium containing fuels. Refs, figs and tabs

  9. Characterization of wastes from fission 99 Mo production

    International Nuclear Information System (INIS)

    Endo, L.S.; Dellamano, J.C.

    1992-07-01

    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)

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

    International Nuclear Information System (INIS)

    Martin Deidier, Loick.

    1979-12-01

    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 [fr

  11. MA-burners efficiency parameters allowing for the duration of transmutation process

    International Nuclear Information System (INIS)

    Gulevich, A.; Zemskov, E.; Kalugin, A.; Ponomarev, L.; Seliverstov, V.; Seregin, M.

    2010-01-01

    Transmutation of minor actinides (MA) means their transforming into the fission products. Usually, MA-burner's transmutation efficiency is characterized by the static parameters only, such as the number of neutrons absorbed and the rate of MA feeding. However, the proper characterization of MA-burner's efficiency additionally requires the consideration of parameters allowing for the duration of the MA transmutation process. Two parameters of that kind are proposed: a) transmutation time τ - mean time period from the moment a mass of MA is loaded into the burner's fuel cycle to be transmuted to the moment this mass is completely transmuted; b) number of reprocessing cycles n rep - effective number of reprocessing cycles a mass of loaded MA has to undergo before being completely transmuted. Some of MA-burners' types have been analyzed from the point of view of these parameters. It turned out that all of them have the value of parameters too high from the practical point of view. It appears that some new approaches to MA-burner's design have to be used to significantly reduce the value of these parameters in order to make the large-scale MA transmutation process practically reasonable. Some of such approaches are proposed and their potential efficiency is discussed. (authors)

  12. MA-burners efficiency parameters allowing for the duration of transmutation process

    Energy Technology Data Exchange (ETDEWEB)

    Gulevich, A.; Zemskov, E. [Institute of Physics and Power Engineering, Bondarenko Square 1, Obninsk, Kaluga Region 249020 (Russian Federation); Kalugin, A.; Ponomarev, L. [Russian Research Center ' ' Kurchatov Institute' ' Kurchatov Square 1, Moscow 123182 (Russian Federation); Seliverstov, V. [Institute of Theoretical and Experimental Physics ul.B. Cheremushkinskaya 25, Moscow 117259 (Russian Federation); Seregin, M. [Russian Research Institute of Chemical Technology Kashirskoe Shosse 33, Moscow 115230 (Russian Federation)

    2010-07-01

    Transmutation of minor actinides (MA) means their transforming into the fission products. Usually, MA-burner's transmutation efficiency is characterized by the static parameters only, such as the number of neutrons absorbed and the rate of MA feeding. However, the proper characterization of MA-burner's efficiency additionally requires the consideration of parameters allowing for the duration of the MA transmutation process. Two parameters of that kind are proposed: a) transmutation time {tau} - mean time period from the moment a mass of MA is loaded into the burner's fuel cycle to be transmuted to the moment this mass is completely transmuted; b) number of reprocessing cycles n{sub rep} - effective number of reprocessing cycles a mass of loaded MA has to undergo before being completely transmuted. Some of MA-burners' types have been analyzed from the point of view of these parameters. It turned out that all of them have the value of parameters too high from the practical point of view. It appears that some new approaches to MA-burner's design have to be used to significantly reduce the value of these parameters in order to make the large-scale MA transmutation process practically reasonable. Some of such approaches are proposed and their potential efficiency is discussed. (authors)

  13. Yields of fission products produced by thermal-neutron fission of 245Cm

    International Nuclear Information System (INIS)

    Dickens, J.K.; McConnell, J.W.

    1981-01-01

    Absolute yields have been determined for 105 gamma rays emitted in the decay of 95 fission products representing 54 mass chains created during thermal-neutron fission of 245 Cm. These results include 17 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays between 30 sec and 0.3 yr after very short irradiations of thermal neutrons on a 1 μg sample of 245 Cm. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 84 and 156. The absolute overall normalization uncertainty is 239 Pu and for 252 Cf(s.f.); the influences of the closed shells Z=50, N=82 are not as marked as for thermal-neutron fission of 239 Pu but much more apparent than for 252 Cf(s.f.). Information on the charge distribution along several isobaric mass chains was obtained by determining fractional yields for 12 fission products. The charge distribution width parameter, based upon data for the heavy masses, A=128 to 140, is independent of mass to within the uncertainties of the measurements. Gamma-ray assignments were made for decay of short-lived fission products for which absolute gamma-ray transition probabilities are either not known or in doubt. Absolute gamma-ray transition probabilities were determined as (51 +- 8)% for the 374-keV gamma ray from decay of 110 Rh, (35 +- 7)% for the 1096-keV gamma ray from decay of 133 Sb, and (21.2 +- 1.2)% for the 255-keV gamma ray from decay of 142 Ba

  14. Fission product removal from molten salt using zeolite

    International Nuclear Information System (INIS)

    Pereira, C.; Babcock, B.D.

    1996-01-01

    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

  15. Fission product released experiment of coated fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Shijiang, Xu; Bing, Yang; Chunhe, Tang; Junguo, Zhu; Jintao, Huang; Binzhong, Zhang [Inst. of Nucl. Energy Technology, Tsinghua Univ., Beijing (China); Jinghan, Luo [Inst. of Atomic Energy, Beijing (China)

    1992-01-15

    Four samples of coated fuel particles were irradiated in the Heavy-Water Research Reactor of the Institute of Atomic Energy. Each of them was divided into two groups and irradiated to the burn up of 0.394% fima and 0.788% fima in two static capsules, respectively. After irradiation and cooling, post irradiation annealing experiment was carried out, the release ratios of the fission product {sup 133}Xe and {sup 131}I were measured, they are in the order of 10{sup -6}{approx}10{sup -7}. The fission product release ratio of naked kernel was also measured under the same conditions as for the coated fuel particles, the ratio of the fission product release of the coated fuel particles and of the naked kernel was in the order of 10{sup -5}{approx}10{sup -4}.

  16. Map of calculated radioactivity of fission product, (4)

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1978-07-01

    The overall radioactivities of fission products depending on irradiation time and cooling time were calculated for 18 different neutron fluxes, which are presented in contour maps and tables. Irradiation condition etc. are the followings: neutron flux (n sub(th)) 1 x 10 12 - 6.8 x 10 14 n/cm 2 /sec, uranium quantity 1 mole (6 x 10 23 atoms, ca. 271 g UO 2 ), U-235 enrichment 2.7%, irradiation time 60. - 6 x 10 7 sec (1 min - 1.9 y), cooling time 0. and 60. - 6 x 10 7 sec (1 min - 1.9 y). The enrichment value represents those for LWRs. To calculate the overall radioactivities, 595 fission product nuclides were introduced. Overall radioactivities calculations were made for 68,000 combinations of irradiation time, cooling time and neutron flux. The many complex decay chains of fission products were treated with CODAC-No.6 computer code. (author)

  17. The release of fission products from uranium metal: a review

    International Nuclear Information System (INIS)

    Minshall, P.C.

    1989-03-01

    The literature on the release of fission products as gaseous species from irradiated uranium metal in oxidising atmospheres has been reviewed. Release of actinides and of fission products as spalled particulate were not considered. Data is given on the release in air, carbon dioxide, steam and mixtures of steam and air. The majority of data discussed lie between 800 and 1200 0 C though some results for xenon, krypton and iodine releases below 800 0 C are given. Two measures of fission product release are discussed: the release fraction, F(tot), which is the ratio of the total release to the initial inventory, and the fractional release, F(ox), which is the fraction released from the oxidised metal. The effect of burn-up, atmosphere and temperature on F(tot) and F(ox) is examined and the conditions under which the release fraction, F(tot) is proportional to the extent of oxidation discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    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. Refs, figs, tabs.

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

    International Nuclear Information System (INIS)

    1997-11-01

    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

  20. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

    International Nuclear Information System (INIS)

    Selby, H.D.; Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C.

    2010-01-01

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99 Mo, 95 Zr, 137 Cs, 140 Ba, 141,143 Ce, and 147 Nd. Modest incident-energy dependence exists for the 147 Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ∼5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except

  1. Properties of Fission-Product decay heat from Minor-Actinide fissioning systems

    International Nuclear Information System (INIS)

    Oyamatsu, Kazuhiro; Mori, Hideki

    2000-01-01

    The aggregate Fission-Product (FP) decay heat after a pulse fission is examined for Minor Actinide (MA) fissiles 237 Np, 241 Am, 243 Am, 242 Cm and 244 Cm. We find that the MA decay heat is comparable but smaller than that of 235 U except for cooling times at about 10 8 s (approx. = 3 y). At these cooling times, either the β or γ component of the FP decay heat for these MA's is substantially larger than the one for 235 U. This difference is found to originate from the cumulative fission yield of 106 Ru (T 1/2 = 3.2x10 7 s). This nuclide is the parent of 106 Rh (T 1/2 = 29.8 s) which is the dominant source of the decay heat at 10 8 s (approx. = 3 y). The fission yield is nearly an increasing function of the fissile mass number so that the FP decay heat is the largest for 244 Cm among the MA's at the cooling time. (author)

  2. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nichols, A.L.

    1990-09-01

    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)

  3. The chemistry of fission products for accident analysis

    International Nuclear Information System (INIS)

    Potter, P.E.

    1985-01-01

    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. The potential for large scale uses for fission product xenon

    International Nuclear Information System (INIS)

    Rohrmann, C.A.

    1983-01-01

    Of all fission products in spent, low enrichment, uranium, power reactor fuels xenon is produced in the highest yield - nearly one cubic meter, STP, per metric ton. In aged fuels which may be considered for processing in the U.S. radioactive xenon isotopes approach the lowest limits of detection. The separation from accompanying radioactive 85 Kr is the essential problem; however, this is state of the art technology which has been demonstrated on the pilot scale to yield xenon with pico-curie levels of 85 Kr contamination. If needed for special applications, such levels could be further reduced. Environmental considerations require the isolation of essentially all fission product krypton during fuel processing. Economic restraints assure that the bulk of this krypton will need to be separated from the much more voluminous xenon fraction of the total amount of fission gas. Xenon may thus be discarded or made available for uses at probably very low cost. In contrast with many other fission products which have unique radioactive characteristics which make them useful as sources of heat, gamma and x-rays and luminescence as well as for medicinal diagnostics and therapeutics fission product xenon differs from naturally occurring xenon only in its isotopic composition which gives it a slightly higher atomic weight, because of the much higher concentrations of the 134 X and 136 Xe isotopes. Therefore, fission product xenon can most likely find uses in applications which already exist but which can not be exploited most beneficially because of the high cost and scarcity of natural xenon. Unique uses would probably include applications in improved incandescent light illumination in place of krypton and in human anesthesia

  5. Separation and utilization of fission products considering economic aspects

    International Nuclear Information System (INIS)

    Beer, M.; Gorski, B.; Hennrich, M.; Pfrepper, G.; Richter, M.

    1982-01-01

    The quantity of usable fission products which will be obtained by nuclear fission till the year 2000 is estimated on the basis of prognostics for the development of nuclear energy in the world considering especially the development in the U.S.S.R. and the CMEA. The possibilities of utilization of cesium as gamma-ray source are discussed, and the present fields of application of palladium and the development of its price on the world market are shown. The fields of application of technetium, which wasn't available as artificial element in a greater quantity till now, have to be developed. The economic estimations base on data of a project for the separation of fission products in connection with a reprocessing plant, which was developed in the U.S.A. in 1978. The data show, that it is possible to produce the platinum metals and cesium with profit, the same can be expected for technetium. (author)

  6. Progress in fission product nuclear data. Information about activities in the field of measurements and compilations/evaluations of fission product nuclear data (FPND)

    International Nuclear Information System (INIS)

    Lammer, G.

    1978-07-01

    This is the fourth 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 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.)

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

    International Nuclear Information System (INIS)

    Lee, Jun Sig

    2013-01-01

    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

  8. Fission product behaviour in the primary circuit of an HTR

    International Nuclear Information System (INIS)

    Decken, C.B. von der; Iniotakis, N.

    1981-01-01

    The knowledge of fission product behaviour in the primary circuit of a High Temperature Reactor (HTR) is an essential requirement for the estimations of the availability of the reactor plant in normal operation, of the hazards to personnel during inspection and repair and of the potential danger to the environment from severe accidents. On the basis of the theoretical and experimental results obtained at the ''Institute for Reactor Components'' of the KFA Juelich /1/,/2/ the transport- and deposition behaviour of the fission- and activation products in the primary circuit of the PNP-500 reference plant has been investigated thoroughly. Special work had been done to quantify the uncertainties of the investigations and to calculate or estimate the dose rate level at different components of the primary cooling circuit. The contamination and the dose rate level in the inspection gap in the reactor pressure vessel is discussed in detail. For these investigations in particular the surface structure and the composition of the material, the chemical state of the fission products in the cooling gas, the composition of the cooling gas and the influence of dust on the transport- and deposition behaviour of the fission products have been taken into account. The investigations have been limited to the nuclides Ag-110m; Cs-134 and Cs-137

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

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1994-01-01

    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

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

    International Nuclear Information System (INIS)

    Rodliffe, R.S.

    1983-08-01

    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)

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

    International Nuclear Information System (INIS)

    Mathews, G.J.

    1983-10-01

    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

  12. Fission product release by fuel oxidation after water ingress

    International Nuclear Information System (INIS)

    Schreiber.

    1990-01-01

    On the basis of data obtained by a literature search, a computer code has been established for the calculation of the degree of oxidation of the fuel in the damaged fuel particles, and hence of the fission product release as a function of the time period of steam ingress. (orig.) [de

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

    International Nuclear Information System (INIS)

    Lammer, G.; Lammer, M.

    1980-06-01

    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

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

    International Nuclear Information System (INIS)

    Rudstam, G.; Aagaard, P.; Aleklett, K.; Lund, E.

    1981-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Roux, J.P.

    1976-03-01

    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 [fr

  16. Mo-99 production by fission and future projections

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  17. Reactivity effects of fission product decay in PWRs

    International Nuclear Information System (INIS)

    Aragones, J.M.; Ahnert, C.

    1988-01-01

    The purpose of the work reported in this paper is to analyze the effects of fission product chains with radioactive decay on the reactivity in pressurized water reactor (PWR) cores, calculating their accumulation and absorption rates along fuel burnup at continuous operation and after shutdown periods extending from 1 day to a few months. The authors PWR version of the WIMS-D4 code is first used to obtain the individual number densities, absorption rates, and averaged cross sections for every nuclide of the fission product chains with significant decay rates, as a function of fuel burnup at continuous irradiation. Next, by an auxiliary ad hoc code, these data, have been processed together with the required one for fissile nuclides and boron, also taken from WIMS at each burnup step, to calculate the average or effective values relevant for the analysis and the decay and change in overall absorption after several shutdown times. (1) The reactivity effect of fission product decay changes significantly with the shutdown time. The maximum absorption increase by decay is reached in ∼ 10 days' shutdown. (2) The dependence with fuel type, enrichment, and burnup is slight, but the change with previous power density is nearly linear, which might be significant after coast-down in previous cycles. (3) For long shutdown periods, the overall reactivity effect of decay in the three fission product chains considered is much less than if only the samarium peak due to 149 Nd is considered

  18. 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.

  19. Yields of fission products produced by thermal-neutron fission of 249Cf

    International Nuclear Information System (INIS)

    Dickens, J.K.; McConnell, J.W.

    1981-01-01

    Absolute yields have been determined for 107 gamma rays emitted in the decay of 97 fission products representing 54 mass chains created during thermal-neutron fission of 249 Cf. These results include 14 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays emanating from a 0.4 μg sample of 249 Cf between 45 s and 0.4 yr after very short irradiations of the 249 Cf by thermal neutrons. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 89 and 156. The absolute overall normalization uncertainty is approx.8%. The measured A-chain cumulative yields make up 77% of the total light mass (A 249 Cf

  20. A proposal for a Los Alamos international facility for transmutations (LIFT)

    International Nuclear Information System (INIS)

    Venneri, F.; Williamson, M.A.; Li, Ning; Doolen, G.

    1996-01-01

    The major groups engaged in transmutation research are converging towards a common objective and similar technology. It is now possible to envision an international program of research aimed at the destruction of reactor-generated (and other) nuclear waste using a series of multipurpose experimental facilities in the near future. Los Alamos National Laboratory, as the home of the highest power LINAC and a very active transmutation technology project, is the ideal host for the first of such facilities. The next step in the international program (a facility 10 times more powerful, for engineering-scale demonstrations) could be built in Europe, where there is substantial interest in the construction of such a device in the framework of international cooperation. A series of experiments at Las Alamos could explore the key transmutation technologies. Liquid lead loops, a liquid lead spallation target, and a large size liquid lead facility with provision for irradiation, cooling and diagnostics of several types of 'transmutation assemblies', where different transmutation concepts will be tested in different media and environments, from transmutation of fission products to destruction by fission of higher actinides, to other waste management applications. The engineering-scale facility, which will follow the initial testing phase, will extend the best concepts to full scale implementation

  1. FISA-2009 Conference on Euratom Research and Training Activities: Nuclear Fission - Past, Present and Future (Generation-II, -III and -IV + Partitioning and Transmutation)

    International Nuclear Information System (INIS)

    Bhatnagar, V.; Deffrennes, M.; Hugon, M.; Manolatos, P.; Ptackova, K.; Van Goethem, G.; Webster, S.

    2011-01-01

    This paper is an introduction to the research and training activities carried out under the Euratom 7th Framework Programme (FP7, 2007-2011) in the field of nuclear fission science and technology, covering in particular nuclear systems and safety, and including innovative reactor systems and partitioning and transmutation. It is based on the more than 40 invited lectures that were delivered by Euratom project coordinators and keynote speakers at the FISA-2009 Conference (), organised by the European Commission DG Research, 22-24 June 2009, Prague, Czech Republic. The Euratom programme must be considered in the context of current and future nuclear technology and the respective research effort: ·Generation-II (i.e. yesterday, NPP construction 1970-2000): safety and reliability of nuclear facilities and energy independence in order to ensure security of supply worldwide; ·Generation-III (i.e. today, construction 2000-2040+): continuous improvement of safety and reliability, and increased industrial competitiveness in a growing energy market; ·Generation-IV (i.e. tomorrow, construction from 2040) for increased sustainability though optimal utilisation of natural resources and waste minimisation, and increased proliferation resistance. Consequently, the focus of the lectures devoted to Generation-II and -III is on the major scientific challenges and technological developments needed to guarantee safety and reliability, in particular issues associated with plant lifetime extension and operation. The focus of the lectures devoted to Generation-IV is on the design objectives and associated research issues that have been agreed upon internationally, in particular the ambitious criteria and technology goals established at the international level by the Generation-IV International Forum (GIF). In the future, electricity must continue to be produced competitively, and in addition high temperature process heat may also be required, while exploiting a maximum of fissile and

  2. The wastes of nuclear fission

    International Nuclear Information System (INIS)

    Doubre, H.

    2005-01-01

    In this paper the author presents the problems of the radioactive wastes generated by the nuclear fission. The first part devoted to the fission phenomenon explains the incident neutron energy and the target nuclei role. The second part devoted to the nuclear wastes sources presents the production of wastes upstream of the reactors, in the reactors and why these wastes are dangerous. The third part discusses the radioactive wastes management in France (classification, laws). The last part details the associated research programs: the radionuclides separation, the disposal, the underground storage, the transmutation and the thorium cycle. (A.L.B.)

  3. Accelerator-driven transmutation of spent fuel elements

    Science.gov (United States)

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    2002-01-01

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  4. Spallation reaction study for fission products in nuclear waste: Cross section measurements for 137Cs, 90Sr and 107Pd on proton and deuteron

    Directory of Open Access Journals (Sweden)

    Wang He

    2017-01-01

    Full Text Available Spallation reactions for the long-lived fission products 137Cs, 90Sr and 107Pd have been studied for the purpose of nuclear waste transmutation. The cross sections on the proton- and deuteron-induced spallation were obtained in inverse kinematics at the RIKEN Radioactive Isotope Beam Factory. Both the target and energy dependences of cross sections have been investigated systematically. and the cross-section differences between the proton and deuteron are found to be larger for lighter fragments. The experimental data are compared with the SPACS semi-empirical parameterization and the PHITS calculations including both the intra-nuclear cascade and evaporation processes.

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

    International Nuclear Information System (INIS)

    Slaughter, D.R.; Accatino, M.R.; Alford, O.J.; Bernstein, A.; Descalle, M.; Gosnell, T.B.; Hall, J.M.; Loshak, A.; Manatt, D.R.; McDowell, M.R.; Moore, T.L.; Petersen, D.C.; Pohl, B.A.; Pruet, J.A.; Prussin, S.G.

    2004-01-01

    Full text: A concept for detecting the presence of special nuclear material ( 235 U or 239 Pu) 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

  6. Study on the calculation method of source term from fission products

    International Nuclear Information System (INIS)

    Zhou Jing; Gong Quan; Qiu Haifeng

    2014-01-01

    As a major part of radioactive nuclides, fission products play an important role in nuclear power plant design. The paper analyzes the calculation model of core activity inventory, the model of fission products releasing from the pellets to RCS, the balance model of fission products in RCS, and then proves them by calculation of the typical pressurized water reactor. The model is proved applicable for calculating fission products of pressurized water reactors. (authors)

  7. Proposal to represent neutron absorption by fission products by a single pseudo-fragment

    International Nuclear Information System (INIS)

    Tsibulya, A.M.; Kochetkov, A.L.; Kravchenko, I.V.; Nikolaev, M.N.

    1991-01-01

    The concentration of fission products during reactor operation is analyzed. The dependence of a composite fission product capture cross-section as a function of time and on the nature of the A of the fissile nuclide are investigated, and the neutron radiative capture in fission products of a thermal reactor is evaluated. It is concluded that neutron absorption by fission products can be described by pseudo-fragments. (author). 18 refs, 2 figs, 3 tabs

  8. Investigation of spatial distribution of fission-rate of natural uranium nuclei in the blanket of electronuclear setup 'Energy plus Transmutation' at Dubna Nuclotron proton beam at energy 1.5 GeV

    International Nuclear Information System (INIS)

    Hashemi-Nezhad, S.R.; Zhuk, I.V.; Kievets, M.; Krivopustov, M.I.; Sosnin, A.N.; Westmeier, W.; Brandt, R.

    2008-01-01

    The 'Energy plus Transmutation' experimental setup of the Veksler and Baldin Laboratory of High Energy Physics within the Joint Institute for Nuclear Research (JINR) in Dubna, Russia, is a lead target (with a diameter of 8.4 cm and length of 45.6 cm) surrounded by a uranium blanket (weight 206.4 kg of natural uranium). A polyethylene plus cadmium shield is placed around the target-blanket assembly to modify the spallation and fission neutron spectra in the system. The setup was irradiated by a proton beam of energy 1.5 GeV using the Nuclotron accelerator. The spatial distribution of natural uranium fission-rate in the assembly and fission-rate in the blanket was determined experimentally and compared with Monte Carlo predictions using the MCNPX 2.6C code. Besides neutron-induced fission the calculations include the Nat U(p,f), Nat U(π,f) as well as Nat U(γ,f) reactions. Good agreement between the experimental and calculation results was obtained. The possible sources of errors in the experiment and calculations are discussed in detail

  9. Proton-fission for the accelerator production of Mo-99

    International Nuclear Information System (INIS)

    Lagunas-Solar, M.C.; Jungerman, J.A.; Castaneda, C.M.

    1993-01-01

    The production of Mo-99 (66.0 h) via de U-238(p,f) Mo-99 fission reaction is proposed as a non-reactor source of this essential precursor of 6.6-h Tc-99m, an isotope of wide use of diagnostic nuclear medicine applications. Measurements of the total excitation function for the U-238(p,f) reaction indicated a maximum and fairly constant cross section of 1.4 barns at > 30 MeV. Combining the advances of high-current (mA) H-accelerators with dual beam (dual target) operation, and assuming a 5% fission yield, estimates of Mo-99 reaches 5 to 14 Ci/h at 1 mA. The proton fission production of Mo-99 appears to more advantageous than the reactor produced via evaporation neutron-induced fission. An accelerator method could allow securing ample supply of Mo-99 independently of the current scarce reactor operation, while also simplifying the associated waste management problems as well as some of the environmental concerns

  10. Fission product release from TRIGA-LEU reactor fuels

    International Nuclear Information System (INIS)

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

    1980-01-01

    Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

  11. Gamma ray beam transmutation

    International Nuclear Information System (INIS)

    Imasaki, K.; Li, D.; Miyamoto, S.; Amano, S.; Motizuki, T.

    2007-01-01

    We have proposed a new approach to nuclear transmutation by a gamma ray beam of Compton scattered laser photon. We obtained 20 MeV gamma ray in this way to obtain transmutation rates with the giant resonance of 1 97Au and 1 29Iodine. The rate of the transmutation agreed with the theoretical calculation. Experiments on energy spectrum of positron, electron and neutron from targets were performed for the energy balance and design of the system scheme. The reaction rate was about 1.5∼4% for appropriate photon energies and neutron production rate was up to 4% in the measurements. We had stored laser photon more than 5000 times in a small cavity which implied for a significant improvement of system efficiency. Using these technologies, we have designed an actual transmutation system for 1 29Iodine which has a 16 million year's activity. In my presentation, I will address the properties of this scheme, experiments results and transmutation system for iodine transmutation

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

    International Nuclear Information System (INIS)

    Tong, L.L.; Huang, G.F.; Cao, X.W.

    2015-01-01

    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

  13. Modification of the fission product inventory program FISPIN

    International Nuclear Information System (INIS)

    Thomas, R.B.

    1977-05-01

    The fission product inventory program FISPIN calculates inventories of fission products, actinides and activation products, during and after irradiation in a nuclear reactor, estimates also being given for heat output and radioactive activity of the isotopes. The program has been developed further by making provision for the simulation of fuel reprocessing and in providing subroutines to make the program compatible with nuclear data in a slightly modified ENDF/B4 format. Continuous development of FISPIN over the years has however involved many program alterations and additions, and this has resulted in a generally untidy and cumbersome program. An attempt has therefore been made to improve the basic structure of the program. The subject is dealt with under the following headings: modularisation, direct access data, override facility, selective output, flowcharts, summary. (U.K.)

  14. Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

    Science.gov (United States)

    Le Gall, C.; Geiger, E.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

    2017-11-01

    Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

  15. Study on neutron spectrum for effective transmutation of minor actinides in thermal reactors

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yokoyama, Kenji

    1997-01-01

    The transmutation of minor actinides (MAs) has been investigated in thermal reactor cells using mixed oxide fuel with MAs. The effect of neutron spectra on transmutation is studied by changing the neutron spectra. Five transmutation rates are compared: direct fission incineration rate, capture transmutation rate, consumption rate, overall fission incineration rate and inventory difference transmutation rate. The relations between these transmutation rates and their dependence on the neutron spectrum were investigated. To effectively incinerate MAs it is necessary to maximize the overall fission incineration rate and the inventory difference transmutation rate. These transmutation rates become maximum when the fraction of neutrons below 1 eV is about 8% for the case where the MA addition is 1-3%. When the MA addition is over 5%, the transmutation rates become maximum for very hard neutron spectrum. (Author)

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

    Science.gov (United States)

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

    2014-05-01

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

  17. Methodology and experimental setup for measuring short-lives fission product yields in actinides induced fission by charged particles

    International Nuclear Information System (INIS)

    Bellido, A.V.

    1995-07-01

    The theoretical principles and the laboratory set-up for the fission products yields measurements are described. The procedures for the experimental determinations are explain in detail. (author). 43 refs., 5 figs

  18. Study on the LLFPs transmutation in a super-critical water-cooled fast reactor

    International Nuclear Information System (INIS)

    Lu Haoliang; Ishiwatari, Yuki; Oka, Yoshiaki

    2011-01-01

    Research highlights: → Transmutation of LLFPs with a super-criticial water cooled fast reactor. → Transmutation of iodine and cesium without the isotopic separation. → The transmuted isotope was mixed with UO 2 to reduce the effect of self-shielding. → A weak neutron moderator Al 2 O 3 was used to suppress the creation of 135 Cs from 133 Cs. - Abstract: The performance of the super-critical water-cooled fast reactor (Super FR) for the transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with the soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super FR. First region is in the blanket assembly due to the ZrH 1.7 layer which was utilized to slow down the fast neutrons to achieve a negative void reactivity. Second region is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected in the transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR or fast reactor. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered to avoid the separation. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe year and 2.79%/GWe year can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the yields from 11.8 and 6.2 1000 MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000 MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained in the Super FR. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super FR. It turns out that the

  19. Accelerator transmutation of waste economics

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1995-01-01

    A parametric systems model of the accelerator transmutation of (nuclear) waste (ATW) is used to examine key system trade-offs and design drivers on the basis of unit costs. This model is applied primarily to a fluid-fuel blanket concept for an ATW that generates net electric power from the fissioning of spent commercial reactor fuel. An important goal of this study is the development of essential parametric trade-offs to aid in any future conceptual engineering design of an ATW that would burn spent commercial fuel and generate net electric power. As such, costing procedures and methodologies used to estimate and compare advanced nuclear power generation systems are applied. The cost of electricity required by an electrical power-generating ATW fueled with spent commercial fuels is generally found to be above that projected for other advanced fission power plants. The accelerator and the chemical plant equipment cost accounts are quantitatively identified as main cost drivers, with the capital cost of radio-frequency power dominating the former. Significant reductions of this cost differential are possible by increased blanket neutron multiplication, increased plant capacity, or increased thermal-to-electric conversion efficiency. The benefits of reduced long-lived fission products and spent commercial fuel actinides provided by the ATW approach translate into a less tangible source of revenue to be provided by a charge that must be levied on the client fission power plants being serviced. The main goal of this study, however, is not a direct cost comparison but is instead a quantitative determination of cost-based sensitivity of key cost drivers and operational modes for an ATW concept that would address the growing spent commercial fuel problem; parametric results presented focus on this goal, and a specific ATW ''straw man'' is given to achieve this main objective

  20. Tracking of fission products release during refueling operations

    International Nuclear Information System (INIS)

    Agarwal, Sharad; Prajapat, M.K.; Vyas, Shyam; Hussain, S.A.

    2001-01-01

    It has been always observed that the release of fission products increase during refueling operations. At RAPP-3 and 4 an attempt has been made to follow-up the change in fission products activity release at each stage of refueling operation and quantification of concentrations of various radionuclides. This exercise was also extended to refueling operation of the channels containing suspected failed fuel. A level of FPNG ( 133 Xe) was observed to increase by a factor of about 10-40 during refueling of failed channel as compared to healthy channel. It can be concluded that by monitoring FPNG levels in exhaust status of the healthiness of spent fuel can be found out. This report discusses in detail the experiment conducted for this purpose. (author)

  1. Measurement of fission product release during LWR fuel failure

    International Nuclear Information System (INIS)

    Osetek, D.J.; King, J.J.

    1979-01-01

    The PBF is a specialized test reactor consisting of an annular core and a central test space 21 cm in diameter and 91 cm high. A test loop circulates coolant through the central experimental section at typical power reactor conditions. Light-water-reactor-type fuel rods are exposed to power bursts simulating reactivity insertion transients, and to power-cooling-mismatch conditions during which the rods are allowed to operate in film boiling. Fission product concentrations in the test loop coolant are continuously monitored during these transients by a Ge(Li) detector based gamma spectrometer. Automatic batch processing of pulse height spectra results in a list of radionuclide concentrations present in the loop coolant as a function of time during the test. Fission product behavior is then correlated to test parameters and posttest examination of the fuel rods. Data are presented from Test PCM-1

  2. Release of fission products from contaminated sodium fires

    International Nuclear Information System (INIS)

    Jordan, S.

    1976-01-01

    Leaks in the primary coolant system of a LMFBR and also serious incidents with tank rupture may entail the escape of fission products into the containment of the reactor. For incident analysis it is important to know the retention capability of sodium for the different fission products. The release of cesium and strontium from pools contaminated with 100 to 1000 ppM was investigated by experiments. The cesium content of airborne aerosols depends on oxygen concentration: at 21 percent oxygen concentration the Cs content of sodium-oxide aerosols is 3 times and at 0.5 percent 15 times as high as the initial Cs concentration in the pool. Strontium content of aerosols over burning contaminated sodium pools is 10 3 times smaller than the strontium pool concentration

  3. Recoil release of fission products from nuclear fuel

    International Nuclear Information System (INIS)

    Wise, C.

    1985-01-01

    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. (orig.)

  4. Quantitative analysis of fission products by γ spectrography

    International Nuclear Information System (INIS)

    Malet, G.

    1962-01-01

    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 ( 144 Ce + 144 Pr activity)/ 137 Cs 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: 141 Ce, 144 Ce + 144 Pr, 103 Ru, 106 Ru + 106 Rh, 137 Cs, 95 Zr + 95 Nb. 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) [fr

  5. Fission product release from HTGR coated microparticles and fuel elements

    International Nuclear Information System (INIS)

    Gusev, A.A.; Deryugin, A.I.; Lyutikov, R.A.; Chernikov, A.S.

    1991-01-01

    The article presents the results of the investigation of fission products release from microparticles with UO 2 core and five-layer HII PyC- and SiC base protection layers of TRICO type as well as from spherical fuel elements based thereon. It is shown that relative release of short-lived xenon and crypton from microparticles does not exceed (2-3) 10 -7 . The release of gaseous fission products from fuel elements containing no damaged coated microparticles, is primarily determined by the contamination of matrix graphite with fuel. An analytical dependence is derived, the dependence described the relation between structural parameters of coated microparticles, irradiation conditions and fuel burnup at which depressurization of coated microparticles starts

  6. Study of deep subcritical electronuclear systems and feasibility of their application for energy production and radioactive waste transmutation

    International Nuclear Information System (INIS)

    Adam, J.; Baldin, A.; Vladimirova, N.

    2010-01-01

    Physical substantiation for investigation of new schemes of electronuclear power production and transmutation of long-lived radioactive wastes based on nuclear relativistic technologies is presented. 'E and T RAW' ('Energy and Transmutation of Radioactive Wastes') is aimed at complex study of interaction of relativistic beams of the Nuclotron-M with energies up to 10 GeV in quasi-infinite targets. Feasibility of application of natural/depleted uranium or thorium without the use of uranium-235, as well as utilization of spent fuel elements of atomic power plants is demonstrated based on analysis of results of known experiments, numerical, and theoretical works. The 'E and T RAW' project will provide fundamentally new data and numerical methods necessary for design of demonstration experimental-industrial setups based on the proposed scheme

  7. Metallic fission product releases from HTR-spherical fuel elements

    International Nuclear Information System (INIS)

    Helmbold, M.; Amian, W.; Stoever, D.; Hecker, R.

    1978-01-01

    Fission product releases from fuel determines to a large extent the feasibility of a special reactor concept. Basic data describing the diffusion behaviour from coated particle fuel are presented concerning isotopes Cs 137 , Sr 90 and Agsup(110m). Taking into account these data for typical 3000MWth plants release calculations are performed. Sensitive release parameters could be defined and the results show low release figures for all the considered reactor concepts. (author)

  8. Reactions of newly formed fission products in the gas phase

    International Nuclear Information System (INIS)

    Strickert, R.G.

    1976-01-01

    A dynamic gas-flow system was constructed which stopped fission products in the gas phase and rapidly separated (in less than 2 sec) volatile compounds from non-volatile ones. The filter assembly designed and used was shown to stop essentially all non-volatile fission products. Between 5 percent and 20 percent of tellurium fission-product isotopes reacted with several hydrocarbon gases to form volatile compounds, which passed through the filter. With carbon monoxide gas, volatile tellurium compound(s) (probably TeCO) were also formed with similar efficiencies. The upper limits for the yields of volatile compounds formed between CO and tin and antimony fission products were shown to be less than 0.3 percent, so tellurium nuclides, not their precursors, reacted with CO. It was found that CO reacted preferentially with independently produced tellurium atoms; the reaction efficiency of beta-produced atoms was only 27 +- 3 percent of that of the independently formed atoms. The selectivity, which was independent of the over-all reaction efficiency, was shown to be due to reaction of independently formed atoms in the gas phase. The gas phase reactions are believed to occur mainly at thermal energies because of the independence of the yield upon argon moderator mole-fraction (up to 80 percent). It was shown in some experiments that about one-half of the TeCO decomposed in passing through a filter and that an appreciable fraction (approximately 20 percent) of the tellurium atoms deposited on the filter reacted agin with CO. Other tellurium atoms on the filter surface (those formed by beta decay and those formed independently but not reacting in the gas phase) also reacted with CO, but probably somewhat less efficiently than atoms formed by TeCO decomposition. No evidence was found for formation of TeCO as a direct result of beta-decay

  9. Fission product source terms and engineered safety features

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1984-01-01

    The author states that new, technically defensible, methodologies to establish realistic source term values for nuclear reactor accidents will soon be available. Although these methodologies will undoubtedly find widespread use in the development of accident response procedures, the author states that it is less clear that the industry is preparing to employ the newer results to develop a more rational approach to strategies for the mitigation of fission product releases. Questions concerning the performance of existing engineered safety systems are reviewed

  10. Process for ultimate storage of radioactive fission products

    International Nuclear Information System (INIS)

    Baukal, W.; Gruenthaler, K.H.; Neumann, K.

    1980-01-01

    In order to exclude cracking in the cooling phase during sealing of radioactive oxidic fission products in glass melts, metallic filling elements - e.g. wires, tissues - are proposed to be incorporated in the mould before the glass melt is poured in. Especially nickel alloys with corrosion proof surface layers, e.g. titanium nitride, silicon carbide, silicon nitride, aluminium oxide, suit best. These elements reduce thermal stresses and effect high thermal conductance towards the mould wall. (UWI) [de

  11. Determination of {sup 90}Sr in uranium fission products

    Energy Technology Data Exchange (ETDEWEB)

    Bajo, S; Tobler, L [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-02-01

    A previously published radiochemical procedure for the determination of {sup 90}Sr 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 {sup 90}Sr in environmental materials following a nuclear accident. The procedure is based on tributylphosphate extraction of {sup 90}Y, precipitation of Y-oxalate, and counting in a proportional counter. (author) figs., tabs., 10 refs.

  12. Biological effects induced by low amounts of nuclear fission products

    International Nuclear Information System (INIS)

    Vasilenko, I.Ya.; Shishkin, V.F.; Khudyakova, N.V.

    1991-01-01

    The review deals with the problem of biological hazard of low radiation doses for animals and human beings taking into the danger of internal and external irradiation by nuclear fission products under conditions of enhancing anthropogenic radiation contamination of biosphere. An attention is paid to the estimation of life span carcinogenesis, genetic and delayed effects. A conclusion is made on a necessity of multiaspect investigation of biological importance of low radiation doses taking into account modifying effects of other environmental factors

  13. Forced decontamination of fission products deposited on urban areas

    International Nuclear Information System (INIS)

    Warming, L.

    1984-12-01

    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)

  14. Thermochemical data for reactor materials and fission products

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1990-01-01

    This volume presents a collection of critically assessed data on inorganic compounds which are of special interest in nuclear reactor safety studies. Thermodynamic equilibrium calculations are an important and widely used instrument in the understanding of the chemical behavior and release of fission products in the course of nuclear reactor accidents. The reliability of such calculations is, nevertheless, limited by the availability of accurate input data for relevant compounds

  15. The Metabolic Properties of the Fission Products and Actinide Elements

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton M.D., J.G.

    1948-03-01

    An investigation of the assimilation, distribution, retention, an excretion of the fission products and actinide elements in the rat has been conducted at the Crocker Radiation Laboratory, University of California, Berkeley, California. These studies were initiated October 15, 1942, and are continuing at the present time. An extensive survey has been made of the metabolism of twenty-two different radio elements in the rat.

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

    International Nuclear Information System (INIS)

    Chrien, R.E.; Burrows, T.W.

    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

  17. 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)

  18. Characteristic relation for the mass and energy distribution of the nuclear fission products

    International Nuclear Information System (INIS)

    Alexandru, G.

    1977-01-01

    The dispersion relation for nuclear fission is written in the two part fragmentation approach which allows to obtain the characteristic relation for the mass and energy distribution of the nuclear fission products. One explains the resonance approximation in the mass distribution of the fission products taking into account the high order resonances too. (author)

  19. The universal library of fission products and delayed neutron group yields

    International Nuclear Information System (INIS)

    Koldobskiy, A.B.; Zhivun, V.M.

    1997-01-01

    A new fission product yield library based on the Semiempirical method for the estimation of their mass and charge distribution is described. Contrary to other compilations, this library can be used with all possible excitation energies of fissionable actinides. The library of delayed neutron group yields, based on the fission product yield compilation, is described as well. (author). 15 refs, 4 tabs

  20. Radiation research contracts: Distribution of fission products in the biosphere

    International Nuclear Information System (INIS)

    Schoenfeld, T.

    1959-01-01

    According to its Statute the IAEA has to fulfil a dual function - to help individual countries in solving their specific problems and to undertake tasks in the common interest of all its Member States. With this latter aim in mind the Agency has placed a number of research contracts with national research institutes. One of them deals with the distribution of fission products in the biosphere. The Agency has contributed to this work by putting at the institutes' disposal scientists from its own staff apparatus and financial aid. Protection against ionizing radiation given off in nuclear transformations is one of the foremost safety problems in all atomic energy operations. While every effort is being made to prevent reactors, processing plants and all other installations from releasing radioactive materials into the biosphere - air, water and earth - under any foreseeable conditions, small amounts of it are actually released into man's living space. Undoubtedly, this will continue to be so, at least for the time being. For example, low activity liquid wastes from some chemical processing plants are decontaminated in special processes, but traces of fission products remain in the liquids finally discharged on the ground or to nearby waterways. In some installations low and medium activity liquid wastes are even released on the ground or into swamps without prior decontamination. It is also to be expected that in accidents larger amounts of fission products may occasionally be released. (author)

  1. Radiation research contracts: Distribution of fission products in the biosphere

    Energy Technology Data Exchange (ETDEWEB)

    Schoenfeld, T [Vienna University, First Chemical Institute, Vienna (Austria)

    1959-04-15

    Protection against ionizing radiation given off in nuclear transformations is one of the foremost safety problems in all atomic energy operations. While every effort is being made to prevent reactors, processing plants and all other installations from releasing radioactive materials into the biosphere - air, water and earth - under any foreseeable conditions, small amounts of it are actually released into man's living space. Undoubtedly, this will continue to be so, at least for the time being. For example, low activity liquid wastes from some chemical processing plants are decontaminated in special processes, but traces of fission products remain in the liquids finally discharged on the ground or to nearby waterways. In some installations low and medium activity liquid wastes are even released on the ground or into swamps without prior decontamination. It is also to be expected that in accidents larger amounts of fission products may occasionally be released. To make the routine release of small amounts of fission products safe and to be able to estimate the possible effect of larger releases in accidents, a considerable amount of information is required

  2. Radiation research contracts: Distribution of fission products in the biosphere

    International Nuclear Information System (INIS)

    Schoenfeld, T.

    1959-01-01

    Protection against ionizing radiation given off in nuclear transformations is one of the foremost safety problems in all atomic energy operations. While every effort is being made to prevent reactors, processing plants and all other installations from releasing radioactive materials into the biosphere - air, water and earth - under any foreseeable conditions, small amounts of it are actually released into man's living space. Undoubtedly, this will continue to be so, at least for the time being. For example, low activity liquid wastes from some chemical processing plants are decontaminated in special processes, but traces of fission products remain in the liquids finally discharged on the ground or to nearby waterways. In some installations low and medium activity liquid wastes are even released on the ground or into swamps without prior decontamination. It is also to be expected that in accidents larger amounts of fission products may occasionally be released. To make the routine release of small amounts of fission products safe and to be able to estimate the possible effect of larger releases in accidents, a considerable amount of information is required

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

    International Nuclear Information System (INIS)

    Allen, M.D.; Stockman, H.W.; Reil, K.O.; Grimley, A.J.; Camp, W.J.

    1988-01-01

    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

  4. Behavior of fission products released from severely damaged fuel during the PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hagrman, D.L.; Broughton, J.M.; Rest, J.

    1984-01-01

    The results of fission product release behavior during the first two Power Burst Facility Severe Fuel Damage tests are presented. Measured fission product release is compared with calculated release using temperature dependent release rate correlations and FASTGRASS analysis. The test results indicate that release from fuel of the high volatility fission products (Xe, Kr, I, Cs, and Te) is strongly influenced by parameters other than fuel temperature; namely fuel/fission product morphology, fuel and cladding oxidation state, extent of fuel liquefaction, and quench induced fuel shattering. Fission product transport from the test fuel through the sample system was strongly influenced by chemical effects. Holdup of I and Cs was affected by fission product chemistry, and transport time while Te release was primarily influenced by the extent of zircaloy oxidation. Analysis demonstrates that such integral test data can be used to confirm physical, chemical, and mechanistic models of fission product behavior for severe accident conditions

  5. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

    2011-07-01

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

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

    CERN Document Server

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

    1999-01-01

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

  7. A novel method for fission product noble gas sampling

    International Nuclear Information System (INIS)

    Jain, S.K.; Prakash, Vivek; Singh, G.K.; Vinay, Kr.; Awsthi, A.; Bihari, K.; Joyson, R.; Manu, K.; Gupta, Ashok

    2008-01-01

    Noble gases occur to some extent in the Earth's atmosphere, but the concentrations of all but argon are exceedingly low. Argon is plentiful, constituting almost 1 % of the air. Fission Product Noble Gases (FPNG) are produced by nuclear fission and large parts of FPNG is produced in Nuclear reactions. FPNG are b-j emitters and contributing significantly in public dose. During normal operation of reactor release of FPNG is negligible but its release increases in case of fuel failure. Xenon, a member of FPNG family helps in identification of fuel failure and its extent in PHWRs. Due to above reasons it becomes necessary to assess the FPNG release during operation of NPPs. Presently used methodology of assessment of FPNG, at almost all power stations is Computer based gamma ray spectrometry. This provides fission product Noble gases nuclide identification through peak search of spectra. The air sample for the same is collected by grab sampling method, which has inherent disadvantages. An alternate method was developed at Rajasthan Atomic Power Station (RAPS) - 3 and 4 for assessment of FPNG, which uses adsorption phenomena for collection of air samples. This report presents details of sampling method for FPNG and noble gases in different systems of Nuclear Power Plant. (author)

  8. Photofission observations in reactor environments using selected fission-product yields

    International Nuclear Information System (INIS)

    Gold, R.; Ruddy, F.H.; Roberts, J.H.

    1982-01-01

    A new method for the observation of photofission in reactor environments is advanced. It is based on the in-situ observation of fission product yield. In fact, at a given in-situ reactor location, the fission product yield is simply a weighted linear combination of the photofission product yield, Y/sub gamma/, and the neutron induced fission product yield, Y/sub n. The weight factors arising in this linear combination are the photofission fraction and neutron induced fission fraction, respectively. This method can be readily implemented with established techniques for measuring in-situ reactor fission product yield. For example, one can use the method based on simultaneous irradiation of radiometric (RM) and solid state track recorder (SSTR) fission monitors. The sensitivity and accuracy and current knowledge of fission product yields. Unique advantages of this method for reactor applications are emphasized

  9. Angular momentum distribution of primary fission fragments by measurement of the relative yield of isomeric fission products

    International Nuclear Information System (INIS)

    Dornhoefer, H.

    1980-01-01

    The fission products 132 I and 136 I produced in the fission reactions 238 U(α,f) and 238 U(d,f) were spectroscoped using a gas transport system. Thereby was taken advantage of the fact that at the transport with pure helium without aerosols only iodine activities were collected in a membrane filter. The relative independent yields of the isomeric fission products of 132 I and 136 I were determined for different excitation energies. Thereby was taken advantage of the fact that the transport yield of the gas transport system for 136 I directly produced from the fission was greater than for iodine indirectly produced by β-decay. (orig./HSI) [de

  10. Research activities related to accelerator-based transmutation at PSI

    International Nuclear Information System (INIS)

    Wydler, P.

    1993-01-01

    Transmutation of actinides and fission products using reactors and other types of nuclear systems may play a role in future waste management schemes. Possible advantages of separation and transmutation are: volume reductions, the re-use of materials, the avoidance of a cumulative risk, and limiting the duration of the risk. With its experience in reactor physics, accelerator-based physics, and the development of the SINQ spallation neutron source, PSI is in a good position to perform basic theoretical and experimental studies relating to the accelerator-based transmutation of actinides. Theoretical studies at PSI have been concentrated, so far, on systems in which protons are used directly to transmute actinides. With such systems and appropriate recycling schemes, the studies showed that considerable reduction factors for long-term toxicity can be obtained. With the aim of solving some specific data and method problems related to these types of systems, a programme of differential and integral measurements at the PSI ring accelerator has been initiated. In a first phase of this programme, thin samples of actinides will be irradiated with 590 MeV protons, using an existing irradiation facility. The generated spallation and fission products will be analysed using different experimental techniques, and the results will be compared with theoretical predictions based on high-energy nucleon-meson transport calculations. The principal motivation for these experiments is to resolve discrepancies observed between calculations based on different high-energy fission models. In a second phase of the programme, it is proposed to study the neutronic behaviour of multiplying target-blanket assemblies with the help of zero-power experiments set up at a separate, dedicated beam line of the accelerator. (author) 3 figs., 2 tabs., 8 refs

  11. Separation of actinides and long-lived fission products from high-level radioactive wastes (a review)

    International Nuclear Information System (INIS)

    Kolarik, Z.

    1991-11-01

    The management of high-level radioactive wastes is facilitated, if long-lived and radiotoxic actinides and fission products are separated before the final disposal. Especially important is the separation of americium, curium, plutonium, neptunium, strontium, cesium and technetium. The separated nuclides can be deposited separately from the bulk of the high-level waste, but their transmutation to short-lived nuclides is a muchmore favourable option. This report reviews the chemistry of the separation of actinides and fission products from radioactive wastes. The composition, nature and conditioning of the wastes are described. The main attention is paid to the solvent extraction chemistry of the elements and to the application of solvent extraction in unit operations of potential partitioning processes. Also reviewed is the behaviour of the elements in the ion exchange chromatography, precipitation, electrolysis from aqueous solutions and melts, and the distribution between molten salts and metals. Flowsheets of selected partitioning processes are shown and general aspects of the waste partitioning are shortly discussed. (orig.) [de

  12. Current Status of the Transmutation Reactor Technology and Preliminary Evaluation of Transmutation Performance of the KALIMER Core

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Ser Gi; Sim, Yoon Sub; Kim, Yeong Il; Kim, Young Gyum; Lee, Byung Woon; Song, Hoon; Lee, Ki Bog; Jang, Jin Wook; Lee, Dong Uk

    2005-08-15

    Recently the most countries using the nuclear power plants for electricity generation have been faced with the problem of the preparation of the repository for the disposition of the nuclear waste generated from LWR. It was well-known that the issues related with long term risk of the radioactive wastes for the future generations are due only to 1% of the total waste. This small fraction of 1% consists of transuranic (TRU) nuclides such as Pu, Np, Am, Cm and the long lived fission products such as Tc and I. For the transuranic (TRU) nuclides, their half lives range from several years to several hundred thousands years and hence their radioactive toxicity can be lasted over very long time period. This has made the change of the rule of the fast spectrum reactor from the economical use of uranium resource through breeding to the reduction of the nuclear waste through the transmutation. The purpose of this study is to obtain the basic knowledge on the nuclear transmutation technology and to suggest the technical solution ways for the future technology development and enhancement through a survey of the state-of-art of the international research on the nuclear transmutation. The increase of the transmutation rate requires the reduction of the breeding ratio. In fact, the transmutation rate is determined by the breeding ratio. The reduction of the breeding ratio can be achieved by reducing the U-238 content in fuel or increasing the neutron leakage through core boundary or absorbing the neutrons by using some absorbers. However, the reduction of the U-238 content results in the degradation of the fuel Doppler coefficient that is one of the most important safety-related parameters and the reduction of the effective delayed neutron fraction that is related with the controllability of the reactor core. Also, the increase of the transmutation rate can lead to the increase of the coolant void reactivity worth unless some ways to reduce the coolant void reactivity are not

  13. Evaluation of actinide partitioning and transmutation

    International Nuclear Information System (INIS)

    1982-01-01

    After a few centuries of radioactive decay the long-lived actinides, the elements of atomic numbers 89-103, may constitute the main potential radiological health hazard in nuclear wastes. This is because all but a very few fission products (principally technetium-99 and iodine-129) have by then undergone radioactive decay to insignificant levels, leaving the actinides as the principal radionuclides remaining. It was therefore at first sight an attractive concept to recycle the actinides to nuclear reactors, so as to eliminate them by nuclear fission. Thus, investigations of the feasibility and potential benefits and hazards of the concept of 'actinide partitioning and transmutation' were started in numerous countries in the mid-1970s. This final report summarizes the results and conclusions of technical studies performed in connection with a four-year IAEA Co-ordinated Research Programme, started in 1976, on the ''Environmental Evaluation and Hazard Assessment of the Separation of Actinides from Nuclear Wastes followed by either Transmutation or Separate Disposal''. Although many related studies are still continuing, e.g. on waste disposal, long-term safety assessments, and waste actinide management (particularly for low and intermediate-level wastes), some firm conclusions on the overall concept were drawn by the programme participants, which are reflected in this report

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Feuerstein, H.; Thorley, A.W.

    1987-08-01

    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)

  16. Adsorption of fission products on mediterranean mud; Adsorption des produits de fission sur des vases de mediterranee

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, P; Gailledreau, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    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) [French] On a mesure les coefficients de partage dans l'eau de mer de quelques produits de fission a longue periode sur des echantillons de vase preleves en Mediterranee. Les valeurs trouvees sont elevees. Le comportement de ces isotopes radioactifs est discutee. (auteur)

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

    International Nuclear Information System (INIS)

    Marques, R.O.; Cristini, P.R.; Marziale, D.P.; Furnari, E.S.; Fernandez, H.O.

    1987-01-01

    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)

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

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2005-01-01

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

  19. Analysis on Radioactive Waste Transmutation in Light Water cooled Hyb-WT

    International Nuclear Information System (INIS)

    Hong, Seonghee; Kim, Myung Hyun

    2014-01-01

    A feasibility of realization is much higher in FFHR compared with pure fusion. A combination of plasma fusion source for neutrons with a subcritical reactor at the blanket side has much higher capability in transmutation of waste as well as reactor safety compared with fission reactor options. Fusion-Fission Hybrid Reactor (FFHR) uses various coolants depending on the purpose. It is important that coolant being used should be suitable to reactor purpose, because reactor performance and the design constraints may change depending on the coolant. There are basically two major groups of coolants for FFHR. One group of coolant does not contain Li. They are Na, Pb-Bi, H 2 O and D 2 O. The other group contains Li for tritium breeding. They are Li, LiPb, LiSN, FLIBE and FLiNaBe. Currently, the issue in FFHR is its implication for radioactive waste transmutation (FFHR for WT). Because radioactive wastes of spent nuclear fuel (SNF) are transmuted using fusion neutron source. Therefore a suitable coolant should be used for effective waste transmutation. . In FFHR for WT, LiPb coolant is being used mainly because of tritium production in Li and high neutron economic through reaction in Pb. However different coolants use such as Na, Pb-Bi are used in fast reactors and accelerator driven systems (ADS) having same purpose. In this study, radioactive waste transmutation performance of various coolants mentioned above will be compared and analyzed. Through this study, the coolants are judged primarily for their support to waste transmutation disregarding their limitation to reactor design and tritium breeding capability. First, performance of the light water coolant regarding radioactive waste transmutation was analyzed among various coolants mentioned above. In this paper, performance of radioactive waste transmutation can be known depending on different volume fractions (54.53, 60.27, 97.94vol.%) of the light water. Light water dose required fusion power lower than LiPb due to

  20. Direct irradiation of long-lived fission products in an ATW system

    Energy Technology Data Exchange (ETDEWEB)

    Carter, T.F. [Univ. of Tennessee, Knoxville, TN (United States); Henderson, D. [Univ. of Wisconsin, Madison, WI (United States); Sailor, W.C. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    The feasibility of directly irradiating five long-lived fission products (LLFPs: {sup 79}Se, {sup 93}Zr, {sup 107}Pd, {sup 126}Sn, and {sup 135}Cs, each with a half-life greater than 10,000 years), by incorporating them into the target of an Accelerator Transmutation of Waste (ATW) system is discussed. The important parameters used to judge the feasibility of a direct irradiation system were the target`s neutron spallation yield (given in neutrons produced per incident proton), and the removal rate of the LLFP, with the baseline incineration rate set at two light water reactors (LWRs) worth of the LLFP waste per year. A target was constructed which consisted of a LLFP cylindrical {open_quotes}plug{close_quotes} inserted into the top (where the proton beam strikes) of a 30 cm radius, 100 cm length lead target. {sup 126}Sn and {sup 79}Se were each found to have high enough removal rates to support two LWR`s production of the LLFP per year of ATW operation. For the baseline plug geometry (5 cm radius, 30 cm length) containing {sup 126}Sn, 3.5 LWRs could be supported per year (at 75% beam availability). Furthermore, the addition of a {sup 126}Sn plus had a slightly positive effect on the target`s neutron yield. The neutron production was 36.83 {plus_minus}.0039 neutrons per proton with a pure lead target having a yield of 36.29 {plus_minus}.0038. It was also found that a plug composed of a tin-selenide compound (SnSe) had high enough removal rates to burn two or more reactor years of both LLFPs simultaneously.

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

    International Nuclear Information System (INIS)

    Barrachina, M.; Villar, M. A.

    1965-01-01

    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

  2. Oxide fuels and targets for transmutation

    International Nuclear Information System (INIS)

    Sudreau, F.; Bonnerot, J.M.; Warin, D.; Gaillard-Groleas, G.; Ferroud-Plattet, M.P.

    2007-01-01

    Full text of publication follows. Direction 1 of the French Act dated 30 December 1991 on the management of high-level, long-lived radioactive waste involves exploring solutions designed to separate long-lived radionuclides from the spent fuel and to transmute them under neutron flux into shorter half-lives or stable elements. In the French research programme conducted by CEA, these radionuclides are mainly minor actinides (americium, neptunium and curium) and fission products (particularly caesium, iodine and technetium). Within this context, this paper aims at illustrating the vast programme that CEA has performed in order to demonstrate the scientific and technical feasibility of minor actinide transmutation. An important part of the research was carried out in collaboration with French research (CNRS) and industrial (EDF, AREVA) organisations, and also in the framework of international co-operation programmes with the European Institute for Transuranium Elements in Karlsruhe (ITU), the US Department of Energy (DOE), the Japanese Atomic Energy Research Institute (now JAEA) and Central Research Institute of Electric Power Industry (CRIEPI) and the Russian Ministry for Atomic Energy (ROSATOM). Such research made it possible to evaluate the capacity of MOX fuels to be used as a support for minor actinide transmutation (homogeneous method). Simulations of pressurised water reactor (PWR) fuels have revealed the limits of this transmutation method, which are mainly related to the pressurization of the fuel rods and the formation of high active californium. On the contrary, for sodium-cooled fast reactor fuels possibly designed with large expansion plenums a first experimental demonstration of the transmutation of americium and neptunium has been successful in the Phenix reactor. Various studies designed to demonstrate the theoretical and experimental feasibility of transmutation using an inert support (heterogeneous method) have been carried out in HFR (EFTTRA

  3. 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.

  4. The status of nuclear data for transmutation calculations

    International Nuclear Information System (INIS)

    Wilson, W.B.; England, T.R.; MacFarlane, R.E.; Muir, D.W.; Young, P.G.

    1995-01-01

    At this point, the accurate description of transmutation products in a radiation environment is more a nuclear data problem than a code development effort. We have used versions of the CINDER code for over three decades to describe the transmutation of nuclear reactor fuels in radiation environments. The need for the accurate description of reactor neutron-absorption, decay-power, and decay-spectra properties have driven many AEC, ERDA, and DOE supported nuclear data development efforts in this period. The level of cross-section, decay, and fission-yield data has evolved from rudimentary to a comprehensive ENDF/B-VI library permitting great precision in reactor calculations. The precision of the data supporting reactor simulations provides a sturdy foundation for the data base required for the wide range of transmutation problems currently studied. However, such reactor problems are typically limited to neutron energies below 10 MeV or so; reaction and decay data are required for actinides of, say, 90 ≤ Z ≤ 96 neutron-rich fission products of 22 ≤ Z ≤ 72. The expansion into reactor structural materials and fusion systems extends these ranges in energy and Z somewhat. The library of nuclear data, constantly growing in breadth and quality with international cooperation, is now described in the following table

  5. Post-irradiation studies on knock-out and pseudo-recoil releases of fission products from fissioning UO2

    International Nuclear Information System (INIS)

    Yamagishi, S.; Tanifuji, T.

    1976-01-01

    By using post-irradiation techniques, in-pile releases of 133 Xe, sup(85m)Kr, 88 Kr, 87 Kr and 138 Xe from UO 2 fissioning at low temperatures below about 200 0 C are studied: these are analyzed into a time-dependent knock-out and time-independent pseudo-recoil releases. For the latter, a 'self knock-out' mechanism is proposed: when a fission fragment loses thoroughly its energy near the UO 2 surface and stops there, it will knock out the surface substances and accordingly the fragment (i.e. the fission product) will be released. The effective thickness of the layer where the self knock-out occurs is found to be approximately 7A. As for the knock-out release, the following is estimated from its dependence on various factors: the knock-out release of fission products occurs from the surface layer with the effective thickness of approximately 20A: the shape of UO 2 matrix knocked out by one fission fragment passing through the surface is equivalent to a cylinder approximately 32A diameter by approximately 27A thick, (i.e. the knock-out coefficient for UO 2 is approximately 660 uranium atoms per knock-out event). On the basis of the above estimations, the conclusions derived from the past in-pile studies of fission gas releases are evaluated. (Auth.)

  6. Preparation of a primary target for the production of fission products in a nuclear reactor

    International Nuclear Information System (INIS)

    Arino, H.; Cosolito, F.J.; George, K.D.; Thornton, A.K.

    1976-01-01

    A primary target for the production of fission products in a nuclear reactor, such as uranium or plutonium fission products, is comprised of an enclosed, cylindrical vessel, preferably comprised of stainless steel, having a thin, continuous, uniform layer of fissionable material, integrally bonded to its inner walls and a port permitting access to the interior of the vessel. A process is also provided for depositing uranium material on to the inner walls of the vessel. Upon irradiation of the target with neutrons from a nuclear reactor, radioactive fission products, such as molybdenum-99, are formed, and thereafter separated from the target by the introduction of an acidic solution through the port to dissolve the irradiated inner layer. The irradiation and dissolution are thus effected in the same vessel without the necessity of transferring the fissionable material and fission products to a separate chemical reactor. Subsequently, the desired isotopes are extracted and purified. Molybdenum-99 decays to technetium-99m which is a valuable medical diagnostic radioisotope. 3 claims, 1 drawing figure

  7. Nuclear Waste Separation and Transmutation Research with Special Focus on Russian Transmutation Projects Sponsored by ISTC

    International Nuclear Information System (INIS)

    Conde, Henri; Blomgren, Jan; Olsson, Nils

    2003-03-01

    High-level nuclear reactor waste is made up of relatively few long-lived radioactive species, among them plutonium, that contribute to difficulties with its storage and disposal. Separation of these species from larger waste volumes mainly constituting of uranium (about 95 %) coupled with nuclear incineration to fission products of plutonium and the so called minor actinides (Neptunium, Americium, and Curium) and transmutation of some of the long lived fission products to short lived or stable isotopes represents a viable nuclear waste management strategy to drastically reduce the time and space requirements for a bed-rock repository of the remaining waste. A remarkable increase in the international research and development on partitioning and transmutation has occurred during the recent years. The road-map report published in April 2001 by The European Technical Working Group on ADS for the development of a European demonstration facility for nuclear waste transmutation has high-lighted the ongoing European research and pointed out the need for further research. The road-map has given the different research activities a position in the ultimate goal of producing an ADS demonstrator and is guiding research planning on the national as well as on the EU level. The Advanced Accelerator Application (3A) program in the US, with the long term goals to enhance long term public safety, provide benefits for the repository, reduce proliferation risks and improve prospects for nuclear power has focused the research on nuclear waste transmutation. The reports on the 3A program indicates a change of the US former abandonment position towards reprocessing and fast reactors due to a strong incentive to eliminate the reactor plutonium and to lower the amount of high level reactor waste for the Yucca Mountain repository. The SKB's proposed research and development program for the next 3 years (FUD01) was presented by SKB in September 2001. It is proposed that the research program

  8. A study of γ-ray source for the transmutation

    International Nuclear Information System (INIS)

    Nomura, Masahiro; Takahashi, Hiroshi.

    1996-07-01

    PNC is developing high power CW electron linac for various applications, those are the transmutation of the fission products, Free Electron Laser (FEL), the positron source and so on. Especially, the transmutation by the electron linac has been studied for several years. As the results, high flux and high energy γ-ray (∼15 MeV) is required, one of the big problems is that plenty of transmutation energy is needed and the narrow γ-ray energy spectrum can reduce the transmutation energy. The γ-rays can be produced by synchrotron radiation, FEL and laser compton scattering. Those methods were described briefly and compared. As a result, the laser compton scattering is one of the good methods to produce high energy γ-ray. However the cross section between electron and photon is small and the scattered photon energy spectrum is not so narrow that the transmutation energy is reduced drastically. To enhance the interaction between electron and photon, the super cavity is proposed. And some experiments are in progress. To reduce the transmutation energy, scattered electron must be reused by the storage ring. If the scattered electrons are not used for producing γ-ray, the efficiency is less than 1%. In our system, the efficiency can be increased to 20% by reusing scattered electrons. But this efficiency is still low. To increase the efficiency, the RF bucket must be enlarged. If the momentans compaction factor α can be reduced, the RF bucket can be enlarged. And the storage ring must be designed to have small value of the α. The electron energy dependency of efficiency is investigated, too. In short word, it is difficult to increase the efficiency drastically by changing electron energy. This work was conducted as a part of the collaboration work between PNC and BNL. (author)

  9. Fuel and target programs for the transmutation at Phenix and other reactors; Programmes combustibles et cibles pour la transmutation dans Phenix et autres reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Gaillard-Groleas, G

    2002-07-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

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

    International Nuclear Information System (INIS)

    Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

    1984-01-01

    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 1400 0 C to >50% at 2000 0 C; 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

  11. Transient fission product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.

    1995-01-01

    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)

  12. Behavior of Nb fission product during nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Gue, J.P.

    1977-02-01

    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 [fr

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

    International Nuclear Information System (INIS)

    Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

    1984-01-01

    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 1400 0 C to >50% at 2000 0 C; 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)

  14. Fission-product burnup chain model for research reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

    1990-12-01

    A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

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

    International Nuclear Information System (INIS)

    Morlevat, J.-P.; Uny, Gisele; Jacquet-Francillon, Noel.

    1981-06-01

    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 [fr

  16. Equilibrium Temperature Profiles within Fission Product Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, Michael D. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-10-01

    We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.

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

    International Nuclear Information System (INIS)

    Yamaura, Mitiko

    1999-01-01

    The presence of small amount of mixed actinides and long-lived heat generators fission products as 137 Cs and 90 Sr 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 HNO 3 and hydroxylamine nitrate + HNO 3 as eluent. In the second one, CMP0-TBP(XAD7) column, the actinides were retained on the column and the separations were done by using (NH 4 ) 2 C 2 O 4 , DTPA, HNO 3 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

  18. Estimated effects of interfacial vaporization on fission product scrubbing

    International Nuclear Information System (INIS)

    Moody, F.J.; Nagy, S.G.

    1983-01-01

    When bubbles containing non-condensible gas rise through a water pool, interfacial evaporation causes a flow of vapor into the bubbles. The inflow reduces the outward particle motion toward the bubble wall, diminishing the effectiveness of fission product particle removal. This analysis provides an estimate of evaporation on pool scrubbing effectiveness. It is shown that hot gas, which boils water at the bubble wall, reduces the effective scrubbing height by less than five centimeters. Although the evaporative humidification in a rising bubble containing non-condensible gas has a diminishing effect on scrubbing mechanisms, substantial decontamination is still expected even for the limiting case of a saturated pool

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

    International Nuclear Information System (INIS)

    Kim, J.I.

    1989-06-01

    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.)

  20. Method of processing liquid waste containing fission product

    International Nuclear Information System (INIS)

    Funabashi, Kiyomi; Kawamura, Fumio; Matsuda, Masami; Komori, Itaru; Miura, Eiichi.

    1988-01-01

    Purpose: To prepare solidification products of low surface dose by removing cesium which is main radioactive nuclides from re-processing plants. Method: Liquid wastes containing a great amount of fission products are generated accompanying the reprocessing for spent nuclear fuels. After pH adjustment, the liquid wastes are sent to a concentrator to concentrate the dissolved ingredients. The concentrated liquid wastes are pumped to an adsorption tower in which radioactive cesium contributing much to the surface dose is removed. Then, the liquid wastes are sent by way of a surge tank to a mixing tank, in which they are mixed under stirring with solidifying agents such as cements. Then, the mixture is filled in a drum-can and solidified. According to this invention, since radioactive cesium is removed before solidification, it is possible to prepare solidification products at low surface dose and facilitate the handling of the solidification products. (Horiuchi, T.)

  1. A new technique to measure fission-product diffusion coefficients in UO2 fuel

    International Nuclear Information System (INIS)

    Hocking, W.H.; Verrall, R.A.; Bushby, S.J.

    1999-01-01

    This paper describes a new out-reactor technique for the measurement of fission-product diffusion rates in UO 2 . The technique accurately simulates in-reactor fission-fragment effects: a thermal diffusion that is due to localized mixing in the fission track, radiation-enhanced diffusion that is due to point-defect creation by fission fragments, and bubble resolution. The technique utilizes heavy-ion accelerators - low energy (40 keV to 1 MeV) for fission-product implantation, high energy (72 MeV) to create fission-fragment damage effects, and secondary ion mass spectrometry (SIMS) for measuring the depth profile of the implanted species. Preliminary results are presented from annealing tests (not in the 72 MeV ion flux) at 1465 deg. C and 1650 deg. C at low and high concentrations of fission products. (author)

  2. Development of the external cooling device of increase the productivity of neutron-transmutation-doped silicon semiconductor (NTD-Si) (Joint research)

    International Nuclear Information System (INIS)

    Hirose, Akira; Wada, Shigeru; Sasajima, Fumio; Kusunoki, Tsuyoshi; Kameyama, Iwao; Aizawa, Ryouji; Kikuchi, Naoyuki

    2007-01-01

    Neutron-Transmutation-Doped Silicon Semiconductor (hereinafter referred as 'NTD-Si') is the best semiconductor for the power device. The needs of NTD-Si increase recently in proportion to the popularization of hybrid-cars. A fission research reactor, which is a steady state neutron source, is being expected as the best device to meet the needs. So far, we have reconsidered the existing approach which is employed for NTD-Si production works at the research reactors JRR-3, JRR-4 and JMTR of JAEA so as to meet the needs. As one of the effective measures, we found out that the productivity can be increased by incorporating a new device to cool down radioactivity of irradiated silicon ingots at the place outside the main stream from the loading of silicon ingots to the withdrawal of irradiated ingots to the existing JRR-3 Uniformity Irradiation System. Consequently, we developed and installed the device (hereinafter referred as 'external cooling device'). After an ingot was irradiated once, it is turned over manually and irradiated again in order to irradiate the ingot uniformly. With the conventional system, it was necessary to wait the radioactivity of ingot decrease less than the permissible level with holding the ingot in the irradiation equipment. It was effective to shorten the waiting period by using an external cooling device for production increase of NTD-Si. It is expected that the productivity of NTD-Si will be increased by using the external cooling device. This report mentions the design of the external cooling device and verification between its design specifications and the performance of the device completed. (author)

  3. Microprobe study of fission product behavior in high-burnup HTR fuels

    International Nuclear Information System (INIS)

    Kleykamp, H.

    Electron microprobe analysis of irradiated coated particles with high burnup (greater than 50 percent fima) gives detailed information on the chemical state and the transport behavior of the fission products in UO 2 and UC 2 kernels and in the coatings. In oxide fuel kernels, metallic inclusions and ceramic precipitations are observed. The solubility behavior of the fission products in the fuel matrix has been investigated. Fission product inclusions could not be detected in carbide fuel kernels; post irradiation annealed UC 2 kernels, however, give information on the element combinations of some fission product phases. Corresponding to the chemical state in the kernel, Cs, Sr, Ba, Pd, Te and the rare earths are released easily and diffuse through the entire pyrocarbon coating. These fission products can be retained by a silicon carbide layer. The initial stage of a corrosive attack of the SiC coating by the fission products is evidenced

  4. An optimization methodology for heterogeneous minor actinides transmutation

    Science.gov (United States)

    Kooyman, Timothée; Buiron, Laurent; Rimpault, Gérald

    2018-04-01

    In the case of a closed fuel cycle, minor actinides transmutation can lead to a strong reduction in spent fuel radiotoxicity and decay heat. In the heterogeneous approach, minor actinides are loaded in dedicated targets located at the core periphery so that long-lived minor actinides undergo fission and are turned in shorter-lived fission products. However, such targets require a specific design process due to high helium production in the fuel, high flux gradient at the core periphery and low power production. Additionally, the targets are generally manufactured with a high content in minor actinides in order to compensate for the low flux level at the core periphery. This leads to negative impacts on the fuel cycle in terms of neutron source and decay heat of the irradiated targets, which penalize their handling and reprocessing. In this paper, a simplified methodology for the design of targets is coupled with a method for the optimization of transmutation which takes into account both transmutation performances and fuel cycle impacts. The uncertainties and performances of this methodology are evaluated and shown to be sufficient to carry out scoping studies. An illustration is then made by considering the use of moderating material in the targets, which has a positive impact on the minor actinides consumption but a negative impact both on fuel cycle constraints (higher decay heat and neutron) and on assembly design (higher helium production and lower fuel volume fraction). It is shown that the use of moderating material is an optimal solution of the transmutation problem with regards to consumption and fuel cycle impacts, even when taking geometrical design considerations into account.

  5. The EU research activities on partitioning and transmutation. From the 4. to the 5. framework programme

    International Nuclear Information System (INIS)

    Hugon, M.

    1999-01-01

    The European Commission is partly supporting research work on partitioning and transmutation of radioactive waste under the Fourth Framework Programme (1994-1998). This work includes nine research projects. Five strategy studies are evaluating the capabilities of various burners and fuel cycles to limit the production and even destroy the stock of actinides (plutonium and minor actinides). Two experimental projects are aiming at developing techniques for the chemical separation of actinides and two others are dealing with the investigation of transmutation of americium and long-lived fission products. The objectives of these studies are described together with the main results already obtained. The European Union should adopt the 5. Framework Programme (1998-2002) at the end of 1998. The broad lines of the research activities foreseen in partitioning and transmutation and future system under the 5. Framework Programme are briefly presented. (author)

  6. Transmutation of technetium: results of the EFTTRA-T1 experiment

    International Nuclear Information System (INIS)

    Konings, R.J.M.; Stalios, A.D.; Walker, C.T.; Cocuaud, N.

    1998-01-01

    The transmutation of the fission product 99 Tc was studied by neutron irradiation in the High Flux Reactor at Petten. Six metallic rods of pure Tc, distributed over three capsules, were irradiated during eight reactor cycles to a total neutron fluence of about 2 x 10 26 m -2 . Metallographic examinations showed that no changes in the microstructure were visible after the irradiation. Electron probe microanalysis (EPMA) of the radial distribution of Ru, the product of the transmutation process, showed an increase from about 6% in the centre to 12-18% near the rim of the pellets. This effect is due to resonance-shielding of epithermal neutrons in the rim region. Mass spectrometric analysis of the pellet-average Ru concentration showed that the extent of transmutation is about 6-7%. (orig.)

  7. An optimization on strontium separation model for fission products (inactive trace elements) using artificial neural networks

    International Nuclear Information System (INIS)

    Moosavi, K.; Setayeshi, S.; Maragheh, M.Gh.; Ahmadi, S.J.; Kardan, M.R.; Banaem, L.M.

    2009-01-01

    In this study, an experimental design using artificial neural networks for an optimization on the strontium separation model for fission products (inactive trace elements) is investigated. The goal is to optimize the separation parameters to achieve maximum amount of strontium that is separated from the fission products. The result of the optimization method causes a proper purity of Strontium-89 that was separated from the fission products. It is also shown that ANN may be establish a method to optimize the separation model.

  8. Fission product yield evaluation for the USA evaluated nuclear data files

    International Nuclear Information System (INIS)

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

    1994-01-01

    An evaluated set of fission product yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

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

    International Nuclear Information System (INIS)

    Jiacoletti, R.J.; Bailey, P.G.

    1979-08-01

    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

  10. A proposed standard on medical isotope production in fission reactors

    International Nuclear Information System (INIS)

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-01-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  11. Proceedings of the specialists' meeting on accelerator-based transmutation

    International Nuclear Information System (INIS)

    Wenger, H.U.

    1992-09-01

    The meeting was organised under the auspices of OECD Nuclear Agency's International Information Exchange Programme on Actinide and Fission Product Partitioning and Transmutation. In the original announcement for the meeting the following sessions were proposed: 1) Concepts of accelerator-based transmutation systems, 2) Nuclear design problems of accelerator-based transmutation systems with emphasis on target facilities and their interfaces with accelerators, 3) Data and methods for nuclear design of accelerator-based transmutation systems, 4) Related cross-section measurements and integral experiments, 5) Identification of discrepancies and gaps and discussion of desirable R+D and benchmark activities. Due to the large number of papers submitted it was necessary to split session 2 into two parts and to reassign some papers in order to balance the sessions more evenly. No papers were submitted for session 5 and this was replaced by a summary and general discussion session. These proceedings contain all 30 papers in the order they were presented at the meeting. They are copies of the duplication-ready versions given to us during or shortly after the meeting. In the Table of Contents, the papers are listed together with the name of the presenter. (author) figs., tabs., refs

  12. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

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

    International Nuclear Information System (INIS)

    Goncalves, Maria Augusta

    1982-01-01

    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 HNO 3 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)

  14. Advanced system for separation of rare-earth fission products

    International Nuclear Information System (INIS)

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

    1982-01-01

    A microprocessor-controlled radiochemical separation system has been further advanced to separate individual rare-earth elements from mixed fission products in times of a few minutes. The system was composed of an automated chemistry system fed by two approximately 300 μg 252 Cf sources coupled directly by a He-jet to transport the fission products. Chemical separations were performed using two high performance liquid chromatography columns coupled in series. The first column separated the rare-earth group by extraction chromatography using dihexyldiethylcarbamoylmethylphosphonate (DHDECMP) adsorbed on Vydac C 8 resin. The second column isolated the individual rare-earth elements by cation exchange chromatography using Aminex A-9 resin with α-hydroxyisobutyric acid (α-HIBA) as the eluent. Significant results, which have been obtained to date with this advanced system, are the identification of several new neutron-rich rare-earth isotopes including 155 Pm (T=48+-4 s) and 163 Gd (T=68+-3 s). In addition, a half-life of 41+-4 s is reported for 160 Eu. (author)

  15. 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.

  16. Fission Product Release from Spent Nuclear Fuel During Melting

    International Nuclear Information System (INIS)

    Howell, J.P.; Zino, J.F.

    1998-09-01

    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

  17. COMEDIE BD1 experiment: Fission product behaviour during depressurization transients

    International Nuclear Information System (INIS)

    Gillet, R.; Brenet, D.; Hanson, D.L.; Kimball, O.F.

    1996-01-01

    An experimental program in the CEA COMEDIE loop has been carried out to obtain integral test data to validate the methods and transport models used to predict fission product release from the core and plate-out in the primary coolant circuit of the Modular High Temperature Gas Cooled Reactor (MHTGR) during normal operation and liftoff, and during rapid depressurization transients. The loop consists of an in-pile section with the fuel element, deposition section (heat exchanger), filters for collecting condensible Fission Productions (FP) during depressurization tests and an out-of-pile section devoted to chemical composition control of the gas and on-line analysis of gaseous FP. After steady state irradiation, the loop was subjected to a series of in-situ blowdowns at shear ratios (ratio of the wall shear stress during blowdown to that during steady state operation) ranging from 0.7 to 5.6. The results regarding the FP profiles on the plate-out section, before and after blowdowns are given. It appears that: the plate-out profiles depend on the FP chemistry; the depressurization phases have led to significant desorption of I 131, but on the contrary, they have almost no effect for the other FP such as Ag 110m, Cs 134, Cs 137 and Te 132. (author). 1 ref., 15 figs

  18. Tables of RCN-2 fission-product cross section evaluation

    International Nuclear Information System (INIS)

    Gruppelaar, H.

    1979-05-01

    This report (continuation of ECN-13 and ECN-33) describes the third part of the RCN-2 evaluation of neutron cross sections for fission product nuclides in KEDAK format. It contains evaluated data for nine nuclides, i.e. 142 Nd, 143 Nd, 144 Nd, 145 Nd, 146 Nd, 147 Nd, 148 Nd, 150 Nd and 147 Pm. Most emphasis has been given to the evaluation of the radiative capture cross section, in order to provide a data base for adjustment calculations using results of integral measurements. Short evaluation reports are given for this cross section. The evaluated capture cross sections are compared with recent experimental differential and integral data. Graphs are given of the capture cross sections at neutron energies above 1 keV, in which also adjusted point cross sections, based upon integral STEK and CFRMF data have been plotted. Moreover, the results are compared with those of the well-known ENDF/B-IV evaluation for fission product nucleides. Finally, evaluation summaries are given, which include tables of other important neutron cross sections, such as the total, elastic scattering and inelastic scattering cross sections

  19. Research and Development on Coatings for Retaining Fission Product Iodine

    Energy Technology Data Exchange (ETDEWEB)

    Genco, J. M.; Berry, D. A.; Rosenberg, H. S.; Cremeans, G. E.; Morrison, D. L. [Battelle Memorial Institute, Columbus Laboratories, Columbus, OH (United States)

    1968-12-15

    It is well known that elemental iodine has the propensity for forming charge-transfer complexes with amines. These complexes are stable at ambient temperatures and retain much of this stability at elevated temperatures. Amines also react with methyl iodide and hydrogen iodide to form the quaternary ammonium salts and amine salts, respectively. These chemical properties of amines provide the basis for the development of retentive coatings for fission product iodine. Various amine-containing polymers were studied in steam-air environments at elevated temperatures using dilute quantities of tagged iodine. Both non-condensing and condensing steam conditions were investigated. Several of the polymers showed sorption rates and capacities that would be adequate for the chemical removal of accident-released fission-product iodine and were several times more effective than commercial protective coatings currently being used. The removal capabilities for amine polymers also could be enhanced by impregnating the reactant on a matrix material such as asbestos mat, presumably because the impregnation technique leads to enhanced surface area and porosity. The two most promising coating systems found were 1:10-phenanthroline impregnated upon asbestos and a three component composite film of the co-polymer of Genamid 2000 and Epon 828 as a reactive binder with 1,10-phenanthroline as a reactive filler. The use of a reactive coating as a passive safety system should reduce appreciably the airborne iodine half-life and the hazards associated with iodine release during a nuclear reactor accident. (author)

  20. Fission Product Experimental Program: Validation and Computational Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Leclaire, N.; Ivanova, T.; Letang, E. [Inst Radioprotect and Surete Nucl, F-92262 Fontenay Aux Roses (France); Girault, E. [CEA Valduc, Serv Rech Neutron and Critcite, 21 - Is-sur-Tille (France); Thro, J. F. [AREVA NC, F-78000 Versailles (France)

    2009-02-15

    From 1998 to 2004, a series of critical experiments referred to as the fission product (FP) experimental program was performed at the Commissariat a l'Energie Atomique Valduc research facility. The experiments were designed by Institut de Radioprotection et de Surete Nucleaire (IRSN) and funded by AREVA NC and IRSN within the French program supporting development of a technical basis for burnup credit validation. The experiments were performed with the following six key fission products encountered in solution either individually or as mixtures: {sup 103}Rh, {sup 133}Cs, {sup nat}Nd, {sup 149}Sm, {sup 152}Sm, and {sup 155}Gd. The program aimed at compensating for the lack of information on critical experiments involving FPs and at establishing a basis for FPs credit validation. One hundred forty-five critical experiments were performed, evaluated, and analyzed with the French CRISTAL criticality safety package and the American SCALE5. 1 code system employing different cross-section libraries. The aim of the paper is to show the experimental data potential to improve the ability to perform validation of full burnup credit calculation. The paper describes three Phases of the experimental program; the results of preliminary evaluation, the calculation, and the sensitivity/uncertainty study of the FP experiments used to validate the APOLLO2-MORET 4 route in the CRISTAL criticality package for burnup credit applications. (authors)

  1. Very-long-term storage of fission products

    International Nuclear Information System (INIS)

    Sousselier, Y.; Pradel, J.; Cousin, O.

    The large majority of the fission products, with 99.9 percent of the radioactivity content, do not pose actual problems in storage in a geological formation for which we can guarantee total confinement. Safety of storage in a geological formation for the radionuclides of long half-life is based in particular on the absorption capacity of the geological formations and the example of the Oklo fossil reactor and the retention of Pu which is produced is a striking example. But the problems are not the same, and the properties that we look for in the terrain are not the same. We could thus be led to storage in different geological formations for the fission products and the long-half-life emitters. Their separation is interesting from this point of view, but the date at which the separation is made will not be necessarily that of reprocessing. But there is a question of one or the other, and these storages will offer a very high level of insurance and will present only the potential hazards that are very comparable with those presented by natural conditions

  2. Simulating irradiation hardening in tungsten under fast neutron irradiation including Re production by transmutation

    Science.gov (United States)

    Huang, Chen-Hsi; Gilbert, Mark R.; Marian, Jaime

    2018-02-01

    Simulations of neutron damage under fusion energy conditions must capture the effects of transmutation, both in terms of accurate chemical inventory buildup as well as the physics of the interactions between transmutation elements and irradiation defect clusters. In this work, we integrate neutronics, primary damage calculations, molecular dynamics results, Re transmutation calculations, and stochastic cluster dynamics simulations to study neutron damage in single-crystal tungsten to mimic divertor materials. To gauge the accuracy and validity of the simulations, we first study the material response under experimental conditions at the JOYO fast reactor in Japan and the High Flux Isotope Reactor at Oak Ridge National Laboratory, for which measurements of cluster densities and hardening levels up to 2 dpa exist. We then provide calculations under expected DEMO fusion conditions. Several key mechanisms involving Re atoms and defect clusters are found to govern the accumulation of irradiation damage in each case. We use established correlations to translate damage accumulation into hardening increases and compare our results to the experimental measurements. We find hardening increases in excess of 5000 MPa in all cases, which casts doubts about the integrity of W-based materials under long-term fusion exposure.

  3. Separation of fission products by the use of recoil; Separation des produits de fission par utilisation du recul

    Energy Technology Data Exchange (ETDEWEB)

    Henry, R; Beydon, J; Bardy, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    We have studied fission recoil in U{sub 3}O{sub 8} organic solvent mixtures. The organic phase chosen was first naphtalene then terphenyl. Graphite and activated carbon were also tried out as recoil media. We first verified that the fission fragments are ejected from the uranium oxide particles under our experimental conditions. The retention phenomenon observed is due to an adsorption occurring either during irradiation or during the chemical treatment. Using naphthalene or terphenyl, the individual separation of the fission products has made it possible to show the influence of the chemical nature of the recoil medium on the retention of each fission product. We put forward a hypothesis concerning this phenomenon: experiments carried out using 'scavengers', together with kinetic studies make it possible to explain the retention phenomenon and to choose the most favorable conditions for reducing this retention to a low value. The thermal recombination kinetics demonstrate the influence of the fission ion charge on the final value of the retention for a given temperature. The origins of this thermal recombination are discussed. (author) [French] On a etudie le recul de fission dans les melanges U{sub 3}0{sub 8}, phase organique. La phase organique choisie a ete le naphtalene puis le terphenyle. Le graphite et le charbon actif ont egalement ete essayes comme milieux de recul. On a d'abord determine que les fragments de fission sortent des particules d'oxyde d'uranium avec un rendement de 100 pour cent dans nos conditions experimentales. Le phenomene de retention observe est du a une adsorption ayant lieu pendant l'irradiation ou pendant le traitement chimique. Dans le naphtalene et le terphenyle, la separation individuelle des produits de fission a permis de mettre en evidence l'influence de la nature chimique du milieu de recul sur la retention de chaque produit de fission. On avance une hypothese sur ce phenomene: des experiences effectuees avec des 'scavengers

  4. Advances in Metallic Fuels for High Burnup and Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

    2016-10-01

    Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

  5. Background and derivation of ANS-5.4 standard fission product release model. Technical report

    International Nuclear Information System (INIS)

    1982-01-01

    ANS Working Group 5.4 was established in 1974 to examine fission product releases from UO2 fuel. The scope of ANS-5.4 was narrowly defined to include the following: (1) Review available experimental data on release of volatile fission products from UO2 and mixed-oxide fuel; (2) Survey existing analytical models currently being applied to lightwater reactors; and (3) Develop a standard analytical model for volatile fission product release to the fuel rod void space. Place emphasis on obtaining a model for radioactive fission product releases to be used in assessing radiological consequences of postulated accidents

  6. Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.

    1997-01-01

    Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)

  7. Methodology and application of the WIMS-D4M fission product data

    International Nuclear Information System (INIS)

    Mo, S.C.

    1995-01-01

    The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-chain model that includes the 135 Xe and 149 Sm chains, and a lumped fission product to account for the absorption by fission products not explicitly treated. Burn-up calculations were performed for the ANS MEU core using WIMS and EPRI-CELL cross sections. The calculated eigenvalues and material loadings are in good agreements

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

    International Nuclear Information System (INIS)

    Mellinger, G.B.; Jensen, G.A.

    1982-08-01

    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, Bi 2 O 3 , Sb 2 O 3 ). 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 Sb 2 O 3 , Bi 2 O 3 , 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 RuO 4 . A viscosity of approx. 20 poise at the processing temperature offered a good compromise between scavenger separation and Ru recovery. Glasses in which PbO was used as the scavenging agent were homogeneous in appearance. Resistance to leaching was close to that of certain waste glasses reported in the literature. 12 figures. 7 tables

  9. Development of nuclear transmutation technology for transuranic elements

    International Nuclear Information System (INIS)

    Mukaiyama, Takehiko

    1996-01-01

    Partitioning and Transmutation (P-T) of long-lived radioactive nuclides is conceived as the technology to improve the high-level radioactive waste management. This report discusses the incentives of P-T, generation of long-lived nuclides in fission reactors, nuclear transmutation technologies, R and D activities of the partitioning and transmutation technology development programs at JAERI and in the world. (author)

  10. Nuclearization of ionic chromatography system for fission products analysis

    International Nuclear Information System (INIS)

    Dimeglio, Remi

    1996-06-01

    The accident at Tchernobyl in 1986 had entailed the release in the atmosphere of different products coming from the splitting of the fuel. It is to better understand, and also to warn this type of catastrophe that the CEA (Commissariat a L'Energie Atomique) develops many programs of researches, aiming to characterize these fission products and to study their mechanisms of relaxation. Thus, the LESC (Laboratoire d'Etude de la Surete du Combustible) takes part, since several years, in many nuclear safety experiences, and in particular to the project PHEBUS PF, that is a reconstitution, in reduced scale, of an accident entailing the fusion of the reactor core. The aim of the researches that have been led during this training period was to the nuclearization of an HPIC (High Performance Ion Chromatography) system, dedicated to the analysis of the PHEBUS PF fission products analysis. The first step was to develop HPIC lines already settled, so as to reduce the quantity of wastes. Indeed, those one are very difficult to process in a radioactive area. For this purpose, we have implanted a column cationic more effective, so as to decrease analysis times, and, by there even, the quantity of sewage generated. We have equally replaced, on lines cationic and anionic, the system of suppression of the eluent conductivity, to make it thriftier in fluid. But the radioactive products characterization necessitates that all analyses are led within a special box with gloves. The second step of the project was therefore to adapt the system to this type of cell, and to its automation. It has been necessary to modify the system of sample injection, the system of detection, and to put in place a supplementary box with gloves, connected by sieve to the first, for the active products dilution. (author) [fr

  11. Recovery of fission products from acidic waste solutions thereof

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.; Dubois, D.W.

    1975-01-01

    Fission products, e.g., palladium, ruthenium and technetium, are removed from aqueous, acidic waste solutions thereof. The acidic waste solution is electrolyzed in an electrolytic cell under controlled cathodic potential conditions and technetium, ruthenium, palladium and rhodium are deposited on the cathode. Metal deposit is removed from the cathode and dissolved in acid. Acid insoluble rhodium metal is recovered, dissolved by alkali metal bisulfate fusion and purified by electrolysis. In one embodiment, the solution formed by acid dissolution of the cathode metal deposit is treated with a strong oxidizing agent and distilled to separate technetium and ruthenium (as a distillate) from palladium. Technetium is separated from ruthenium by organic solvent extraction and then recovered, e.g., as an ammonium salt. Ruthenium is disposed of as waste by-product. Palladium is recovered by electrolysis of an acid solution thereof under controlled cathodic potential conditions. Further embodiments wherein alternate metal recovery sequences are used are described. (U.S.)

  12. An application program for fission product decay heat calculations

    International Nuclear Information System (INIS)

    Pham, Ngoc Son; Katakura, Jun-ichi

    2007-10-01

    The precise knowledge of decay heat is one of the most important factors in safety design and operation of nuclear power facilities. Furthermore, decay heat data also play an important role in design of fuel discharges, fuel storage and transport flasks, and in spent fuel management and processing. In this study, a new application program, called DHP (Decay Heat Power program), has been developed for exact decay heat summation calculations, uncertainty analysis, and for determination of the individual contribution of each fission product. The analytical methods were applied in the program without any simplification or approximation, in which all of linear and non-linear decay chains, and 12 decay modes, including ground state and meta-stable states, are automatically identified, and processed by using a decay data library and a fission yield data file, both in ENDF/B-VI format. The window interface of the program is designed with optional properties which is very easy for users to run the code. (author)

  13. On fusion driven systems (FDS) for transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Aagren, O (Uppsala Univ., Aangstroem laboratory, div. of electricity, Uppsala (Sweden)); Moiseenko, V.E. (Inst. of Plasma Physics, National Science Center, Kharkov Inst. of Physics and Technology, Kharkov (Ukraine)); Noack, K. (Forschungszentrum Dresden-Rossendorf (Germany))

    2008-10-15

    This report gives a brief description of ongoing activities on fusion driven systems (FDS) for transmutation of the long-lived radioactive isotopes in the spent nuclear waste from fission reactors. Driven subcritical systems appears to be the only option for efficient minor actinide burning. Driven systems offer a possibility to increase reactor safety margins. A comparatively simple fusion device could be sufficient for a fusion-fission machine, and transmutation may become the first industrial application of fusion. Some alternative schemes to create strong fusion neutron fluxes are presented

  14. On fusion driven systems (FDS) for transmutation

    International Nuclear Information System (INIS)

    Aagren, O; Moiseenko, V.E.; Noack, K.

    2008-10-01

    This report gives a brief description of ongoing activities on fusion driven systems (FDS) for transmutation of the long-lived radioactive isotopes in the spent nuclear waste from fission reactors. Driven subcritical systems appears to be the only option for efficient minor actinide burning. Driven systems offer a possibility to increase reactor safety margins. A comparatively simple fusion device could be sufficient for a fusion-fission machine, and transmutation may become the first industrial application of fusion. Some alternative schemes to create strong fusion neutron fluxes are presented

  15. ESOL facility for the generation and radiochemical separation of short half-life fission products

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Meikrantz, D.H.; Baker, J.D.; Anderl, R.A.; Novick, V.J.; Greenwood, R.C.

    1988-01-01

    A facility has been developed at the Idaho National Engineering Laboratory (INEL) for the generation and rapid radiochemical separation of short half-life mixed fission products. This facility, referred to as the Idaho Elemental Separation On Line (ESOL), consists of electro-plated sources of spontaneously fissioning 252 Cf with a helium jet transport arrangement to continuously deliver short half-life, mixed fission products to the radiochemistry laboratory for rapid, computer controlled, radiochemical separations. 18 refs., 13 figs

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

    International Nuclear Information System (INIS)

    Ihara, Hitoshi

    1989-11-01

    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)

  17. Yields of products from thermal-neutron induced fission of 235U

    International Nuclear Information System (INIS)

    Rudstam, G.; Aagaard, P.; Zwicky, H.U.

    1985-01-01

    Methods for fission yield determinations at an ISOL-system connected to a nuclear reactor have been developed. The present report contains detailed descriptions both of the experimental techniques and of the method used to correct the experimental yields for the decay of short-lived nuclear species in the delay between production and measurement. The methods have been applied to the determination of the fission yields of 40 fission products including 2 isometric pairs in the light mass region and those of 99 fission products including 25 isometric pairs or triplets in the heavy mass region. For 64 cases this is the first determination published. (author)

  18. Fuel and target programs for the transmutation at Phenix and other reactors

    International Nuclear Information System (INIS)

    Gaillard-Groleas, G.

    2002-01-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

  19. Installation for the Mo-99 production from fission products

    International Nuclear Information System (INIS)

    Marques, R.O.; Cristini, P.R.; Marziale, D.P.; Furnari, E.S.; Fernandez, H.O.

    1988-01-01

    The installation to produce Mo-99 from nuclear fission started going on August 12th 1985 in Ezeiza Atomic Center. The characteristics of the process, the emplacement of the power plant, target, and irradiation conditions are presented. The targets are plates with a nucleus of Al/U alloy, with U-235 enriched to 90 % covered by Al plates. Each plate consists of about 1.10 -3 Kg of U-235 and 13.10 -3 Kg of Al. The plates are irradiated with a 3.10 13 n cm -2 s -1 flux during five days in the RA-3 nucleus. The Mo-99 separation method, is presented, where it is foreseen te I-131 separation. An account of the treatment of solid, liquid and gaseous waste is provided. An equipment to transfer the filter precipitate was designed in order to recover the U. The installation to continue the U recovery process, to separate I-131 and Xe-133 and to incorporate a Mo-99 purification stage for sublimation is being extended. (M.E.L.) [es

  20. Vortex transmutation.

    Science.gov (United States)

    Ferrando, Albert; Zacarés, Mario; García-March, Miguel-Angel; Monsoriu, Juan A; de Córdoba, Pedro Fernández

    2005-09-16

    Using group theory arguments and numerical simulations, we demonstrate the possibility of changing the vorticity or topological charge of an individual vortex by means of the action of a system possessing a discrete rotational symmetry of finite order. We establish on theoretical grounds a "transmutation pass" determining the conditions for this phenomenon to occur and numerically analyze it in the context of two-dimensional optical lattices. An analogous approach is applicable to the problems of Bose-Einstein condensates in periodic potentials.

  1. Accelerator-driven sub-critical target concept for transmutation of nuclear wastes

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Todosow, M.; Aronson, A.L.; Takahashi, H.; Geiger, M.J.

    1991-01-01

    A means of transmuting key long-lived nuclear wastes, primarily the minor actinides (Np, Am, Cm) and iodine, using a hybrid proton accelerator and sub-critical lattice, is proposed. By partitioning the components of the light water reactor (LWR) spent fuel and by transmuting key elements, such as the plutonium, the minor actinides, and a few of the long-lived fission products, some of the most significant challenges in building a waste repository can be substantially reduced. The proposed machine, based on the described PHOENIX Concept, would transmute the minor actinides and the iodine produced by 75 LWRs, and would generate usable electricity (beyond that required to run the large accelerator) of 850 MW e . 19 refs., 20 figs

  2. Separation and transmutation. A picture of the applications in Sweden; Separation och transmutation. Belysning av tillaempning i Sverige

    Energy Technology Data Exchange (ETDEWEB)

    Grundfelt, Bertil; Lindgren, Maria [Kemakta Konsult AB, Stockholm (Sweden)

    2006-11-15

    This report contains a review of the transmutation technology and an elucidation of the consequences of the use of transmutation in Sweden. Transmutation has often been mentioned in the public debate as a way of rendering harmless the high-level waste from nuclear power such that the final disposal of the waste can be substantially simplified or even completely avoided. However, it can be noted that even with an exploitation of transmutation, significant amounts of radioactive waste requiring qualified final disposal will be generated. The transmutation technology will make it possible to reduce the longevity of the high-level waste by converting primarily the transuranic elements to fission products with shorter half lives. The long-term radiotoxicity of the spent nuclear fuel is dominated by the transuranics. Hence, transmutation will lead to a substantial decrease of the long-term radiotoxicity of the spent fuel. The research on transmutation has been focussed on sub-critical so called ADS-reactors (Accelerator Driven System). In such a system protons are accelerated to very high energy levels (in the order of GeV) in an electromagnetic field. The accelerated protons are impacted on a spallation source consisting of heavy atoms, e.g. lead or a mixture of lead and bismuth. At the impact the heavy nuclei are spalled releasing a number of neutrons that can be used for fissioning the nuclei of the substances to be transmuted, primarily the transuranics. ADS-reactors are still at the research stage. It is a common view that it will take several decades before the technology has reached a maturity that allows the construction of a demonstration facility. Calculations performed at Royal Institute of Technology in Stockholm show that using the ADS-technology would allow a reduction of the inventory of transuranics in the spent fuel from Swedish reactors by 50-85% within a 50-100 years period. The goal to transmute 99% of the transuranics inventory has been achieved in

  3. Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

    International Nuclear Information System (INIS)

    Bultman, J.

    1992-04-01

    Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

  4. Preliminary results utilizing high-energy fission product γ-rays to detect fissionable material in cargo

    Science.gov (United States)

    Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Church, J. A.; Descalle, M. A.; Gosnell, T. B.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Mauger, G. J.; Moore, T. L.; Norman, E. B.; Pohl, B. A.; Pruet, J. A.; Petersen, D. C.; Walling, R. S.; Weirup, D. L.; Prussin, S. G.; McDowell, M.

    2005-12-01

    A concept for detecting the presence of special nuclear material (235U or 239Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their β-delayed neutron emission or β-delayed high-energy γ radiation between beam pulses provide the detection signature. Fission product β-delayed γ-rays above 3 MeV are nearly 10 times more abundant than β-delayed neutrons and are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified.

  5. Transient fission-product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

    1997-12-01

    Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

  6. Early history of soil contamination with fission products

    International Nuclear Information System (INIS)

    Nagel, E.

    1985-01-01

    Underneath a balloon and instrument hut demolished today and on the natural pasture next to it, a series of 2 to 4-layer soil samples was carried out to determine the contents in Cs-137 and Sr-90. The results showed a sudded decrease of the Cs concentration behind the walls of the hut, but no further decrease towards the centre of the hut. As expected, the decrease of Sr-90 concentration was slower, both in a horizontal line and in the depth. The results reveal that the soil underneath the hut has not received further fission products since it was built in 1956 from depositions of later nuclear weapon tests. Furthermore, the radionuclides were not transported much further into lower strata so that by taking into consideration of the half-life of the two nuclides their concentration in the soil can be traced back to the year 1956 for comparison with the results measured at the time. (orig./HP) [de

  7. ENDF/B-5 fission product cross section evaluations

    International Nuclear Information System (INIS)

    Schenter, R.E.; England, T.R.

    1979-12-01

    Cross section evaluations were made for the 196 fission product nuclides on the ENDF/B-5 data files. Most of the evaluations involve updating the capture cross sections of the important absorbers for fast and thermal reactor systems. This included updating thermal values, resonance integrals, resonance parameter sets, and fast capture cross sections. For the fast capture results generalized least-squares calculations were made with the computer code FERRET. Input for these cross section adjustments included nuclear models calculations and both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, 4000. Comparisons of these evaluations with recent capture measurements are shown. 15 figures, 10 tables

  8. Modeling of fission product release in integral codes

    International Nuclear Information System (INIS)

    Obaidurrahman, K.; Raman, Rupak K.; Gaikwad, Avinash J.

    2014-01-01

    The Great Tohoku earthquake and tsunami that stroke the Fukushima-Daiichi nuclear power station in March 11, 2011 has intensified the needs of detailed nuclear safety research and with this objective all streams associated with severe accident phenomenology are being revisited thoroughly. The present paper would cover an overview of state of art FP release models being used, the important phenomenon considered in semi-mechanistic models and knowledge gaps in present FP release modeling. Capability of FP release module, ELSA of ASTEC integral code in appropriate prediction of FP release under several diversified core degraded conditions will also be demonstrated. Use of semi-mechanistic fission product release models at AERB in source-term estimation shall be briefed. (author)

  9. Diffusion of Fission Product Elements in Compacted Bentonite

    International Nuclear Information System (INIS)

    Pratomo-Budiman-Sastrowardoyo; Dewi-Susilowati; Dadang-Suganda

    2000-01-01

    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)

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

    International Nuclear Information System (INIS)

    Henderson, R.W.

    1991-01-01

    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

  11. Evaluation of fission product neutron cross sections for JENDL

    International Nuclear Information System (INIS)

    1984-01-01

    The recent activities on the evaluation of fission product (FP) neutron cross sections for JENDL (Japanese Evaluated Nuclear Data Library) are presented briefly. The integral test of JENDL-1 FP cross section file was performed using the CFRMF sample activation data and the STEK sample reactivity data, and the ratio of experiment to calculation was nearly constant for all the samples in the STEK measurement. Therefore, a tentative analysis was performed by applying the correction to the calculated scattering reactivity component. Better agreement with the experiment was obtained after applying this correction in most cases. The evaluation work on the JENDL-2 FP neutron cross sections is now in progress. The improvement of the data evaluation is presented in an itemized form. The JENDL-2 FP file will contain the evaluated data for 100 nuclides from Kr to Tb. The improvement and the future scope of the integral test for JENDL-2 FP data are summarized. (Asami, T.)

  12. A model for fission product distribution in CANDU fuel

    International Nuclear Information System (INIS)

    Muzumdar, A.P.

    1983-01-01

    This paper describes a model to estimate the distribution of active fission products among the UO 2 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 I 131 . The model also appears to generally agree with the ''free'' inventory release calculated using ANS-5.4. (author)

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

    International Nuclear Information System (INIS)

    Schrock, V.E.

    1978-01-01

    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 10 4 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 10 9 seconds, but provides only an ''upper bound'' on the capture correction in the interval 10 4 9 seconds. (author)

  14. Monitoring of fission products through on-line gamma spectrometry

    International Nuclear Information System (INIS)

    Montagnon, F.; Warlop, R.

    1989-01-01

    Under normal operating conditions, the monitoring of the possible deterioration of the pressurized water reactor core fuel rods is achieved through analysis of the radioactive fission products carried by the primary system. For acquiring results of spectrometric analyses in real time, and avoiding risks of errors linked to manual operations, CEA/DMG and EDF/SEPTEN have jointly developed an entirely automatic system. This system allows measuring permanently the primary system activity of two coupled units, with no human operation nor any handling of active coolant specimens. The PIGAL facility has been set up in the nuclear auxiliary building, common to the two units, and it is used on a demonstration basis for units 2 and 3 of the BUGEY site. This device has been patented

  15. Fission product poisoning in KS-150 reactor operation

    International Nuclear Information System (INIS)

    Rana, S.B.

    1978-01-01

    A three-dimensional model of the KS-150 reactor was used to study reactivity changes induced by reactor poisoning with fission products Xe 135 and Sm 149 . A comparison of transients caused by the poisoning showed the following differences: (1) the duration of the transient Xe poisoning (2 days) is shorter by one order of magnitude than the duration of Sm poisoning (20 days); however, the level of Xe poisoning is greater approximately by one order than the level of the Sm poisoning; (2) the level of steady-state Xe poisoning depends on the output level of the reactor; steady-state Sm poisoning does not depend on this level; (3) following reactor shutdown Xe poisoning may increase to the maximum value of up to Δrhosub(Xe)=20% and will then gradually decrease; Sm poisoning may reach maximum values of up to Δrhosub(Sm)=2% and does not decrease. (J.B.)

  16. Partitioning and Transmutation. Annual Report 2004

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Sofie; Drouet, Francois; Ekberg, Christian; Liljenzin, Jan-Olov; Magnusson, Daniel; Nilsson, Mikael; Retegan, Teodora; Skarnemark, Gunnar [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Materials and Surface Chemistry

    2005-01-01

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products ({sup 129}I, {sup 99}Tc, {sup 135}Cs, {sup 93}Zr and {sup 126}Sn and activation products ({sup 14}C and {sup 36}Cl). To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. This separation is necessary to obtain the desired efficiency in the transmutation process in order not to create any unnecessary waste thus rendering the process useless. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. Chalmers University of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in the European Union sixth framework program project EUROPART. This is a continuation of the projects we participated in within the fourth and fifth framework programmes NEWPART and PARTNEW respectively. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development.

  17. Partitioning and Transmutation. Annual Report 2005

    International Nuclear Information System (INIS)

    Andersson, Sofie; Ekberg, Christian; Fermvik, Anna; Hervieux, Nadege; Liljenzin, Jan-Olov; Magnusson, Daniel; Nilsson, Mikael; Retegan, Teodora; Skarnemark, Gunnar

    2006-01-01

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products ( 79 Se, 87 Rb, 99 Tc, 107 Pd, 126 Sn, 129 I, 135 Cs) and activation products ( 14 C, 36 Cl, 59 Ni, 93 Zr, 94 N To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. These separations are necessary to obtain the desired efficiency of the transmutation process and in order not to create any unnecessary waste thus rendering the process useless. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. Chalmers Univ. of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in the EUROPART project within the European Union sixth framework program. This is a continuation of the projects we participated in within the fourth and fifth framework programmes, NEWPART and PARTNEW respectively. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. However, since the basic understanding is still needed we have our main focus on the chemical processes and understanding of how they work

  18. The Phebus Fission Product and Source Term International Programmes

    International Nuclear Information System (INIS)

    Clement, B.; Zeyen, R.

    2005-01-01

    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 5 th 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)

  19. Analysis of Dust and Fission Products in PBMR Turbine

    International Nuclear Information System (INIS)

    Stempniewicz, M.M.; Wessels, D.

    2014-01-01

    A 400 MWth direct cycle Pebble Bed Modular reactor was under development in South Africa. The work performed included design and safety analyses. In HTR/PBMR, graphite dust is generated during normal reactor operation due to pebble-to-pebble scratching. This dust will be deposited throughout the primary system. Furthermore, the dust will become radioactive due to sorption of fission products released, although in very small quantities, during normal operation. This paper presents a model and analyses of the PBMR turbine with the SPECTRA code. The purpose of the present work was to estimate the amount and distribution of deposited dust and the fission products, namely cesium, iodine, and silver, during plant life-time, which was assumed to be 40 full-power years. The performed work showed that after 40 years of plant life-time deposited layers are very small. The largest deposition is of course observed on the dust filters. Apart from the dust filters, the largest dust deposition is observed on the: • Outer Casing (inner walls) • Turbine Rotor Cooling Cavity (inner walls) • HPC Cold Cooling Gas Header (inner walls) This is caused by relatively low gas velocities in these volumes. The low velocities allow a continuous build-up of the dust layer. About 90% of cesium, 40% of iodine, and 99.9% of silver is adsorbed on the metallic structures of the turbine. The sorption rate increases along the turbine due to decreasing temperatures. In case of cesium and iodine the highest concentrations are observed in the last stage (stage 12) of the turbine. In the case of silver the sorption is so large that the silver vapor is significantly depleted in the last stages of the turbine. This is a reason for having a maximum in silver concentration in the stage 10. In the following stages the concentration decreases due to very small silver vapor fraction in the gas. (author)

  20. Partitioning and Transmutation. Annual Report 2006

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, Isabelle; Englund, Sofie; Fermvik, Anna; Liljenzin, Jan-Olov; Neumayer, Denis; Retegan, Teodora; Skarnemark, Gunnar [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Chemical and Biological Engineering

    2007-01-15

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products ({sup 79}Se, {sup 87}Rb, {sup 99}Tc, {sup 107}Pd, {sup 126}Sn, {sup 129}I, {sup 135}Cs) and activation products ({sup 14}C, {sup 36}Cl, {sup 59}Ni, {sup 93} Zr, {sup 94} To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. These separations are necessary to obtain the desired efficiency of the transmutation process and in order not to create any unnecessary waste thus rendering the process useless. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. Chalmers University of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in the EUROPART project within the European Union sixth framework program. This is a continuation of the projects we participated in within the fourth and fifth framework programmes, NEWPART and PARTNEW, respectively. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. However, since the basic understanding is still needed we have our main focus on the chemical processes and understanding of how they work. Work is progressing in relation to a proposal for the 7th framework programme. This proposal will be aiming at a pilot plant for separation for transmutation purposes.

  1. The Impact of Partitioning and Transmutation on the Risk Assesment of a Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Amrani, Naima

    2006-01-01

    Partitioning and transmutation of radioactive and long lived component from the highly radioactive waste stream in order to reduce or probably eliminate their radiotoxic inventory was the important option for the nuclear waste management. The principal radionuclides contribution to the long term radiotoxic inventory is mostly due to Pu, minor actinides and some long-lived fission products. The conditioning operation can present artificial barriers which are potentially capable of confining the radionuclides within their package for thousands of years. After this time nothing can be predicted. The solubility of the actinides (except Np) is generally low whereas the long lived fission product. particularly 137 Cs, 129 I and in some case 99 Tc, display high mobility In the geosphere. Conditioning of separated long-lived nuclides in appropriate matrices which could serve as irradiation matrix in a delayed transmutation option is a possible outcome for the next decades. The general strategy of introducing Partitioning and Transmutation as an additional waste management option is based on the radiological benefit which is expected from such an option. The short term impact of partitioning would be to reduce long-term radiotoxic inventory of the resulting HLW at the expense of an increase of the operational requirements for the nuclear facilities concerned. Fast neutron spectrum devices (FR or ADS facilities) are more efficient than current LWRs for recycling and transmuting long-lived radionuclides

  2. Transmutation technologies to solve the problem of long-term spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Hosnedl, P.; Valenta, V.; Blahut, O.

    2000-01-01

    The paper gives a brief description of the transmutation process for actinides and long-lived fission products which are present in spent nuclear fuel. Transmutation technologies can solve the problem of long-term spent nuclear fuel storage and reduce the requirements for storage time and conditions. The basic data and requirements for the detailed design of the transmutor are summarized, and the views upon how to address the fuel purification and dry reprocessing issues are discussed. The results of activities of SKODA JS are highlighted; these include, for instance, the fluoride salt-resistant material MONICR, test loops, and electrowinners. The preliminary design of the transmutor is also outlined. Brief information regarding activities in the field of transmutation technologies in the Czech Republic and worldwide is also presented. The research and design activities to be developed for the whole design of the demonstration and basic units are summarized. It is emphasized that SKODA JS can join in international cooperation without constraints. The Attachment presents a simple assessment of how the radioactivity balance can be reduced, based on the actinide and long-lived fission product transmutation half-lives, is presented in the Attachment. (author)

  3. Waste partitioning and transmutation as a means towards long-term risk reduction

    International Nuclear Information System (INIS)

    Merz, E.R.

    1993-09-01

    It has been an idea for some time to reduce the long-term potential hazard of the waste by chemical removal of the actinides as well as some long-lived fission products and their subsequent transmutation in an intense neutron flux. Transmutation would thus shorten the required containment period of radioactive material in a repository. It is estimated, that development of such technology would take at least 40 years because facilities would be required to perform a clean actinide and fission product isolation and to fabricate the fuel elements that contained the separated nuclides. This latter requirements would involve a major expansion of new chemical process steps which are not available as yet. Development of new equipment to maintain occupational exposures as low as reasonably achievable and to minimize releases of radioactivity to the environment would also be necessary. Partitioning and transmutation should be introduced, if at all, as a long-term decision about new nuclear power technology as a future energy source. With regard to this, R and D work dealing with basic questions seems to be worthwhile, However, the introduction of partitioning and transmutation will not eliminate the need for radioactive waste disposal. (orig./HP) [de

  4. Energy distribution of antineutrinos originating from the decay of fission products in a nuclear reactor

    International Nuclear Information System (INIS)

    Rudstam, G.; Aleklett, K.

    1979-01-01

    The energy spectrum of antineutrinos around a nuclear reactor has been derived by summing contributions from individual fission products. The resulting spectrum is weaker at energies above approx. 8 MeV than earlier published antineutrino spectra. The reason may be connected to the strong feeding of high-lying daughter states in the beta decay of fission products with high disintegration energies

  5. Role of fission product in whole core accidents: research in the USA

    International Nuclear Information System (INIS)

    Jackson, J.F.; Deitrich, L.W.

    1977-01-01

    The techniques being developed in the United States for analyzing postulated whole-core accidents in LMFBRs are briefly reviewed. The key mechanistic analysis methods are discussed in detail. Important research projects in the area of fission product effects are examined. Some typical results on the role of fission products in whole-core accidents are presented

  6. Calculation and Evaluation of Fission Yields and Capture Cross Sections Leading to the Production of Therapeutic Radionuclide by Means of Nuclear Reactors

    International Nuclear Information System (INIS)

    Sublet, J.C.

    2009-01-01

    Much progress has been made in nuclear medicine that involves the use of radionuclides for both diagnosis and therapy. Because of this qualitative and quantitative growth, the adoption of a set of established radionuclides for various applications, the methods of nuclide production need to be addressed and consideration given to other, emerging radionuclides that are judged to be developing in importance. The methods involved are characterized by the transmutation of isotopes by neutron-induced reactions and decays. Therefore, newly evaluated cross sections, fission yields and decay characteristics of relevance to the reactor production of those therapeutic radionuclides have been reviewed. Considerations of the decay schemes of all the nuclides involved are also included. (author)

  7. Impact of partitioning and transmutation on repository design

    International Nuclear Information System (INIS)

    Carter, D. 'Buzz' Savage

    2004-01-01

    The U.S. Department of Energy's Advanced Fuel Cycle Initiative (AFCI) program is investigating spent nuclear fuel treatment technologies that have the potential to improve the performance of the proposed geologic repository at Yucca Mountain. Separating actinides and selected fission products from spent fuel, storing some of them as low level waste and transmuting them in thermal and/or fast reactors has the potential to reduce the volume, short and long-term heat load and radiotoxicity of the high level waste destined for the repository, effectively increasing its capacity by a factor of 50 or more above the current legislative limit. (author)

  8. Calculated investigation of actinide transmutation in the BOR-60 reactor

    International Nuclear Information System (INIS)

    Zhemkov, I.Yu.; Ishunina, O.V.; Yakovleva, I.V.

    2001-01-01

    In the course of reactor operation the formation of fission products and accumulation of minor-actinides and plutonium take place in the nuclear fuel. These materials define the radiation hazard to a great extent. Of one possible ways lowering the activity of irradiated nuclear fuel is transmutation of long-lived radioactive isotopes in the stable or short-lived ones, that allows to facilitate the problem of the high-level waste and to improve the efficiency of nuclear fuel use at the expense of its recycling and burnup increasing. (authors)

  9. The status of fission product yield data (FPND) in 1977

    International Nuclear Information System (INIS)

    Cuninghame, J.G.

    1977-05-01

    The topics covered is this paper are:- (a) cumulative yields in thermal neutron fission and in fast fission up to 14 MeV incident neutron energy; (b) dependence of the yields on incident neutron energy and spectrum; (c) independent yields; (d) charge dispersion and distribution, and (e) yields of light particles from ternary fission. The paper reviews information on these subjects for fission of actinides from 232 Th upwards, with special emphasis on data published since the 1973 Bologna FPND Panel, compares data sets and discusses the gaps still to be found in them. (author)

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

    International Nuclear Information System (INIS)

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

    2003-01-01

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

  11. Plutonium and surrogate fission products in a composite ceramic waste form

    International Nuclear Information System (INIS)

    Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

    1999-01-01

    Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known

  12. The wastes of nuclear fission; Les dechets de la fission nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Doubre, H. [Paris-11 Univ., Centre de Spectrometrie Nucleaire et de Spectrometrie de Masse, IN2P3/CNRS, 91 - Orsay (France)

    2005-07-01

    In this paper the author presents the problems of the radioactive wastes generated by the nuclear fission. The first part devoted to the fission phenomenon explains the incident neutron energy and the target nuclei role. The second part devoted to the nuclear wastes sources presents the production of wastes upstream of the reactors, in the reactors and why these wastes are dangerous. The third part discusses the radioactive wastes management in France (classification, laws). The last part details the associated research programs: the radionuclides separation, the disposal, the underground storage, the transmutation and the thorium cycle. (A.L.B.)

  13. Measurement of the fission yields of selected prompt and decay fission product gamma-rays of spontaneously fissioning 252Cf and 244Cm

    International Nuclear Information System (INIS)

    Reber, E.L.; Gehrke, R.J.; Aryaeinejad, R.; Hartwell, J.K.

    2005-01-01

    Gamma-ray spectrometry measurements have been made of the fission yields of selected γ -rays emitted by the spontaneously fissioning isotopes 252 Cf and 244 Cm. The measured γ-rays were selected based on their relative abundance in the spectrum and their freedom from interference or, in a few instances, ease of interference correction. From these data and the cumulative and independent yield data of England and Rider, those γ-rays that are primarily produced by radioactive decay, as opposed to direct yield, were converted into the decays per spontaneous fission expressed in percent and compared to cumulative yield values of England and Rider. For those γ-rays whose production is dominated by direct (independent) yield, the ratio of γ-rays per spontaneous fission is reported. The γ-ray yield can be compared to the independent yield values of England and Rider when 100% of the direct feeding passes through the γ-ray. In those cases where both cumulative and independent yields contribute to the observed γ-ray emission rate, a direct comparison is not possible but a method to quantify the contribution from each is proposed. (author)

  14. Vortex Transmutation

    International Nuclear Information System (INIS)

    Ferrando, Albert; Garcia-March, Miguel-Angel; Zacares, Mario; Monsoriu, Juan A.; Cordoba, Pedro Fernandez de

    2005-01-01

    Using group theory arguments and numerical simulations, we demonstrate the possibility of changing the vorticity or topological charge of an individual vortex by means of the action of a system possessing a discrete rotational symmetry of finite order. We establish on theoretical grounds a 'transmutation pass rule' determining the conditions for this phenomenon to occur and numerically analyze it in the context of two-dimensional optical lattices. An analogous approach is applicable to the problems of Bose-Einstein condensates in periodic potentials

  15. Application of dynamic pseudo fission products and actinides for accurate burnup calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

    1996-09-01

    The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (author)

  16. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  17. Brief description of out-of-pile test facilities for study in corrosion and fission product behaviour in flowing sodium

    International Nuclear Information System (INIS)

    Iizawa, K.; Sekiguchi, N.; Atsumo, H.

    1976-01-01

    The experimental methods to perform tests for study in corrosion and fission products behaviour in flowing sodium are outlined. Flow diagrams for the activated materials and fission products behaviour test loop are given

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

    International Nuclear Information System (INIS)

    Bail, A.

    2009-05-01

    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 235 U(n th ,f), 239 Pu(n th ,f) and 241 Pu(n th ,f). For these higher masses an isotopic separation is no longer possible. So, a new method was undertaken with the reaction 239 Pu(n th ,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)

  19. Facilities for preparing actinide or fission product-based targets

    CERN Document Server

    Sors, M

    1999-01-01

    Research and development work is currently in progress in France on the feasibility of transmutation of very long-lived radionuclides such as americium, blended with an inert medium such as magnesium oxide and pelletized for irradiation in a fast neutron reactor. The process is primarily designed to produce ceramics for nuclear reactors, but could also be used to produce targets for accelerators. The Actinide Development Laboratory is part of the ATALANTE complex at Marcoule, where the CEA investigates reprocessing, liquid and solid waste treatment and vitrification processes. The laboratory produces radioactive sources; after use, their constituents are recycled, notably through R and D programs requiring such materials. Recovered americium is purified, characterized and transformed for an experiment known as ECRIX, designed to demonstrate the feasibility of fabricating americium-based ceramics and to determine the reactor transmutation coefficients.

  20. Nudatra: nuclear data for transmutation in IP-Eurotrans

    International Nuclear Information System (INIS)

    Gonzalez, E.M.; Koning, A.; Leray, S.; Plompen, A.; Sanz, J.

    2007-01-01

    The objective of NUDATRA, Domain 5 of the EU Integrated Project EUROTRANS (FI6W-CT-2004- 516520), is to improve and validate the nuclear data and simulation tools required for the development and optimisation of nuclear waste transmutation, ADS dedicated transmutation systems and the associated fuel cycle. Activities are essentially aimed at supplementing the evaluated nuclear data libraries and improving the reaction models for materials in transmutation fuels, coolants, spallation targets, internal structures, and reactor and accelerator shielding, relevant for the design and optimisation of the ETD and XT-ADS. These activities are distributed over four Work Packages: Sensitivity Analysis and Validation of Nuclear Data and Simulation Tools; Low- and Intermediate-energy Nuclear Data Measurements; Nuclear Data Libraries Evaluation and Low-intermediate Energy Models; and High-energy Experiments and Modelling.The main accomplishments expected from NUDATRA are: 1) new measurements and evaluations of Pb-Bi cross-sections, i.e. inelastic, (n,xn) and isomer branching ratios (Po production); 2) new measurements and evaluations for minor actinides particularly the capture in 243 Am and fission on 244 Cm; 3) improvement of TALYS as an evaluation tool and as an a priori model for the estimation of low- and intermediate-energy reaction cross-section; 4) high-energy model improvement based on measurements, particularly for the prediction of the spallation products, and gas (H, He) production cross-sections; 5) sensitivity and uncertainty analysis of ETD fuel cycle and related covariance issues. (authors)

  1. 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

  2. Evaluation of Neutron Induced Reactions for 32 Fission Products

    International Nuclear Information System (INIS)

    Kim, Hyeong Il

    2007-02-01

    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, 95 Mo, 101 Ru, 103 Rh, 105 Pd, 109 Ag, 131 Xe, 133 Cs, 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 sections for almost all reaction channels

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

    International Nuclear Information System (INIS)

    Dickens, J.K.; Emery, J.F.; Love, T.A.; McConnell, J.W.; Northcutt, K.J.; Peelle, R.W.; Weaver, H.

    1977-10-01

    Fission-product decay energy-releases rates were measured for thermal-neutron fission of 235 U. 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

  4. Transmutation of transuranium elements in a gas-cooled accelerator-driven system

    International Nuclear Information System (INIS)

    Biss, Klaus Hendrik

    2014-01-01

    that a transmutation time of more than 338 years, even at low reprocessing losses of 0.1 %, is necessary for transmutation efficiency of 98.8 %. One reason for this duration is due to a reprocessing time of 10 years. The development of repository-relevant quantities radiotoxicity, decay heat and activity for transuranic elements on repository periods up to 1 million years is presented for direct disposal, after the P and T-process and for not transmutable assumed waste. The waste is divided into already vitrified heat-generating waste, uranium and fission products. Efficiencies for decay heat and radiotoxicity are deduced from the vitrified waste, which limits the additional contribution of transmuted transuranic elements to this level. The efficiencies differ with 96 % and 96.6 %, only slightly from each other and lead to transmutation durations of 232 and 245 years. With the consideration of the initial hazard potential, the simulation of the operation of a transmutation system, the calculation of the final storage masses after the P and T-phase, the development of the resulting repository-relevant variables, and the consideration of the transmutation time, this thesis provides an in-depth picture of the transmutation process.

  5. Design considerations of fission and corrosion product in primary system of MONJU

    International Nuclear Information System (INIS)

    Yanagisawa, T.; Akagane, K.; Yamamoto, K.; Kawashima, K.

    1976-01-01

    General influence of fission and corrosion products in primary system on MONJU plant design is reviewed. Various research and development works are now in progress to decrease the generation rate, to remove the products more effectively and to develop the methods of evaluation the behaviour of radioactive products. The inventory and distribution of fission and corrosion products in the primary circuit of MONJU are given. The radiation levels on the primary components are estimated to be several roentgens per hour. (author)

  6. Calculating the mass distribution of heavy nucleus fission product by neutrons

    International Nuclear Information System (INIS)

    Gudkov, A.N.; Koldobskij, A.B.; Kolobashkin, V.M.; Semenova, E.V.

    1981-01-01

    The technique of calculating the fission product mass yields by neutrons which are necessary for performing nucleus physical calculations in designing nuclear reactor cores is considered. The technique is based on the approximation of fission product mass distribution over the whole mass range by five Gauss functions. New analytical expressions for determining energy weights of used gaussians are proposed. The results of comparison of experimental data with calculated values for fission product mass obtained for reference processes in the capacity of which the fission reactions are chosen: 233 U, 235 U fission by thermal neutrons, 232 Th, 233 U, 235 U, 238 U by fission spectrum neutrons and 14 MeV neutrons and for 232 Th fission reactions by 11 MeV neutrons and 238 U by 7.7 MeV neutrons. On the basis of the analysis of results obtained the conclusion is drawn on a good agreement of fission product mass yield calculation values obtained using recommended values of mass distribution parameters with experimental data [ru

  7. Research on transmutation and accelerator-driven systems at the Forschungszentrum Karlsruhe

    International Nuclear Information System (INIS)

    Knebel, J.U.; Heusener, G.

    2000-01-01

    Transmutation is considered a promising technology worldwide for significantly reducing the amount and, thereby, the long-term radiotoxicity of high active waste (HAW) produced by the operation of nuclear power plants such as light water reactors (LWR). The maximum reduction of radiotoxicity could be by a factor of about 100. Transmutation is thus an alternative to the direct deposition of large volumes of highly radioactive waste. Transmutation presents the possibility of closing the fuel cycle including the minor actinides. Plutonium, minor actinides and long-lived fission products can be transmuted in a so called Accelerator Driven Sub-critical System (ADS), which consists of an accelerator, a target module and a subcritical blanket. This paper describes the work performed at Forschungszentrum Karlsruhe which is critically evaluating an ADS mainly with respect to its potential for transmuting minor actinides, to its feasibility and to safety aspects. The work is being done in the area of core design, neutronics, safety, system analyses, materials and corrosion. (orig.) [de

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

    International Nuclear Information System (INIS)

    Soriano R, J. M.

    2011-01-01

    In this research work, sorption properties of hydroxyapatite in aqueous solutions were studied using Na + and K + ion behavior. In addition, the fission products 99 Tc and 107 Pd 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 24 Na and 42 K 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: 99m Tc and 109 Pd. Tc presented a low sorption affinity on hydroxyapatite in aqueous medium 0.02 M NaH 2 PO 4 and the results also show that Tc is not sorbed from perchlorate medium (0.01 M Ca(ClO 4 ) 2 ). Sorption behaviour of Pd(II) on hydroxyapatite was studied for different experimental conditions, with parameter such as: ph, aqueous medium (0.01 M NaClO 4 , 0.01 M and 0.025 M Ca(ClO 4 ) 2 , and 0.02 M NaH 2 PO 4 ), 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

  9. Radiation Damage and Fission Product Release in Zirconium Nitride

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-08-29

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

  10. Chemical reactions of fission products with ethylene using the gas jet technique

    International Nuclear Information System (INIS)

    Contis, E.T.; Rengan, Krish; Griffin, Henry C.

    1994-01-01

    An understanding of the nature of the chemical reactions taking place between fission products and their carrier gases, and the designing of a fast separation procedure were the purposes of this investigation. Chemical reactions of short-lived (less than one minute half-life) fission products with carrier gases lead to various chemical species which can be separated in the gas phase. The Gas Jet Facility at the Ford Nuclear Reactor was used to study the yields of volatile selenium and bromine fission products of 235 U using a semi-automatic batch solvent extraction technique. Heptane and water were used as organic and inorganic solvents. A carrier gas mixture of ethylene to pre-purified nitrogen (1 : 3) was used to sweep the fission products from the target to the chemistry area for analysis. The results indicated that the volatile selenium products generated by the interaction of selenium fission fragments with ethylene were predominantly organic in nature (84%), possibly organoselenides. The selenium values were used to resolve the fractions of the bromine nuclides, which come from two major sources, viz., directly from fission and from the beta-decay of selenium. The data showed that the fractions of independent bromine fission products in the organic phase were much lower compared to selenium; the bromine values range from 10 to 22% and varied with mass number. Results indicated that the bromine products were inorganic in nature, as possibly hydrogen chloride. ((orig.))

  11. Convective-diffusive transport of fission products in the gap of a failed fuel element

    International Nuclear Information System (INIS)

    Lian, Z.W.; Carlucci, L.N.; Arimescu, V.I.

    1995-03-01

    A model is presented to describe the transport behaviour of gaseous fission products along the axial fuel-to-sheathe gap of a failed fuel element to the coolant system. The model is applicable to an element having failed under normal operating conditions or loss-of coolant-accident conditions. Because of the large differences in operating parameters, the transport characteristics of gaseous fission products in a failed element under these two operating conditions are significantly different. However, in both cases the transport process can be described by convection-diffusion caused by the continuous release of fission products from the fuel to the gap. Under normal operating conditions, the bulk-flow velocity is found to be negligible, due to the low release rate of fission products from fuel. The process can be well approximated by the diffusion of fission products in a stagnant gas-steam mixture. The effect of convection on the fission product transport, however, becomes significant under loss-of-coolant-accident conditions, where the release rates of fission products from fuel can be several orders of magnitude higher that that under normal operating conditions. The convection of the mixture in the gap not only contributes an additional flux to the gas-mixture transport, but also increases the gradient of fission products concentration across the opening, and therefore increases the diffusion flux to the coolant. As a result of the bulk flow, the transport of fission products along the gap is accelerated and the hold-up of short-lived isotopes in the gap is significantly reduced. Steam ingress through the opening into the gap is obstructed by the bulk flow, resulting in low steam concentrations in the gap under loss-of-coolant-accident conditions. (author). 6 refs., 8 figs

  12. Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

  13. Transport and release of fission products during nuclear reactor accident

    International Nuclear Information System (INIS)

    Lee, K.W.; Kuhlman, M.R.; Gieseke, J.A.

    1984-01-01

    This study represents the identification and formulation of a systematic, mechanistic approach to estimating source terms and the implementation of this approach through calculations of fission products release to the environment for a large PWR reactor under a selected set of accident conditions. The development and improvement of calculational procedures is an evolutionary process and in the long term must be verified through experimental studies. It is anticipated that as additional information is obtained the accuracy of predictions can be improved and uncertainties reduced. Transport and deposition of radionuclides were found to be quite dependent on the accident sequences and the corresponding thremal hydraulic conditions. Reduced temperatures led to increased deposition of vapor species, and reduced flow rates to increased aerosol deposition. It is to be recognized that the estimates of release fractions are subject to uncertainties in the data and computer models employed in the calculations and are expected to have been influenced by assumptions regarding plant geometry, thermal hydraulics, deposition mechanisms, and sequence events. The effects of these assumptions will be investigated as this study continues. (Author)

  14. Geochemistry of long lived transuranic actinides and fission products

    International Nuclear Information System (INIS)

    1992-01-01

    The IAEA initiated in 1987 a new Co-ordinated Research Programme (CRP) on geochemistry of long lived transuranic actinides and fission products for a duration of 5 years. The framework of the CRP consists of three main components: (1) development of a working hypothesis with focus on laboratory studies; (2) testing of the working hypothesis with the focus on the field studies; and (3) transport modelling. The contents of this document reflect the results reported on by a number of Member States who participated in this Co-ordinated Research Programme which investigated the geochemical processes and mechanisms which affect rock-water interactions and migration of the chemical elements in geological media as scientific background in support of safety assessments of repositories for high level radioactive wastes. Studies conducted considered the migration of the long lived radionuclides of Tc, I, Np and Pu in both the near and far field. The programme investigated natural occurrences and geochemical processes and mechanisms which may affect migration of the chemical elements under consideration in geological media which may be used for disposal of radioactive wastes. 47 refs, 9 figs, 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O

    1967-09-15

    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.

  16. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  17. Fission product separation from seawater by electrocoagulation method

    International Nuclear Information System (INIS)

    Kitagaki, T.; Hoshino, T.; Sambommatsu, Y.; Yano, K.; Takeuchi, M.; Igarashi, T.; Suzuki, T.

    2013-01-01

    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)

  18. Preliminary treatment of chlorinated waste streams containing fission products

    Energy Technology Data Exchange (ETDEWEB)

    Hudry, Damien; Bardez, Isabelle; Bart, Florence [CEA Marcoule DTCD/SECM/LM2C, BP 17171, 30207 Bagnols sur Ceze (France); Deniard, Philippe; Jobic, Stephane [Institut des Materiaux Jean Rouxel, Universite de Nantes, CNRS, BP 32229, 44322 Nantes cedex 3 (France); Rakhmatullin, Aydar [Conditions Extremes et Materiaux: Hautes Temperatures et Irradiations, CEMHTI-CNRS, 45071 Orleans cedex 2 (France); Bessada, Catherine [Conditions Extremes et Materiaux: Hautes Temperatures et Irradiations, CEMHTI-CNRS, 45071 Orleans cedex 2 (France); Universite d' Orleans, Faculte des Sciences, BP 6749, 45067 Orleans cedex 2 (France)

    2008-07-01

    Separating actinides from fission products (FP) by electrolytic techniques in a molten chloride medium produces high-level waste which, because of its high chlorine content, cannot be directly and quantitatively loaded in a glass matrix and therefore requires the development of new management methods. In this regard the strategy of submitting chlorinated waste streams to a preliminary treatment consists in separating the various types of FP from the solvent to minimize the ultimate high-level waste volume. Selective precipitation of the rare earth elements by NH{sub 4}H{sub 2}PO{sub 4} was investigated in a LiCl-KCl medium, and could constitute the first step in the purification process. Unlike the use of alkali orthophosphate, this method provides similar conversion factors with the simple addition of stoichiometric phosphorus (P:rare-earth = 1) and does not require excess phosphate (P:rare-earth = 5). This prevents the formation of a secondary Li{sub 3}PO{sub 4} phase. Moreover, NH{sub 4}H{sub 2}PO{sub 4} also allows chlorine bound to rare earth elements to be eliminated as NH{sub 4}Cl. The formation of HCl is highly probable.

  19. Fission product release from core-concrete mixtures

    International Nuclear Information System (INIS)

    Roche, M.F.; Settle, J.; Leibowitz, L.; Johnson, C.E.; Ritzman, R.L.

    1988-01-01

    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% H 2 -0.06% H 2 O (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

  20. Baseline Glass Development for Combined Fission Products Waste Streams

    International Nuclear Information System (INIS)

    Crum, Jarrod V.; Billings, Amanda Y.; Lang, Jesse B.; Marra, James C.; Rodriguez, Carmen P.; Ryan, Joseph V.; Vienna, John D.

    2009-01-01

    Borosilicate glass was selected as the baseline technology for immobilization of the Cs/Sr/Ba/Rb (Cs), lanthanide (Ln) and transition metal fission product (TM) waste steams as part of a cost benefit analysis study.(1) Vitrification of the combined waste streams have several advantages, minimization of the number of waste forms, a proven technology, and similarity to waste forms currently accepted for repository disposal. A joint study was undertaken by Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) to develop acceptable glasses for the combined Cs + Ln + TM waste streams (Option 1) and Cs + Ln combined waste streams (Option 2) generated by the AFCI UREX+ set of processes. This study is aimed to develop baseline glasses for both combined waste stream options and identify key waste components and their impact on waste loading. The elemental compositions of the four-corners study were used along with the available separations data to determine the effect of burnup, decay, and separations variability on estimated waste stream compositions.(2-5) Two different components/scenarios were identified that could limit waste loading of the combined Cs + LN + TM waste streams, where as the combined Cs + LN waste stream has no single component that is perceived to limit waste loading. Combined Cs + LN waste stream in a glass waste form will most likely be limited by heat due to the high activity of Cs and Sr isotopes.