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Sample records for actinides fission products

  1. Chemistry of actinides and fission products

    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

  2. Actinide and fission product separation and transmutation

    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.)

  3. Actinide and fission product separation and transmutation

    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

    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 partitioning and transmutation

    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.)

  6. Actinide and fission product partitioning and transmutation

    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.)

  7. Actinide and fission product partitioning and transmutation

    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.)

  8. Actinide and fission product separation and transmutation

    NONE

    1991-07-01

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

  9. Actinide and fission product separation and transmutation

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

  10. Status report on actinide and fission product transmutation studies

    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

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

    Kaminski, Michael D.; Mertz, Carol J.

    2016-06-07

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

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

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

  13. High flux transmutation of fission products and actinides

    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

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

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

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

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

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

    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.)

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

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

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

    Schenter, R.E.

    1978-05-01

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

  19. Recycling of actinides and fission products, the Dutch RAS research programme

    An ECN, a research programme has been started to contribute to current international research efforts in the field of P and T. The name of this programme is RAS, which is the dutch acronym for recycling of actinides and fission products. This multidisciplinary programme consists of the following components: - Nuclear data ('cross-section libraries') - Reactor physics and scenario studies - Chemical studies ('actinide chemistry') - Technological studies and irradiations. (orig./HP)

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

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

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

    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)

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

    Salahuddin Asif

    2013-01-01

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

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

    Salahuddin Asif; Iqbal Masood

    2013-01-01

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

  4. Partitioning of actinides and fission products using molten salt electrorefining process

    Electrorefining is the key step of pyrometallurgical processing for separating actinides from fission products. In this work, the electrorefining process is carried out in a electrorefining cell that contains molten salts (49% LiCl- 51% KCL) floating on a liquid cadmium. The cell is operated under an inert atmosphere at 500 degree C. In this work we describe in detail the construction of the cell and the way of operation

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

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

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

    Gay, R. L.; Grantham, L. F.; Fusselman, S. P.; Grimmett, D. L.; Roy, J. J.

    1995-09-01

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

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

    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)

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

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

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

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

    2015-06-01

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

  10. Facilities for preparing actinide or fission product-based targets

    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.

  11. Immobilization of actinides and fission products in Synroc

    Synroc containing simulated JW-A waste decreased in density as it sustained radiation damage due to doping with 0.69 wt% of 244Cm. The rate of change in density with increasing α-fluence showed an increase beyond ∼9.5 x 1017 αg-1, and the increase may reflect the onset of intergranular cracking. This may also explain the increases in leach rates of many elements above about the same α-fluence. Additional leach rate increases were deduced to occur in the perovskite phase as a result of intragranular radiation damage. Leach rates of Np, Pu, Am, and Cm were found to be much lower than the matrix elements. Nd3+ and U4+ incorporation in the very durable zirconolite phase is extensive suggesting that it is suitable for actinide encapsulation. (author)

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

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

  13. IAEA activity on partitioning and transmutation of actinides and fission products

    In 1990, the IAEA received a request from Member States to review the status of research and development on partitioning and transmutation of actinides and fission products. In response to this request the Advisory Group Meeting (AG) was held in the fall of 1991. AG advised the Agency to play an active role in coordinating international activities in this area. A series of meetings that followed identified considerable interest among many Member States and international organizations in the P and T options as a potential complement to the reference concepts of the back-end of nuclear fuel cycle. Inherent difficulties for the Agency to actively explore this programme were identified including non-proliferation concerns from some Member States about partitioning technology and possible duplication of effort in other international organizations, especially OECD/NEA. But, there remain fundamental questions to be addressed on the objectives of and motivations for P and T and it is clear that some common international understanding would be necessary. In order to contribute to the solution of this problem, and considering the existence of programmes being implemented by OECD/NEA, the Agency has initiated a new CRP entitled 'Safety, environmental and non-proliferation aspects of partitioning and transmutation of actinides and fission products' (1994-1998). This presentation will explain about this Agency's new CRP and how the Agency's work is co-ordinated with other international activities. (author)

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

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

  15. Actinide, lanthanide and fission product speciation and electrochemistry in high and low temperature ionic melts

    There is currently a great deal of research interest in the development of molten salt technology, both classical high temperature melts and low temperature ionic liquids, for the electrochemical separation of the actinides from spent nuclear fuel. We are interested in gaining a better understanding of actinide and key fission product speciation and electrochemical properties in a range of melts. Our studies in high temperature alkali metal melts (including LiCl and LiCl-KCl and CsCl-NaCl eutectics) have focussed on in-situ species of U, Th, Tc and Ru using X-ray absorption spectroscopy (XAS, both EXAFS and XANES) and electronic absorption spectroscopy (EAS). We report unusual actinide speciation in high temperature melts and an evaluation of the likelihood of Ru or Tc volatilization during plant operation. Our studies in lower temperature melts (ionic liquids) have focussed on salts containing tertiary alkyl group 15 cations and the bis(tri-fluor-methyl)sulfonyl)imide anion, melts which we have shown to have exceptionally wide electrochemical windows. We report Ln, Th, U and Np speciation (XAS, EAS and vibrational spectroscopy) and electrochemistry in these melts and relate the solution studies to crystallographic characterised benchmark species. (authors)

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

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

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

    Herrera-Martínez, Adonai

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

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

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

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

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

  20. Selective extraction of actinides and long lived fission products by functionalized calixarenes

    Selective extraction can be used for the removal of minor actinides or long lived fission products (93Zr, 99Tc,135Cs....) from Purex raffinate in order to enable their destruction by incineration or their disposal in very specific matrices. On the other hand, the selective removal of 90Sr, 137Cs and actinides allows waste to be de-categorized and to send the greatest part of the waste to the existing subsurface disposal. To achieve this aim, calix[4]arenes, macrocycles made up of phenolic units linked by methylene bridges, are used as building blocks for functionalized and pre-organized assemblies able to remove cations with a high efficiency and an important selectivity from acidic or high sodium content liquid waste. Calix[4]arene-crown-6, consisting of a calixarene frame maintained in a special conformation (1-3 alternate) by one or two polyethylene glycol bridges containing six oxygen atoms, display an exceptional efficiency for the extraction of cesium from acidic, neutral or alkaline media even in presence of large amounts of sodium. Various calix[4]arene tetra-alkyl ethers, substituted at the upper rim by CMPO-like functional group -NH-C(O)-CH2-P(O)Ph2, have been synthesized for the extraction of lanthanides and actinides. All these compounds are tremendously strong extractants compared with the commonly used CMPO. Complementary studies demonstrated the role of the pre-organization of these calixarenes: While CMPO displays a low discrimination of lanthanides, the extraction of lanthanides by calixarenes-like-CMPO strongly decreases as the atomic number of lanthanides increases. So this class of extractants makes possible a partial separation of minor actinides from lanthanides, it has to be noticed that this separation can be carried out from very acidic media. These examples demonstrate the interest of the calixarene frame for the synthesis of very specific extractants. We have to point out the promising role of the molecular modeling, that enables us, in

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

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

  2. Stainless steel-zirconium alloy waste forms for metallic fission products and actinides during treatment of spent nuclear fuel

    Stainless steel-zirconium waste form alloys are being developed for the disposal of metallic wastes recovered from spent nuclear fuel using an electrometallurgical process developed by Argonne National Laboratory. The metal waste form comprises the fuel cladding, noble metal fission products and other metallic constituents. Two nominal waste form compositions are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels. The noble metal fission products are the primary source of radiation and their contribution to the waste form radioactivity has been calculated. The disposition of actinide metals in the waste alloys is also being explored. Simulated waste form alloys were prepared to study the baseline alloy microstructures and the microstructural distribution of noble metals and actinides, and to evaluate corrosion performance

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

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

  4. Structural study and properties of peraluminous formulations for the fission products and minor actinides confinement

    In this work, peraluminous glasses (lack of alkaline and alkaline earth ions regarding aluminum) are under study to assess the potentiality of these matrices to confine fission products and minor actinides (FPA) at higher rate than current R7T7 glass (18,5 wt % FPA). The first part of this work aims at studying the physical and chemical properties of complex peraluminous glasses containing increasing FPA rate (18.5 to 32 wt %) to compare them with the specifications. The very low crystallization tendency of complex glasses containing up to 22.5 wt % as well as the very good chemical durability observed are major assets. The other part focuses on the lanthanides incorporation in simplified glass compositions in the SiO2-B2O3-Al2O3-Na2O-CaO-Ln2O3 system (Ln = Nd or La). The glass homogeneity and devitrification tendency are investigated at different scales by XRD, SEM, TEM and structural techniques such as NMR (MAS, MQMAS, REDOR, HMQC, DHMQC) and neodymium optical spectroscopy that appear very powerful to determine the lanthanides structural role regarding aluminum and describe more precisely the structural organization of peraluminous network, as still unknown in such systems. The glass homogeneity was demonstrated in a large composition domain and new structural data were put in evidence at high lanthanides content. (author)

  5. Actinide, Activation Product and Fission Product Decay Data for Reactor-based Applications

    The UK Activation Product Decay Data Library was first released in September 1977 as UK-PADD1, to be followed by regular improvements on an almost yearly basis up to the assembly of UKPADD6.12 in March 2013. Similarly, the UK Heavy Element and Actinide Decay Data Library followed in December 1981 as UKHEDD1, with the implementation of various modifications leading to UKHEDD2.6, February 2008. Both the data content and evaluation procedures are defined, and the most recent evaluations are described in terms of specific radionuclides and the resulting consistency of their recommended decay-data files. New versions of the UKPADD and UKHEDD libraries are regularly submitted to the NEA Data Bank for possible inclusion in the JEFF library

  6. Technological research on Recycling of Actinides and fission products (RAS). Irradiations in the High Flux Reactor (HFR), Petten, Netherlands

    The purpose of the title irradiations is to study the efficiency and technical feasibility of possible transmutation processes for those long-lived actinides and fission products, that contribute to long-term radiotoxicity and leaking risks of geological storage. A cooperative research program (EFFTRA or Experimental Feasibility of Targets for TRAnsmutation) has been set up for irradiations of technetium, iodine and americium in the thermal reactor HFR and the fast reactor Phenix. A radiation program for fission products is in progress in the HFR. An inert matrix concept is developed, in which the actinide is mixed with a ceramic material, which hardly reacts with neutrons and actinides and containment materials. Irradiation experiments with candidate inert matrices will be carried out in the HFR. Also, the feasibility of transmutation of americium in a thermal spectrum will be demonstrated by means of a long-range experiment in the HFR. Plans are elaborated for the irradiation of plutonium in inert matrices in the HFR to realize an efficient transmutation of existing supplies, both military and civil, of plutonium. 8 figs., 4 tabs., 18 refs

  7. Chromatographic separation of actinides and lanthanides from fission products on CMPO/SiO2-P extraction resin

    Since several years, IRI is developing an alternative for the PUREX process, based on chromatographic separations. Next to other advantages, this process shall allow partitioning of the minor actinides. At SCK x CEN, previous chromatographic experiments using LWR fuel dissolved in concentrated nitric acid, showed that Am and Cm do not adsorb onto the proposed AR-01 anion exchanger and elute simultaneously with the fission products, while U, Pu, and Np are retained quantitatively and can then be further separated and purified. After such a test, the fractions containing Am, Cm and fission products were pooled and loaded onto a CMPO/SiO2-P column. IRI had prepared this silica-based extraction resin by impregnating CMPO into a styrene-divinylbenzene copolymer, which was immobilized in porous silica particles. Earlier experiments with simulated HL W solutions in nitric acid had revealed that it allows separating the trivalent actinides (Am, Cm) and lanthanides (Ce, Nd, Eu,..) from fission products such as Cs, Sr and Ru. The present hot cell experiment using real LWR fuel confirms these results. (orig.)

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

    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.)

  9. Partition of actinides and fission products between metal and molten salt phases: Theory, measurement, and application to IFR pyroprocess development

    The chemical basis of Integral Fast Reactor fuel reprocessing (pyroprocessing) is partition of fuel, cladding, and fission product elements between molten LiCl-KCl and either a solid metal phase or a liquid cadmium phase. The partition reactions are described herein, and the thermodynamic basis for predicting distributions of actinides and fission products in the pyroprocess is discussed. The critical role of metal-phase activity coefficients, especially those of rare earth and the transuranic elements, is described. Measured separation factors, which are analogous to equilibrium constants but which involve concentrations rather than activities, are presented. The uses of thermodynamic calculations in process development are described, as are computer codes developed for calculating material flows and phase compositions in pyroprocessing

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

    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, 99Tc and 129I 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 4th 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)

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

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

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

    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

  13. Retention behavior of actinides and long lived fission products on Smectite rich clays

    In the present work, sorption of Am(llI), Cs(I) and Sr(ll) by the Smectite rich clay from western India has been studied in detail under the varying experimental conditions, viz., pH, ionic strength, and metal ion concentration. The experimental data on sorption have been modeled using the surface complexation model. Am(llI) sorption by smectite rich clay was found to increase with the pH of the suspension. At lower pH values, the sorption decreased with increasing ionic strength of the suspension, but remained constant at higher pH values. This is reminiscent of the ion exchange mechanism at lower pH and predominantly inner sphere complexation at higher pH. Surface complexation modeling using FITEQL could successfully explain these two mechanisms operating in the different pH values. Sorption of Cs(I) and Sr(II) by the smectite rich clay was studied under the varying experimental conditions. Though the sorption of both the metal ions increased with pH, it decreased with the increasing ionic strength, at all pH values, suggesting ion exchange as the predominant mechanism at all pH values. Further, the ionic strength dependence was different in the case of Cs(I) and Sr(II) depending upon the metal ion concentration. At same metal ion concentration of Cs(I) and Sr(II) (10-5 M) the extent of decrease with ionic strength was same in both cases, while at 10-9 M, Cs(I), the decrease was much smaller than that at 10-5M. This indicates the existence of ion exchange sites having different affinities. These studies have shown high retention capacity of the clay for actinides and long lived fission products with the sorption following ion exchange mechanism in the case of Cs(I) and Sr(II) and a combination of ion exchange and surface complexation in the case of Am(III) depending upon the pH. The sorption data could be successfully explained within the framework of FITEQL, taking into account both the types of binding sites

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

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

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

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

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

    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.)

  17. Migration of the fission products strontium, technetium, iodine, cesium, and the actinides neptunium, plutonium, americium in granitic rock

    Rock samples were taken from drilling cores in granitic and granodioritic rock, and small (2x2x2 cm) rock tablets from the drilling cores were exposed to a groundwater solution containing one of the studied elements at race levels. The concentration of the element versus penetration depth in the rock tablet was measured radiometrically. The sorption on the mineral faces and the migration into the rock was studied, by an autoradiographic technique. The cationic fission products strontium and cesium had apparent diffusivities of 10-13-10-14 m2/s. They migrate mainly in fissures or filled fractures containing e.g., calcite, epidote or chlorite or in veins with hgih capacity minerals (e.g. biotite). The anionic fission products iodine and technetium had apparent diffusivities of about 10-14 m2/s. These species migrate along mineral boundaries and in open fractures and to a minor extent in high capacity mineral veins. The migration of the actinides neptunium, plutonium and americium is very slow (in the mm-range after 2-3 years contact time). The apparent diffusivities were about 10-15 m2/s. The actinide migration into the rock was largely confined to fissures. (orig./HP)

  18. Transmutation of nuclear waste. State-of-the-art national and international research and strategy studies on partitioning and transmutation of actinides and fission products

    Since 1991 the Netherlands Energy Research Foundation (ECN) in Petten, Netherlands, runs a programme on recycling and transmutation of actinides and long-lived fission products that are present in the spent fuel from nuclear power generation. This programme, which is known under the Dutch acronym RAS, is concentrated on the following topics: reactor physics and scenario studies for transmutation, non-proliferation, thorium cycle, irradiations in the High Flux Reactor at Petten, chemical and material studies of fuels and targets, radiological effects and risks. In the present paper a short description of the achievements of the RAS programme is given. Next, the status of the international research on recycling of actinides and fission products is described. Strategies and (innovative) fuel cycle technology required for the recycling of plutonium, minor actinides and fission products are discussed and their possibilities and limits are identified. Also the potential of future options with low actinide production (thorium cycles, accelerators) is considered. Recommendations for future research in this field are given, taking into account the results of a review by a national committee of experts from government, science and industry. The future work should concentrate on: advanced partitioning methods for trivalent actinides, for which a break-through is required, transmutation of actinides using inert matrices as support (non-fissionable materials), studies using 100% MOX-PWRs, HWRs, HTRs and fast burners, innovative systems for future 'clean' energy production using thorium cycle and/or accelerators. It is emphasized that the radiological effects of all new concepts to be developed for recycling and transmutation should be analysed adequately. 6 figs., 14 tabs., 97 refs

  19. Comparison of Computational Estimations of Reactivity Margin From Fission Products and Minor Actinides in PWR Burnup Credit

    This paper has presented the results of a computational benchmark and independent calculations to verify the benchmark calculations for the estimation of the additional reactivity margin available from fission products and minor actinides in a PWR burnup credit storage/transport environment. The calculations were based on a generic 32 PWR-assembly cask. The differences between the independent calculations and the benchmark lie within 1% for the uniform axial burnup distribution, which is acceptable. The Δk for KENO - MCNP results are generally lower than the other Δk values, due to the fact that HELIOS performed the depletion part of the calculation for both the KENO and MCNP results. The differences between the independent calculations and the benchmark for the non-uniform axial burnup distribution were within 1.1%

  20. Selective separation of actinides and long lived fission products from aged liquid wastes produced by the EUREX plant at Saluggia

    The chemical process for the selective separation of actinides and long lived fission products from aged MTR liquid wastes is described. To perform this selective separation, some chemical procedures such as precipitation and ion exchange, both in acidic and alkaline media, have been investigated, mainly at the laboratory scale, using simulated and traced waste solutions. To confirm the results obtained with simulated or traced solutions, additional tests with true MTR wastes are in progress in the analytical hot cell of ENEA EUREX pilot plant in Saluggia. With all these results, it is possible to perform a preliminary selection of the reference process. Further larger scale information will be available after experimental runs on a cold pilot plant, named SERSE, designed and built at Saluggia for engineering scale demonstration of the chemical separation process. This R and D work was performed and partially funded in the frame of the research programmes of the European Communities. (author)

  1. Fission production and actinides in the spent graphite of the reactor stacks of the Siberian chemical integrated plant

    The peculiarity of the accomplished studies consisted in the representative selection of the reactor graphite stacks samples and in the performance of the complex analysis of their radioactive contamination. The role of incidents in forming the graphite contamination by individual radionuclides is identified and their distribution in stacks is studied. The correlation between the content of various radionuclides is investigated. The schemes for evaluating their reserve in the graphite stack are plotted. The results on evaluating the radionuclides reserve in the graphite stack highly differ from the earlier forecasted ones. The fission products and actinides reserves are by 10 times lesser as it was fore coated earlier, which may essentially simplify dismantling and selection of utilization technologies

  2. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses

  3. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  4. Non-compound nucleus fission in actinide and pre-actinide regions

    R Tripathi; S Sodaye; K Sudarshan

    2015-08-01

    In this article, some of our recent results on fission fragment/product angular distributions are discussed in the context of non-compound nucleus fission. Measurement of fission fragment angular distribution in 28Si+176Yb reaction did not show a large contribution from the non-compound nucleus fission. Data on the evaporation residue cross-sections, in addition to those on mass and angular distributions, are necessary for better understanding of the contribution from non-compound nucleus fission in the pre-actinide region. Measurement of mass-resolved angular distribution of fission products in 20Ne+232Th reaction showed an increase in angular anisotropy with decreasing asymmetry of mass division. This observation can be explained based on the contribution from pre-equilibrium fission. Results of these studies showed that the mass dependence of anisotropy may possibly be used to distinguish pre-equilibrium fission and quasifission.

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

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

  6. Development of ARTIST process, extraction and separation of actinides and fission products by TODGA

    Total recovery of TRU from nitric acid solution was studied by using TODGA whose extractant is a tridentate ligand showing the high extractability for An(III) and An(IV). TODGA extracts Pu(IV), Am(III) and Cm(III) effectively but not Np(V), therefore Np(V) has to be reduced to Np(IV) prior to extraction. From the results of the extraction of fission products, Sr(II), Zr(IV) and lanthanides(III) give high distribution ratios in TODGA-HNO3 system. After extraction of TRU with Sr(II) and Zr(IV), these FP elements can be stripped by using 0.2 M oxalic acid in 2 M HNO3 or diluted HNO3, while TRU remains in the organic phase. After separation of Sr(II) and Zr(IV), the backward extraction of total TRU into aqueous phase was also studied. (authors)

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

    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

  8. First burnup credit application including actinides and fission products for transport and storage cask by using French experiments

    The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C-E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA-AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TN's application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies. (author)

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

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

    2016-07-01

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

  10. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's Commission at Three Mile Island

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis

  11. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's commission on the accident at Three Mile Island

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis

  12. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's Commission at Three Mile Island

    England, T.R.; Wilson, W.B.

    1979-10-01

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis.

  13. Fission cross section measurements for minor actinides

    Fursov, B. [IPPE, Obninsk (Russian Federation)

    1997-03-01

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

  14. Advancement of reprocessing technology. The forefront of the actinides/fission products separation

    The subject which is important for building the future back end process of nuclear fuel is the better compatibility of the sharp rise of economic efficiency with global environmental conditions, taking up the fuel cycle system for fast reactors as the object. Wet reprocessing PUREX process is excellent in its reliability and safety, but from the viewpoint of economic efficiency and the load on waste disposal, same pointing-out has been done. In high level waste liquid, trace minor actinides and large amount of Na salt are the problems. As the advancement of PUREX process, the research on the reduction of Na waste liquid is reported. As for the recent improvement, emphasis has been placed on the control of the behavior of Np, Tc and Pt family. As the wet type actinide separation process, transuranium extraction (TRUEX) process is the relatively new, powerful solvent extraction process. Its development is described. By using the real waste liquid generated by the PUREX test of the spent fuel from fast reactors, the multi-stage, opposite flow extraction test on bench scale has been carried out at the hot cell of Chemical Processing Facility. The separation of actinides using macrocyclic compounds is reported. (K.I.)

  15. Investigation of single-cycle separation process based on forward and backward extractions of actinides and fission products

    We have been developing a new partitioning method of high-level radioactive waste by the single-cycle extraction process. This process is composed of the extraction of actinides (An) and fission products (FP, e.g., Pd, Ru, Mo and Tc), and mutual separation by reverse extraction. The extractant employed in this process is required to extract soft, hard acid metals and oxonium anions simultaneously. The NTAamide (N,N,N',N',N”,N”,-hexaoctyl-nitrilotriacetamide) is one of the candidate extractants. After the extraction of An and FP, the mutual separation by reverse-extraction should be set up. Distribution ratios of Pd and Ru, which are obtained by NTAamide extraction, can be suppressed by masking agents, thiourea, systeine, diethylenetriamine, and trisaminoethylamine. The masking of Mo can be performed using methylimino-N,N'-diethylacetamide (MIDEA), NTAamide(C2) and iminodimethylphosphoric acid, and Re can be stripped using an aqueous phase with high pH. The information on extraction and masking for these metals will be utilized in the development of the single-cycle process. (author)

  16. Isotope ratio analysis of actinides, fission products, and geolocators by high-efficiency multi-collector thermal ionization mass spectrometry

    A ThermoFisher 'Triton' multi-collector thermal ionization mass spectrometer (MC-TIMS) was evaluated for trace and ultra-trace level isotoperatioanalysis of actinides (uranium, plutonium, and americium), fission products and geolocators (strontium, cesium, and neodymium). Total efficiencies (atoms loaded to ions detected) of up to 0.5-2% for U, Pu, and Am, and 1-30% for Sr, Cs, and Nd can be reported employing resin bead load techniques onto flat ribbon Re filaments or resin beads loaded into a millimeter-sized cavity drilled into a Re rod. This results in detection limits of 4 atoms to 105 atoms) for 239-242+244Pu, 233+236U, 241-243Am, 89,90Sr, and 134,135,137Cs, and (le) 1 pg for natural Nd isotopes (limited by the chemical processing blank) using a secondary electron multiplier (SEM) or multiple-ion counters (MICs). Relative standard deviations (RSD) as small as 0.1% and abundance sensitivities of 1 x 106 or better using a SEM are reported here. Precisions of RSD ∼ 0.01-0.001% using a multi-collector Faraday cup array can be achieved at sub-nanogram concentrations for strontium and neodymium and are suitable to gain crucial geolocation information. The analytical protocols reported herein are of particular value for nuclear forensic and nuclear safeguard applications.

  17. Preparation and characterization of polyphase ceramic for fixation of actinides and fission products

    Two basic crystalline phases, a fluorite type, calcia stabilized zirconia, and a magnetoplumbite type, CaAl12O19, have been studied for incorporating the full range of waste compositions in to polyphase ceramic forms. The phase assemblage provides crystalline host phases, with stable mineral analogues, for many radionuclides in the waste. Fluorite is considered to be suitable host phase for the fixation of actinides and lanthanides. Magnetoplumbite-like structure can accommodate a wide range of elements with various charges and ionic radii. These kinds of compounds, in addition, present good chemical inertia. (author)

  18. Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

    Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fission yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice

  19. Fission of actinides using a table-top laser

    Schwoerer, H; Sauerbrey, R; Galy, J; Magill, J; Rondinella, V; Schenkel, R; Butz, T

    2003-01-01

    Powerful table-top lasers are now available in the laboratory and can be used to induce nuclear reactions. We report the first demonstration of nuclear fission using a high repetition rate table-top laser with intensities of 10 sup 2 sup 0 W/cm sup 2. Actinide photo-fission has been achieved in both sup 2 sup 3 sup 8 U and sup 2 sup 3 sup 2 Th from the high-energy Bremsstrahlung radiation produced by laser acceleration of electrons. The fission products were identified by time-resolved gamma-spectroscopy. (authors)

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

    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

  1. Ionic Liquid and Supercritical Fluid Hyphenated Techniques for Dissolution and Separation of Lanthanides, Actinides, and Fission Products

    This project is investigating techniques involving ionic liquids (IL) and supercritical (SC) fluids for dissolution and separation of lanthanides, actinides, and fission products. The research project consists of the following tasks: Study direct dissolution of lanthanide oxides, uranium dioxide and other actinide oxides in [bmin][Tf2N] with TBP(HNO3)1.8(H2O)0.6 and similar types of Lewis acid-Lewis base complexing agents; Measure distributions of dissolved metal species between the IL and the sc-CO2 phases under various temperature and pressure conditions; Investigate the chemistry of the dissolved metal species in both IL and sc-CO2 phases using spectroscopic and chemical methods; Evaluate potential applications of the new extraction techniques for nuclear waste management and for other projects. Supercritical carbon dioxide (sc-CO2) and ionic liquids are considered green solvents for chemical reactions and separations. Above the critical point, CO2 has both gas- and liquid-like properties, making it capable of penetrating small pores of solids and dissolving organic compounds in the solid matrix. One application of sc-CO2 extraction technology is nuclear waste management. Ionic liquids are low-melting salts composed of an organic cation and an anion of various forms, with unique properties making them attractive replacements for the volatile organic solvents traditionally used in liquid-liquid extraction processes. One type of room temperature ionic liquid (RTIL) based on the 1-alkyl-3-methylimidazolium cation [bmin] with bis(trifluoromethylsulfonyl)imide anion [Tf2N] is of particular interest for extraction of metal ions due to its water stability, relative low viscosity, high conductivity, and good electrochemical and thermal stability. Recent studies indicate that a coupled IL sc-CO2 extraction system can effectively transfer trivalent lanthanide and uranyl ions from nitric acid solutions. Advantages of this technique include operation at ambient temperature

  2. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions

    The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for

  3. MANTRA: An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

    Youinou, G.; Vondrasek, R.; Veselka, H.; Salvatores, M.; Paul, M.; Pardo, R.; Palmiotti, G.; Palchan, T.; Nusair, O.; Nimmagadda, J.; Nair, C.; Murray, P.; Maddock, T.; Kondrashev, S.; Kondev, F. G.; Jones, W.; Imel, G.; Glass, C.; Fonnesbeck, J.; Berg, J.; Bauder, W.

    2014-05-01

    This paper presents an update of an on-going collaborative INL-ANL-ISU integral reactor physics experiment whose objective is to infer the effective neutron capture cross sections for most of the actinides of importance for reactor physics and fuel cycle studies in both fast and epithermal spectra. Some fission products are also being considered. The principle of the experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation together with the neutron fluence will allow inference of effective neutron capture cross-sections in different neutron spectra.

  4. 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

    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

  5. Prediction of some fission properties of actinides

    The 2 Z-N correlations are indications for the deuteron-triton clusters structure to most of the nuclei. For N=Z nuclei this approach indicates deuteron clusters only. The space dependence Schroedinger equation for neutron and proton in the same shell for N=Z nuclei shows that part of the time these particles behave like single particles and part of the time as deuteron clusters. The 2 Z-N correlations are used to predict some fission properties of some actinides. (author). 13 refs., 6 Tabs

  6. 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

    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

  7. The seventh international conference on the chemistry and migration behavior of actinides and fission products in the Geosphere MIGRATION'99 abstracts

    Palmer, C

    1999-09-01

    The Migration conferences focus on recent developments in the fundamental chemistry of actinides and fission products in natural aquifer systems, their interactions and migration in the geosphere, and the processes involved in modeling their geochemical behavior. The primary mode dissemination of technical information will be early evening poster sessions designed to encourage intensive communication between the authors and participants. Daily oral sessions will be opened with invited lectures followed by contributed papers within the scope of each session. Sessions cover: (A) Chemistry of actinides and fission products in natural aquatic systems: (1) Solubilities and dissolution reactions; (2) Complexation with inorganic and organic ligands; (3) Redox reactions; (4) Colloid formation; and (5) Experimental methods. (B) Geochemical interactions and transport phenomena: (1) Diffusion and migration in geologic media; (2) Sorption/desorption phenomena; (3) Natural analog studies; (4) Effects of biological activities and organic materials; (5) Colloid transport; (6) Radionuclides in soils; and (7) Soil-remediation chemistries. (C) Data base development and modeling: (1) Data selection and evaluation; (2) Data base management; (3) Geochemical models and modeling; (4) Application of models; and (5) Validation of modeling results.

  8. The seventh international conference on the chemistry and migration behavior of actinides and fission products in the Geosphere MIGRATION'99 abstracts

    The Migration conferences focus on recent developments in the fundamental chemistry of actinides and fission products in natural aquifer systems, their interactions and migration in the geosphere, and the processes involved in modeling their geochemical behavior. The primary mode dissemination of technical information will be early evening poster sessions designed to encourage intensive communication between the authors and participants. Daily oral sessions will be opened with invited lectures followed by contributed papers within the scope of each session. Sessions cover: (A) Chemistry of actinides and fission products in natural aquatic systems: (1) Solubilities and dissolution reactions; (2) Complexation with inorganic and organic ligands; (3) Redox reactions; (4) Colloid formation; and (5) Experimental methods. (B) Geochemical interactions and transport phenomena: (1) Diffusion and migration in geologic media; (2) Sorption/desorption phenomena; (3) Natural analog studies; (4) Effects of biological activities and organic materials; (5) Colloid transport; (6) Radionuclides in soils; and (7) Soil-remediation chemistries. (C) Data base development and modeling: (1) Data selection and evaluation; (2) Data base management; (3) Geochemical models and modeling; (4) Application of models; and (5) Validation of modeling results

  9. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Mueller, Don E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowen, Douglas G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  10. Analytical control of separation tests: Development of fission product/actinide separation by extraction chromatography prior to X-ray fluorescence measurement

    Analytical control of separation tests will require determination of the neptunium, plutonium and uranium present in the column bottom solutions with a detection limit below 0.5 mg/L. The high fission product concentrations in these solutions make it unlikely that direct X-ray fluorescence measurements would be capable of reaching such a low detection limit. Chromatographic separation was therefore considered prior to the X-ray fluorescence measurement. The separation principle consists in fixing the neptunium, uranium and plutonium initially in a chromatography column during scrubbing (elution of the fission products), then eluting the neptunium, uranium and plutonium by modifying the mobile phase. The purified actinide solution leaving the column will be assayed by LXF. The objective was achieved using columns functionalized with phosphonate groups. Because of the radioactivity level involved, the method was developed in two steps. Separation was first tested in a glove box on surrogate extraction raffinate solutions consisting of nitric acid, nonradioactive isotopes of the main fission products, and uranium, plutonium and neptunium. The method was then validated in a shielded cell with an actual extraction raffinate solution. (authors)

  11. Prompt Fission Neutron Spectra of Actinides

    Capote, R; Chen, Y J; Hambsch, F J; Kornilov, N V; Lestone, J P; Litaize, O; Morillon, B; Neudecker, D; Oberstedt, S; Ohsawa, T; Smith, D. L.

    2016-01-01

    The energy spectrum of prompt neutrons emitted in fission (PFNS) plays a very important role in nuclear science and technology. A Coordinated Research Project (CRP) “Evaluation of Prompt Fission Neutron Spectra of Actinides”was established by the IAEA Nuclear Data Section in 2009, with the major goal to produce new PFNS evaluations with uncertainties for actinide nuclei. The following technical areas were addressed: (i) experiments and uncertainty quantification (UQ): New data for neutron-induced fission of 233U, 235U, 238U, and 239Pu have been measured, and older data have been compiled and reassessed. There is evidence from the experimental work of this CRP that a very small percentage of neutrons emitted in fission are actually scission neutrons; (ii) modeling: The Los Alamos model (LAM) continues to be the workhorse for PFNS evaluations. Monte Carlo models have been developed that describe the fission phenomena microscopically, but further development is needed to produce PFNS evaluations meeting the uncertainty targets; (iii) evaluation methodologies: PFNS evaluations rely on the use of the least-squares techniques for merging experimental and model data. Considerable insight was achieved on how to deal with the problem of too small uncertainties in PFNS evaluations. The importance of considering that all experimental PFNS data are “shape” data was stressed; (iv) PFNS evaluations: New evaluations, including covariance data, were generated for major actinides including 1) non-model GMA evaluations of the 235U(nth,f), 239Pu(nth,f), and 233U(nth,f) PFNS based exclusively on experimental data (0.02 ≤ E ≤ 10 MeV), which resulted in PFNS average energies E of 2.00±0.01, 2.073±0.010, and 2.030±0.013 MeV, respectively; 2) LAM evaluations of neutron-induced fission spectra on uranium and plutonium targets with improved UQ for incident energies from thermal up to 30 MeV; and 3) Point-by-Point calculations for 232Th, 234U and 237Np targets; and (v) data

  12. Applications of inductively coupled plasma-mass spectrometry to the determination of actinides and fission products in high level radioactive wastes at the Savannah River Site

    Four years of experience in applying inductively coupled plasma-mass spectrometry (ICP-MS) to the analysis of actinides and fission products in high level waste (HLW) samples at the Savannah River Site has led to the development of a number of techniques to aid in the interpretation of the mass spectral data. The goal has been to develop rapid and reliable analytical procedures that provide the necessary chemical and isotopic information to answer the process needs of the customers. Techniques that have been developed include the writing of computer software to strip the experimental data from the instrumental data files into spreadsheets or into a spectral data processing package so that the raw mass spectra can be overlain for comparison or plotted with higher output resolution. These procedures have been applied to problems ranging from the analysis of the high level waste tanks to reactor moderator water as well as environmental samples. Criticality safety analyses in some HLW waste treatment processes depend upon actinide concentration and isotopic information generated by ICP-MS, particularly in tanks with high concentrations of 137Cs and 90Sr. Experimental results for a number of these applications will be presented. These procedures represent a considerable saving in time and expense as compared to conventional chemical separation followed by radiochemical analyses, as well as decreased radiation exposure for the analysts

  13. Separation of actinides and fission products using solvent extraction, extraction chromatography, supported liquid membrane and biosorption techniques

    The actinide elements play an important role in the development of nuclear energy. The elements, 90Th to 103Lw, occupy a unique position in the periodic table. Seaborg has classified these elements as f-group elements(l) similar to the lanthanides where electrons get filled in the 4f shell. In the case of actinides, electrons get filled in the 5f shell. Some of the properties of the 4f- and 5f-elements are quite similar, such as complex formation, lanthanide/actinide contraction, multiple valency etc. Hence specially the chemistry of actinides has aroused much interest among the scientists, to carry out different types of investigations. (author)

  14. Effect of irradiation on some actinide and fission product ions' extraction using several tetraalkyl diglycolamides

    The radiolytic stability of different substituted diglycolamides such as N,N,N',N'-tetrapentyl diglycolamides(TPDGA), N,N,N',N'-tetrahexyl diglycolamide (THDGA), N,N,N',N'-tetraoctyl diglycolamide (TODGA), N,N,N',N'-tetradecyl diglycolamide (TDDGA) and N,N,N',N'-tetra-2-ethyl hexyl diglycolamide (T2EHDGA) was investigated in solvent systems containing 30% iso-decanol as a phase modifier in n-dodecane kept in contact with a 3 M HNO3 aqueous solution while irradiating in a gamma ray chamber up to 1000 kGy. The degradation of the solvent systems was qualitatively ascertained from measuring the distribution ratio values at 3 M and 0.2 M HNO3 which gave a direct indication of the reusability of the solvent for long term reuse in separation processes such as 'actinide partitioning'. The effect of irradiation on distribution values of Sr, Pu, U was also investigated with diglycolamide extractants both under extraction and stripping conditions. Stoichiometry of the extracted species was determined for Am(III) extraction using the fresh as well as the irradiated solvent systems involving all the five substituted diglycolamides at 3 M HNO3 and the results indicated only marginal changes. GC-MS analysis was done for fresh and irradiated solvent systems of all the diglycolamides and attempts were made to identify the degradation products.

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

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

  16. Systematic study of actinide and pre-actinide fission modes

    Andrade-II, E; Deppman, A; Bernal-Castillo, J L; Balabekyan, A R; Demekhina, N A; Adam, J; Garcia, F; Guzmán, F

    2016-01-01

    In this work, we present new experimental data on mass distribution of fission fragments from $^{241}$Am proton-induced fission at $660$ MeV measured at the LNR Phasotron (JINR). The systematic analysis of several measured fragment mass distributions from different fission reactions available in the literature is also presented. The proton-induced fission of $^{241}$Am, $^{237}$Np and $^{238}$U at 26.5, 62.9 and 660 MeV was studied. The proton-induced fission of $^{232}$Th was studied at 26.5, 62.9 and 190 MeV. The fission of $^{208}$Pb also by a proton was investigated at 190, 500 and 1000 MeV. The fission of $^{197}$Au was studied for 190 and 800 MeV protons. Bremsstrahlung reactions with maximum photon energies of 50 and 3500 MeV were studied for $^{232}$Th and $^{238}$U. The framework of the Random Neck Rupture Model was applied in the analysis. The roles of the neutron excess and of the so called fissility parameter were also investigated.

  17. Measurements of Fission Cross Sections of Actinides

    Wiescher, M; Cox, J; Dahlfors, M

    2002-01-01

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

  18. Microscopic Description of Nuclear Fission: Fission Barrier Heights of Even-Even Actinides

    McDonnell, J; Schunck, N; Nazarewicz, W.

    2013-01-01

    We evaluate the performance of modern nuclear energy density functionals for predicting inner and outer fission barrier heights and energies of fission isomers of even-even actinides. For isomer energies and outer barrier heights, we find that the self-consistent theory at the HFB level is capable of providing quantitative agreement with empirical data.

  19. An independent method for data selection of long-life radionuclides (actinides and fission products) in the geosphere

    An independent method for data selection of long-life radio-nuclides based on the electronegativity equalization principle is proposed to predict the speciation of metal cations as a function of the solution pH. Hydrolysis, condensation and complexation reactions of metal cations in aqueous media are, by this simple model, unified and can be analyzed in terms of electronegativities, oxidation states and coordination numbers with a specific PC software. This paper describes the thermodynamical basis and the underlying concepts of the model in relation to aqueous actinide chemistry of elements such as U and Tc. It is then shown that the model could provide a complementary approach to existing softwares based on thermodynamic data bases allowing to make intelligent and reasonnable choices for the various complexes to consider in complex geochemical codes. (orig.)

  20. The state of the art of partitioning technology for long-lived actinides and fission products by solvent extraction method

    Ozawa, M.; Koma, Y.; Nomura, K.; Sano, Y. [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

    1998-04-01

    Japan launched an ambitious long-term program on partitioning and transmutation (P-T) called OMEGA in 1988. Under the program PNC has being carried out its R and D activities. A check and review process based on progress made was conducted in fall 1998 by STA (Science and Technology Agency). This report was prepared to submit the state of R and D activities on partitioning by solvent extraction program in PNC for seven years (1990-1997) to STA. The paper described the progress, the results and future plans on (a) improved PUREX process for the extraction of Np with Pu by valence control, (b) improved TRUEX process for the extraction of minor Actinides and (c) other potential solvents for the extraction of other long-lived FPs from spent fuels. (H. Itami)

  1. Status of measurements of fission neutron spectra of Minor Actinides

    Drapchinsky, L.; Shiryaev, B. [V.G. Khlopin Radium Inst., Saint Petersburg (Russian Federation)

    1997-03-01

    The report considers experimental and theoretical works on studying the energy spectra of prompt neutrons emitted in spontaneous fission and neutron induced fission of Minor Actinides. It is noted that neutron spectra investigations were done for only a small number of such nuclei, most measurements, except those of Cf-252, having been carried out long ago by obsolete methods and imperfectapparatus. The works have no detailed description of experiments, analysis of errors, detailed numerical information about results of experiments. A conclusion is made that the available data do not come up to modern requirements. It is necessary to make new measurements of fission prompt neutron spectra of transuranium nuclides important for the objectives of working out a conception of minor actinides transmutation by means of special reactors. (author)

  2. Production Pathways and Separation Procedures for High-Diagnostic-Value Activation Species, Fission Products, and Actinides Required for Preparation of Realistic Synthetic Post-Detonation Nuclear Debris: Status Report and FY16 Project Plan

    The objective of this project is to provide a comprehensive study on the production routes and chemical separation requirements for activation products, fission products, and actinides required for the creation of realistic post-detonation surrogate debris. Isotopes that have been prioritized by debris diagnosticians will be examined for their ability to be produced at existing irradiation sources, production rates, and availability of target materials, and chemical separation procedures required to rapidly remove the products from the bulk target matrix for subsequent addition into synthetic debris samples. The characteristics and implications of the irradiation facilities on the isotopes of interest will be addressed in addition to a summary of the isotopes that are already regularly produced. This is a planning document only.

  3. Production Pathways and Separation Procedures for High-Diagnostic-Value Activation Species, Fission Products, and Actinides Required for Preparation of Realistic Synthetic Post-Detonation Nuclear Debris: Status Report and FY16 Project Plan

    Faye, S. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shaughnessy, D. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-08-19

    The objective of this project is to provide a comprehensive study on the production routes and chemical separation requirements for activation products, fission products, and actinides required for the creation of realistic post-detonation surrogate debris. Isotopes that have been prioritized by debris diagnosticians will be examined for their ability to be produced at existing irradiation sources, production rates, and availability of target materials, and chemical separation procedures required to rapidly remove the products from the bulk target matrix for subsequent addition into synthetic debris samples. The characteristics and implications of the irradiation facilities on the isotopes of interest will be addressed in addition to a summary of the isotopes that are already regularly produced. This is a planning document only.

  4. Fusion-Fission Burner for Transuranic Actinides

    Choi, Chan

    2013-10-01

    The 14-MeV DT fusion neutron spectrum from mirror confinement fusion can provide a unique capability to transmute the transuranic isotopes from light water reactors (LWR). The transuranic (TRU) actinides, high-level radioactive wastes, from spent LWR fuel pose serious worldwide problem with long-term decay heat and radiotoxicity. However, ``transmuted'' TRU actinides can not only reduce the inventory of the TRU in the spent fuel repository but also generate additional energy. Typical commercial LWR fuel assemblies for BWR (boiling water reactor) and PWR (pressurized water reactor) measure its assembly lengths with 4.470 m and 4.059 m, respectively, while its corresponding fuel rod lengths are 4.064 m and 3.851 m. Mirror-based fusion reactor has inherently simple geometry for transmutation blanket with steady-state reactor operation. Recent development of gas-dynamic mirror configuration has additional attractive feature with reduced size in central plasma chamber, thus providing a unique capability for incorporating the spent fuel assemblies into transmutation blanket designs. The system parameters for the gas-dynamic mirror-based hybrid burner will be discussed.

  5. Fission product margin in burnup credit analyses

    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

  6. Dependence of Fission-Fragment Properties On Excitation Energy For Neutron-Rich Actinides

    Ramos, D.; Rodríguez-Tajes, C.; Caamaño, M.; Farget, F.; Audouin, L.; Benlliure, J.; Casarejos, E.; Clement, E.; Cortina, D.; Delaune, O.; Derkx, X.; Dijon, A.; Doré, D.; Fernández-Domínguez, B.; de France, G.; Heinz, A.; Jacquot, B.; Navin, A.; Paradela, C.; Rejmund, M.; Roger, T.; Salsac, M. D.; Schmitt, C.

    2016-03-01

    Experimental access to full isotopic fragment distributions is very important to determine the features of the fission process. However, the isotopic identification of fission fragments has been, in the past, partial and scarce. A solution based on the use of inverse kinematics to study transfer-induced fission of exotic actinides was carried out at GANIL, resulting in the first experiment accessing the full identification of a collection of fissioning systems and their corresponding fission fragment distribution. In these experiments, a 238U beam at 6.14 AMeV impinged on a carbon target to produce fissioning systems from U to Am by transfer reactions, and Cf by fusion reactions. Isotopic fission yields of 250Cf, 244Cm, 240Pu, 239Np and 238U are presented in this work. With this information, the average number of neutrons as a function of the atomic number of the fragments is calculated, which reflects the impact of nuclear structure around Z=50, N=80 on the production of fission fragments. The characteristics of the Super Long, Standard I, Standard II, and Standard III fission channels were extracted from fits of the fragment yields for different ranges of excitation energy. The position and contribution of the fission channels as function of excitation energy are presented.

  7. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  8. MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment of advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(Ei), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after

  9. Fission of actinides through quasimolecular shapes

    Royer, Guy; Zhang, Hongfei; Eudes, Philippe; Moustabchir, Rachid; Moreau, Damien; Jaffré, Muriel; Morabit, Youssef; Particelli, Benjamin

    2013-12-01

    The potential energy of heavy nuclei has been calculated in the quasimolecular shape path from a generalized liquid drop model including the proximity energy, the charge and mass asymmetries and the microscopic corrections. The potential barriers are multiple-humped. The second maximum is the saddle-point. It corresponds to the transition from compact one-body shapes with a deep neck to two touching ellipsoids. The scission point lies at the end of an energy plateau well below the saddle-point and where the effects of the nuclear attractive forces between two separated fragments vanish. The energy on this plateau is the sum of the kinetic and excitation energies of the fragments. The shell and pairing corrections play an essential role to select the most probable fission path. The potential barrier heights agree with the experimental data and the theoretical half-lives follow the trend of the experimental values. A third peak and a shallow third minimum appear in asymmetric decay paths when one fragment is close to a double magic quasi-spherical nucleus, while the smaller one changes from oblate to prolate shapes.

  10. Fission of actinides through quasimolecular shapes

    Royer Guy

    2013-12-01

    Full Text Available The potential energy of heavy nuclei has been calculated in the quasimolecular shape path from a generalized liquid drop model including the proximity energy, the charge and mass asymmetries and the microscopic corrections. The potential barriers are multiple-humped. The second maximum is the saddle-point. It corresponds to the transition from compact one-body shapes with a deep neck to two touching ellipsoids. The scission point lies at the end of an energy plateau well below the saddle-point and where the effects of the nuclear attractive forces between two separated fragments vanish. The energy on this plateau is the sum of the kinetic and excitation energies of the fragments. The shell and pairing corrections play an essential role to select the most probable fission path. The potential barrier heights agree with the experimental data and the theoretical half-lives follow the trend of the experimental values. A third peak and a shallow third minimum appear in asymmetric decay paths when one fragment is close to a double magic quasi-spherical nucleus, while the smaller one changes from oblate to prolate shapes.

  11. The contrasting fission potential-energy structure of actinides and mercury isotopes

    Ichikawa, Takatoshi; Iwamoto, Akira; Möller, Peter; Sierk, Arnold J.

    2012-01-01

    Fission-fragment mass distributions are asymmetric in fission of typical actinide nuclei for nucleon number $A$ in the range $228 \\lnsim A \\lnsim 258$ and proton number $Z$ in the range $90\\lnsim Z \\lnsim 100$. For somewhat lighter systems it has been observed that fission mass distributions are usually symmetric. However, a recent experiment showed that fission of $^{180}$Hg following electron capture on $^{180}$Tl is asymmetric. We calculate potential-energy surfaces for a typical actinide ...

  12. Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS

    Perkasa, Y. S. [Department of Physics, Sunan Gunung Djati State Islamic University Bandung, Jl. A.H Nasution No. 105 Cibiru, Bandung (Indonesia); Waris, A., E-mail: awaris@fi.itb.ac.id; Kurniadi, R., E-mail: awaris@fi.itb.ac.id; Su' ud, Z., E-mail: awaris@fi.itb.ac.id [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa No. 10 Bandung 40132 (Indonesia)

    2014-09-30

    Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS have been conducted. In this work, fission cross section resulted from MCNP6 prediction will be compared with result from TALYS calculation. MCNP6 with its event generator CEM03.03 and LAQGSM03.03 have been validated and verified for several intermediate and heavy nuclides fission reaction data and also has a good agreement with experimental data for fission reaction that induced by photons, pions, and nucleons at energy from several ten of MeV to about 1 TeV. The calculation that induced within TALYS will be focused mainly to several hundred MeV for actinide and sub-actinide nuclides and will be compared with MCNP6 code and several experimental data from other evaluator.

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

    Nakagawa, T

    2003-01-01

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

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

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

  15. Review of fission product yields and delayed neutron data for the actinides NP-237, PU-242, AM-242M, AM-243, CM-243 and CM-245

    A review of fission product yields and delayed neutron data for Np-237, Pu-242, Am-242m, Am-243, Cm-243 and Cm-245 has been undertaken. Gaps in understanding and inconsistencies in existing data were identified and priority areas for further experimental, theoretical and evaluation investigation detailed

  16. Fission product yields

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

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

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

  18. Impact of fuel chemistry on fission product behaviour

    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

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

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

  20. Fission product detection

    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

  1. Dependence of Fission-Fragment Properties On Excitation Energy For Neutron-Rich Actinides

    Ramos D; Rodríguez-Tajes C.; Caamaño M.; Farget F.; Audouin L.; Benlliure J.; Casarejos E.; Clement E.; Cortina D.; Delaune O.; Derkx X.; Dijon A.; Doré D.; Fernández-Domínguez B.; France G. de

    2015-01-01

    Experimental access to full isotopic fragment distributions is very important to determine the features of the fission process. However, the isotopic identification of fission fragments has been, in the past, partial and scarce. A solution based on the use of inverse kinematics to study transfer-induced fission of exotic actinides was carried out at GANIL, resulting in the first experiment accessing the full identification of a collection of fissioning systems and their corresponding fission ...

  2. Spectroscopy of neutron rich nuclei using cold neutron induced fission of actinide targets at the ILL: The EXILL campaign

    Blanc A.

    2013-12-01

    Full Text Available One way to explore exotic nuclei is to study their structure by performing γ-ray spectroscopy. At the ILL, we exploit a high neutron flux reactor to induce the cold fission of actinide targets. In this process, fission products that cannot be accessed using standard spontaneous fission sources are produced with a yield allowing their detailed study using high resolution γ-ray spectroscopy. This is what was pursued at the ILL with the EXILL (for EXOGAM at the ILL campaign. In the present work, the EXILL setup and performance will be presented.

  3. Review of the fission decay of the giant resonances in the actinide region

    The fission decay of giant resonances in the actinide region is reviewed. Results from various experiments which are invariably conflicting are discussed. These include inclusive electron and positron-induced fission, as well as experiments in which fission fragments were detected in coincidence with inelastically scattered electrons or hadrons. Attention is focussed on a recent (α,α'f) experiment in which the fission decay of the giant monopole inelastically scattered α-particles at and around 00. 49 references

  4. Actinide production in 136Xe bombardments of 249Cf

    The production cross sections for the actinide products from 136Xe bombardments of 249Cf at energies 1.02, 1.09, and 1.16 times the Coulomb barrier were determined. Fractions of the individual actinide elements were chemically separated from recoil catcher foils. The production cross sections of the actinide products were determined by measuring the radiations emitted from the nuclides within the chemical fractions. The chemical separation techniques used in this work are described in detail, and a description of the data analysis procedure is included. The actinide production cross section distributions from these 136Xe + 249Cf bombardments are compared with the production cross section distributions from other heavy ion bombardments of actinide targets, with emphasis on the comparison with the 136Xe + 248Cm reaction. A technique for modeling the final actinide cross section distributions has been developed and is presented. In this model, the initial (before deexcitation) cross section distribution with respect to the separation energy of a dinuclear complex and with respect to the Z of the target-like fragment is given by an empirical procedure. It is then assumed that the N/Z equilibration in the dinuclear complex occurs by the transfer of neutrons between the two participants in the dinuclear complex. The neutrons and the excitation energy are statistically distributed between the two fragments using a simple Fermi gas level density formalism. The resulting target-like fragment initial cross section distribution with respect to Z, N, and excitation energy is then allowed to deexcite by emission of neutrons in competition with fission. The result is a final cross section distribution with respect to Z and N for the actinide products. 68 refs., 33 figs., 6 tabs

  5. Fission product data library

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

  6. Minor actinide fission induced by multi-nucleon transfer reaction in inverse kinematics

    Taieb J.

    2010-03-01

    Full Text Available In the framework of nuclear waste incineration and design of new generation nuclear reactors, experimental data on fission probabilities and on fission fragment yields of minor actinides are crucial to design prototypes. Transfer-induced fission has proven to be an efficient method to study fission probabilities of actinides which cannot be investigated with standard techniques due to their high radioactivity. We report on the preliminary results of an experiment performed at GANIL that investigates fission probabilities with multi-nucleon transfer reactions in inverse kinematics between a 238U beam on a 12C target. Actinides from U to Cm were produced with an excitation energy range from 0 to 30 MeV. In addition, inverse kinematics allowed to characterize the fission fragments in mass and charge. A key point of the analysis resides in the identification of the actinides produced in the different transfer channels. The new annular telescope SPIDER was used to tag the target-like recoil nucleus of the transfer reaction and to determine the excitation energy of the actinide. The fission probability for each transfer channel is accessible and the preliminary results for 238U are promising.

  7. Current position on fission product behavior

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

  8. Fission product revaporization

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

  9. Summary Report of Second Research Coordination Meeting on Prompt Fission Neutron Spectra of Major Actinides

    A summary is given of the Second Research Coordination Meeting on Prompt Fission Neutron Spectra of Actinides. Experimental data and modelling methods on prompt fission neutron spectra were reviewed. Extensive technical discussions held on theoretical methods to calculate prompt fission spectra. Detailed coordinated research proposals have been agreed. Summary reports of selected technical presentations at the meeting are given. The resulting work plan of the Coordinated Research Programme is summarized, along with actions and deadlines. (author)

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

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

  11. Measurement of fission cross-section of actinides at n_TOF for advanced nuclear reactors

    Calviani, Marco; Montagnoli, G; Mastinu, P

    2009-01-01

    The subject of this thesis is the determination of high accuracy neutron-induced fission cross-sections of various isotopes - all of which radioactive - of interest for emerging nuclear technologies. The measurements had been performed at the CERN neutron time-of-flight facility n TOF. In particular, in this work, fission cross-sections on 233U, the main fissile isotope of the Th/U fuel cycle, and on the minor actinides 241Am, 243Am and 245Cm have been analyzed. Data on these isotopes are requested for the feasibility study of innovative nuclear systems (ADS and Generation IV reactors) currently being considered for energy production and radioactive waste transmutation. The measurements have been performed with a high performance Fast Ionization Chamber (FIC), in conjunction with an innovative data acquisition system based on Flash-ADCs. The first step in the analysis has been the reconstruction of the digitized signals, in order to extract the information required for the discrimination between fission fragm...

  12. Consultants' meeting on prompt fission neutron spectra of major actinides. Summary report

    A Consultants' Meeting on 'Prompt Fission Neutron Spectra of Major Actinides' was held at IAEA Headquarters, Vienna, Austria, to discuss the adequacy and quality of the recommended prompt fission neutron spectra to be found in existing nuclear data applications libraries. These prompt fission neutron spectra were judged to be inadequate, and this problem has proved difficult to resolve by means of theoretical modelling. Major adjustments may be required to ensure the validity of such important data. There is a strong requirement for an international effort to explore and resolve these difficulties and recommend prompt fission neutron spectra and uncertainty covariance matrices for the actinides over the neutron energy range from thermal to 20 MeV. Participants also stressed that there would be a strong need for validation of the resulting data against integral critical assembly and dosimetry data. (author)

  13. 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

    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

  14. The contrasting fission potential-energy structure of actinides and mercury isotopes

    Ichikawa, Takatoshi; Möller, Peter; Sierk, Arnold J

    2012-01-01

    Fission-fragment mass distributions are asymmetric in fission of typical actinide nuclei for nucleon number $A$ in the range $228 \\lnsim A \\lnsim 258$ and proton number $Z$ in the range $90\\lnsim Z \\lnsim 100$. For somewhat lighter systems it has been observed that fission mass distributions are usually symmetric. However, a recent experiment showed that fission of $^{180}$Hg following electron capture on $^{180}$Tl is asymmetric. An earlier experiment has shown fission of $^{198}$Hg and nearby nuclei is symmetric, but with hints of asymmetric yield distributions up to about 10 MeV above the saddle-point energy. We calculate potential-energy surfaces for a typical actinide nucleus and for 12 even isotopes in the range $^{178}$Hg--$^{200}$Hg, demonstrating the radical differences between actinide and mercury potential surfaces. We discuss these differences and how the changing potential-energy structure along the mercury isotope chain affects the observed (a)symmetry of the fission fragments. We show that the ...

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

    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 feed 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 ORIGEN 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

  16. Summary Report of First Research Coordination Meeting on Prompt Fission Neutron Spectra of Major Actinides

    A summary is given of the First Research Coordination Meeting on Prompt Fission Neutron Spectra of Actinides. Experimental data and modelling methods on prompt fission neutron spectra were reviewed. The programme to compile and evaluate prompt fission spectra including uncertainty information over the neutron energy range from thermal to 20 MeV was proposed. Validation of the resulting data against integral critical assembly and dosimetry data is foreseen. Detailed coordinated research proposals have been agreed. Summary reports of technical presentations at the meeting are given. The resulting work plan of the Coordinated Research Programme is summarized, along with actions and deadlines. (author)

  17. Improved Fission Neutron Data Base for Active Interrogation of Actinides

    Pozzi, Sara; Czirr, J. Bart; Haight, Robert; Kovash, Michael; Tsvetkov, Pavel

    2013-11-06

    This project will develop an innovative neutron detection system for active interrogation measurements. Many active interrogation methods to detect fissionable material are based on the detection of neutrons from fission induced by fast neutrons or high-energy gamma rays. The energy spectrum of the fission neutrons provides data to identify the fissionable isotopes and materials such as shielding between the fissionable material and the detector. The proposed path for the project is as follows. First, the team will develop new neutron detection systems and algorithms by Monte Carlo simulations and bench-top experiments. Next, They will characterize and calibrate detection systems both with monoenergetic and white neutron sources. Finally, high-fidelity measurements of neutron emission from fissions induced by fast neutrons will be performed. Several existing fission chambers containing U-235, Pu-239, U-238, or Th-232 will be used to measure the neutron-induced fission neutron emission spectra. The challenge for making confident measurements is the detection of neutrons in the energy ranges of 0.01 – 1 MeV and above 8 MeV, regions where the basic data on the neutron energy spectrum emitted from fission is least well known. In addition, improvements in the specificity of neutron detectors are required throughout the complete energy range: they must be able to clearly distinguish neutrons from other radiations, in particular gamma rays and cosmic rays. The team believes that all of these challenges can be addressed successfully with emerging technologies under development by this collaboration. In particular, the collaboration will address the area of fission neutron emission spectra for isotopes of interest in the advanced fuel cycle initiative (AFCI).

  18. Fission Product Inventory in CANDU Fuel

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

  19. Progress in fission product nuclear data

    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. Study of actinides fission induced by multi-nucleon transfer reactions in inverse kinematics

    The study of actinide fission encounters two major issues. On one hand, measurements of the fission fragment distributions and the fission probabilities allow a better understanding of the fission process itself and the discrimination among the models of nuclear structure and dynamics. On the other hand, new measurements are required to improve nuclear data bases, which are a key component for the design of new generation reactors and radio-toxic waste incinerators. This thesis is in line with different French and American experimental projects using the surrogate method, i.e. transfer reactions leading to the same compound nuclei as in neutron irradiation, allowing the study of fission of actinides which are inaccessible by conventional techniques, whereas they are important for applications. The experiment is based on multi-nucleon transfer reactions between a 238U beam and a 12C target, using the inverse kinematics technique to measure, for each transfer channel, the complete isotopic distributions of the fission fragments with the VAMOS spectrometer. The work presented in this dissertation is focused on the identification of the transfer channels and their properties, as their angular distributions and the distributions of the associated excitation energy, using the SPIDER telescope to identify the target recoil nuclei. This work of an exploratory nature aims to generalize the surrogate method to heavy transfers and to measure, for the first time, the fission probabilities in inverse kinematics. The obtained results are compared with available direct kinematics and neutron irradiation measurements. (author)

  1. Odd–even effect in fragment angular momentum in low-energy fission of actinides

    B S Tomar; R Tripathi; A Goswami

    2007-01-01

    Quantitative explanation for the odd–even effect on fragment angular momenta in the low-energy fission of actinides have been provided by taking into account the single particle spin of the odd proton at the fragment's scission point deformation in the case of odd- fragments along with the contribution from the population of angular momentum bearing collective vibrations of the fissioning nucleus at scission point. The calculated fragment angular momenta have been found to be in very good agreement with the experimental data for fragments in the mass number region of 130–140. The odd–even effect observed in the fragment angular momenta in the low-energy fission of actinides has been explained quantitatively for the first time.

  2. Angular distribution in the neutron-induced fission of actinides

    Leong L.S.

    2013-12-01

    Full Text Available Above 1 MeV of incident neutron energy the fission fragment angular distribution (FFAD has generally a strong anisotropic behavior due to the combination of the incident orbital momentum and the intrinsic spin of the fissioning nucleus. This effect has to be taken into account for the efficiency estimation of devices used for fission cross section measurements. In addition it bears information on the spin deposition mechanism and on the structure of transitional states. We designed and constructed a detection device, based on Parallel Plate Avalanche Counters (PPAC, for measuring the fission fragment angular distributions of several isotopes, in particular 232Th. The measurement has been performed at n_TOF at CERN taking advantage of the very broad energy spectrum of the neutron beam. Fission events were recognized by back to back detection in coincidence in two position-sensitive detectors surrounding the targets. The detection efficiency, depending mostly on the stopping of fission fragments in backings and electrodes, has been computed with a Geant4 simulation and validated by the comparison to the measured case of 235U below 3 keV where the emission is isotropic. In the case of 232Th, the result is in good agreement with previous data below 10 MeV, with a good reproduction of the structures associated to vibrational states and the opening of second chance fission. In the 14 MeV region our data are much more accurate than previous ones which are broadly scattered.

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

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

  4. Fission product solvent extraction

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

  5. Measurement of fast neutron induced fission cross section of minor-actinide

    In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am, Cm). Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA is measured using Dynamitron Accelerator in Tohoku University. The experimental method and the samples, which were developed or introduced during the last year, were improved in this fiscal year: (1) Development of a sealed fission chamber, (2) Intensification of Li neutron target, (3) Improvement of time-resolution of Time-of-Flight (TOF) electronic circuit, (4) Introduction of Np237 samples with large sample mass and (5) Introduction of a U235 sample with high purity. Using these improved tools and samples, the fission cross section ratio of Np237 relative to U235 was measured between 5 to 100 keV, and the fission cross section of Np237 was deduced. On the other hand, samples of Am241 and Am243 were obtained from Japan Atomic Energy Research Institute (JAERI) after investigating fission cross section of two americium isotopes (Am241 and Am 243) which are important for core physics calculation of fast reactors. (author)

  6. Chemical Production using Fission Fragments

    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)

  7. Decontamination of alkaline solution from technetium and other fission products and from some actinides by reductive coprecipitation and sorption on metals

    Peretrukhin, V.F.; Silin, V.I.; Tananaev, I.G.; Kareta, A.V.; Trushina, V.E. [Russian Academy of Sciences, Moscow (Russian Federation). Institute of Physical Chemistry

    1997-09-01

    Effective decontamination of alkaline solutions and Hanford Site tank waste simulants from technetium has been accomplished by reductive coprecipitation with iron(III) hydroxide. Addition of 1 M (NH{sub 4}){sub 2}Fe(SO{sub 4}){sub 2} to 0.5 to 4.0 M NaOH to a final concentration of 0.1 to 0.15 M coprecipitates more than 99% of the technetium. from 0.5 to 1.0 M NaOH and 98 to 96% from 2.0 to 4.0 M NaOH. Similar results were obtained by reduction of Tc(VII) with 0.1 to 0.15 M hydrazine and subsequent addition of FeCl{sub 3} to a final concentration of 0.15 M. Inclusion of four complex-forming agents [0.01 M phosphate, 0.1 M EDTA (ethylenediaminetetraacetate), 0.03 M citrate, and 0.1 M glycolate (HOCH{sub 2}CO{sub 2}{sup -})] to the alkaline solution decreases technetium coprecipitation with iron hydroxide to 85% under otherwise similar conditions. Inclusion of 0.04 M Na{sub 2}CrO{sub 4} drastically decreases reductive coprecipitation of Tc(VII) in 0.5 to 4.0 M NaOH. Iron(II) salt, added to a 0.07 M excess over that of chromate, completely reduces chromate and provides greater than 99% coprecipitation of technetium with product iron(III) and chromium(III) hydroxides. Technetium(VII) reduction by hydrazine is slow in the presence of chromate in alkaline solution, and technetium coprecipitation is incomplete in these conditions. Decontamination of an alkaline Hanford Site tank waste simulant, containing 0.04M chromate and eleven salts and complex-forming agents, by adding 1 M iron(II) salt solution was studied. Coprecipitation of 15 to 28% of the technetium and more than 99% of the plutonium occurred in the Fe/Cr(III) hydroxide precipitate produced by adding 0.05 to 0.10 M iron(II). Chromate reduction was incomplete. About 75% of the technetium was coprecipitated, and the chromate was completely reduced, after adding 0.2 M iron(II) salt.

  8. Actinide production from xenon bombardments of curium-248

    Production cross sections for many actinide nuclides formed in the reaction of 129Xe and 132Xe with 248Cm at bombarding energies slightly above the coulomb barrier were determined using radiochemical techniques to isolate these products. These results are compared with cross sections from a 136Xe + 248Cm reaction at a similar energy. When compared to the reaction with 136Xe, the maxima in the production cross section distributions from the more neutron deficient projectiles are shifted to smaller mass numbers, and the total cross section increases for the production of elements with atomic numbers greater than that of the target, and decreases for lighter elements. These results can be explained by use of a potential energy surface (PES) which illustrates the effect of the available energy on the transfer of nucleons and describes the evolution of the di-nuclear complex, an essential feature of deep-inelastic reactions (DIR), during the interaction. The other principal reaction mechanism is the quasi-elastic transfer (QE). Analysis of data from a similar set of reactions, 129Xe, 132Xe, and 136Xe with 197Au, aids in explaining the features of the Xe + Cm product distributions, which are additionally affected by the depletion of actinide product yields due to deexcitation by fission. The PES is shown to be a useful tool to predict the general features of product distributions from heavy ion reactions

  9. Interplay of fission modes in mass distribution of light actinide nuclei 225,227Pa

    R. Dubey

    2016-01-01

    Full Text Available Fission-fragment mass distributions were measured for 225,227Pa nuclei formed in fusion reactions of 19F+206,208Pb around fusion barrier energies. Mass-angle correlations do not indicate any quasi-fission like events in this bombarding energy range. Mass distributions were fitted by Gaussian distribution and mass variance extracted. At below-barrier energies, the mass variance was found to increase with decrease in energy for both nuclei. Results from present work were compared with existing data for induced fission of 224,226Th and 228U around barrier energies. Enhancement in mass variance of 225,227Pa nuclei at below-barrier energies shows evidence for presence of asymmetric fission events mixed with symmetric fission events. This is in agreement with the results of mass distributions of nearby nuclei 224,226Th and 228U where two-mode fission process was observed. Two-mode feature of fission arises due to the shell effects changing the landscape of the potential-energy surfaces at low excitation energies. The excitation-energy dependence of the mass variance gives strong evidence for survival of microscopic shell effects in fission of light actinide nuclei 225,227Pa with initial excitation energy ∼30–50 MeV.

  10. Interplay of fission modes in mass distribution of light actinide nuclei 225,227Pa

    Dubey, R.; Sugathan, P.; Jhingan, A.; Kaur, Gurpreet; Mukul, Ish; Mohanto, G.; Siwal, D.; Saneesh, N.; Banerjee, T.; Thakur, Meenu; Mahajan, Ruchi; Kumar, N.; Chatterjee, M. B.

    2016-01-01

    Fission-fragment mass distributions were measured for 225,227Pa nuclei formed in fusion reactions of 19F + 206,208Pb around fusion barrier energies. Mass-angle correlations do not indicate any quasi-fission like events in this bombarding energy range. Mass distributions were fitted by Gaussian distribution and mass variance extracted. At below-barrier energies, the mass variance was found to increase with decrease in energy for both nuclei. Results from present work were compared with existing data for induced fission of 224,226Th and 228U around barrier energies. Enhancement in mass variance of 225,227Pa nuclei at below-barrier energies shows evidence for presence of asymmetric fission events mixed with symmetric fission events. This is in agreement with the results of mass distributions of nearby nuclei 224,226Th and 228U where two-mode fission process was observed. Two-mode feature of fission arises due to the shell effects changing the landscape of the potential-energy surfaces at low excitation energies. The excitation-energy dependence of the mass variance gives strong evidence for survival of microscopic shell effects in fission of light actinide nuclei 225,227Pa with initial excitation energy ∼30-50 MeV.

  11. Determination of minor actinides fission cross sections by means of transfer reactions

    We present an original method that allows to determine neutron-induced cross sections of very short-lived minor actinides. This indirect method, based on the use of transfer reactions, has already been applied with success for the determination of the neutron-induced fission and capture cross section of 233Pa, a key nucleus in the 232Th - 233U fuel cycle. A recent experiment using this technique has been performed to determine the neutron-induced fission cross sections of 242,243,244Cm and 241Am which are present in the nuclear waste of the current U-Pu fuel cycle. These cross sections are highly relevant for the design of reactors capable to incinerate minor actinides. The first results will be illustrated. (authors)

  12. Measurement of fast neutron induced fission cross section of minor-actinide

    In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am). Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA was measured using Dynamitron Accelerator in Tohoku University. New or improved techniques and tools with high precision and fast timing capability were developed for this study. Those are as follows: (1) Development of a sealed fission chamber, (2) Intensification of Li neutron target, (3) Improvement of time-resolution of Time-of-Flight (TOF) electronic circuit, (4) Introduction of MA (Np237, Am241 and Am243) samples with large sample mass and (5) Introduction of a U235 sample with high purity. Using these improved tools and samples, fission cross section of Np237 was measured between 10 to 100 keV. On the other hand, averaged fission cross section for Maxwell distribution spectrum with kt = 25.3 keV was measured for Am241 and Am243. (author)

  13. Neutron Capture and Fission Measurements on Actinides at Dance

    Chyzh, A.; Wu, C. Y.; Kwan, E.; Henderson, R. A.; Gostic, J. M.; Ullmann, J. L.; Bredeweg, T. A.; Jandel, M.; Couture, A. J.; O'Donnell, J. M.; Haight, R. C.; Lee, H. Y.

    2013-03-01

    The prompt γ-ray energy and multiplicity distributions in the spontaneous fission of 252Cf have been measured using a highly granular 4π γ-ray calorimeter. Corrections were made for both energy and multiplicity distributions according to the detector response, which is simulated numerically using a model validated with the γ-ray calibration sources. A comparison of the total γray energy distribution was made between the measurement and a simulation by random sampling of the corrected γ-ray energy and multiplicity distributions through the detector response. A reasonable agreement is achieved between the measurement and simulation, indicating weak correlations between γ-ray energy and multiplicity. Moreover, the increasing agreement with increasing multiplicity manifests the stochastic aspect of the prompt γ decay in spontaneous fission. This calorimeter was designed for the study of neutron capture reactions and an example is given, where the238Pu(n, γ) measurement was carried out in the laboratory environment for the first time.

  14. Progress in fission product nuclear data

    This is the 12th issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the IAEA. The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The type of activities included are: measurements, compilations and evaluations of fission product yields (neutron induced and spontaneous fission), neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products and lumped fission product data (decay heat, absorption etc.). The first part of the report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The second part contains recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences

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

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

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

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

    2015-07-11

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

  17. Experimental study of delayed neutron emission from photo-fission of actinides

    Recently there has been a strong interest in a number of applications using the detection of delayed neutrons. Some ongoing projects, as non-destructive characterization of nuclear waste, have shown an urgent need of associated nuclear data. In this context, an experimental program has been launched to measure the delayed neutron yields and their time spectra from photo-fission of actinides. The very first measurements had been performed earlier in our laboratory with the uranium-238 and thorium-232 targets. In this thesis work the follow-up experiments, data analysis and results are reported for uranium-235 and neptunium-237. The high efficiency delayed neutron detector was designed, optimised, constructed and employed for these experiments with the endpoint Bremsstrahlung photons from 15 to 18 MeV. The photo-fission delayed neutron group parameters (ai, λi) were obtained and compared both with earlier work found in the literature and our own modelling results. The energy dependence of the total delayed neutron yield (νd) was also experimentally studied with the endpoint Bremsstrahlung photons in the energy range from 12 to 18 MeV for the uranium-235, 238, thorium-232, and neptunium-237 targets. Finally, some feasibility experiments were successfully performed with mixed samples in order to quantify the different actinide presence in the target. In parallel, the energy dependence of the total delayed neutron yield (νd) was also experimentally studied for thorium-232 but from neutron induced fission. (author)

  18. Actinide Capture and Fission Cross Section Measurements Within the Mini-Inca Project

    Full text of publication follows: The Mini-INCA project is devoted to precise description of the transmutation chain of Actinides within high thermal neutron fluxes. It uses the High Flux Reactor of ILL (Laue Langevin Institute) as an intense thermal neutron source to measure capture and fission cross sections. Two irradiation channels are dedicated for those measurements offering a diversity of fluxes ranging from pure thermal neutrons to 15% epithermal neutrons with intensities as high as 1*1015 n/cm2/s. Standard nuclear techniques for measurements, such as α and γ-spectroscopy of irradiated samples, have been extended in order to stand all constraints due to the irradiation in high fluxes. In particular new types of fission micro-chambers have been developed to follow online the evolution of one actinide and to measure its fission cross section in reference to 235U(n,F) standard reaction. This type of neutron detector will be used within the MEGAPIE target to on-line characterise the neutron flux and to study the potentiality of such target in terms of incineration. (author)

  19. Even-odd effects in the prompt fission emission of even Z actinides

    Tudora Anabella

    2016-01-01

    Full Text Available The investigation of even-odd effects in the prompt emission of even Z actinides showed a sawtooth shape of ν(Z with staggering in the asymmetric fission region. Average prompt emission quantities as a function of A, e.g. ν(A, of even Z fragmentations are higher than those of odd Z fragmentations and they exhibit oscillations with a periodicity of about 5 mass units in the asymmetric fission region. This periodicity is not due to the Z even-odd effect in fragment distributions. The even-odd effect in (TKE is increasing with increasing TKE and it decreases with increasing mass of the fissioning nucleus. The global even-odd effect in total average prompt emission quantities is decreasing with increasing mass of the fissioning nucleus. In the case of an even-odd fissioning nucleus, 234U(n,f, the global even-odd effect in prompt emission quantities exhibits a very slow variation with the incident neutron energy.

  20. Even-odd effects in the prompt fission emission of even Z actinides

    Tudora, Anabella; Hambsch, Franz-Josef; Giubega, Georgiana; Visan, Iuliana

    2016-03-01

    The investigation of even-odd effects in the prompt emission of even Z actinides showed a sawtooth shape of ν(Z) with staggering in the asymmetric fission region. Average prompt emission quantities as a function of A, e.g. ν(A), of even Z fragmentations are higher than those of odd Z fragmentations and they exhibit oscillations with a periodicity of about 5 mass units in the asymmetric fission region. This periodicity is not due to the Z even-odd effect in fragment distributions. The even-odd effect in (TKE) is increasing with increasing TKE and it decreases with increasing mass of the fissioning nucleus. The global even-odd effect in total average prompt emission quantities is decreasing with increasing mass of the fissioning nucleus. In the case of an even-odd fissioning nucleus, 234U(n,f), the global even-odd effect in prompt emission quantities exhibits a very slow variation with the incident neutron energy.

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

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

    2000-07-01

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

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

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

  3. Methodology and experimental setup for measuring short-lives fission product yields in actinides induced fission by charged particles; Metodologia e montagem experimental para a medicao de rendimentos de produtos de fissao de meia vida curta na fissao de actinideos por particulas carregadas

    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.

  4. Actinide production in /sup 136/Xe bombardments of /sup 249/Cf

    Gregorich, K.E.

    1985-08-01

    The production cross sections for the actinide products from /sup 136/Xe bombardments of /sup 249/Cf at energies 1.02, 1.09, and 1.16 times the Coulomb barrier were determined. Fractions of the individual actinide elements were chemically separated from recoil catcher foils. The production cross sections of the actinide products were determined by measuring the radiations emitted from the nuclides within the chemical fractions. The chemical separation techniques used in this work are described in detail, and a description of the data analysis procedure is included. The actinide production cross section distributions from these /sup 136/Xe + /sup 249/Cf bombardments are compared with the production cross section distributions from other heavy ion bombardments of actinide targets, with emphasis on the comparison with the /sup 136/Xe + /sup 248/Cm reaction. A technique for modeling the final actinide cross section distributions has been developed and is presented. In this model, the initial (before deexcitation) cross section distribution with respect to the separation energy of a dinuclear complex and with respect to the Z of the target-like fragment is given by an empirical procedure. It is then assumed that the N/Z equilibration in the dinuclear complex occurs by the transfer of neutrons between the two participants in the dinuclear complex. The neutrons and the excitation energy are statistically distributed between the two fragments using a simple Fermi gas level density formalism. The resulting target-like fragment initial cross section distribution with respect to Z, N, and excitation energy is then allowed to deexcite by emission of neutrons in competition with fission. The result is a final cross section distribution with respect to Z and N for the actinide products. 68 refs., 33 figs., 6 tabs.

  5. Dependence of Fission-Fragment Properties On Excitation Energy For Neutron-Rich Actinides

    Ramos D.

    2016-01-01

    Isotopic fission yields of 250Cf, 244Cm, 240Pu, 239Np and 238U are presented in this work. With this information, the average number of neutrons as a function of the atomic number of the fragments is calculated, which reflects the impact of nuclear structure around Z=50, N=80 on the production of fission fragments. The characteristics of the Super Long, Standard I, Standard II, and Standard III fission channels were extracted from fits of the fragment yields for different ranges of excitation energy. The position and contribution of the fission channels as function of excitation energy are presented.

  6. Progress in fission product nuclear data

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

  7. Fission product behaviour in severe accidents

    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

  8. Mesures de sections efficaces de fission induite par neutrons sur des actinides du cycle du thorium à n_TOF.

    Ferrant, Laure

    2005-01-01

    Dans le contexte des études sur les systèmes innovants de production d'énergie, des réacteurs exploitant le combustible thorium sont envisagés. Les sections efficaces de fission induite par neutrons des actinides qui y sont engagés entrent en jeu dans les simulations de scénarios. Pour les alimenter, des bases de données sont produites à partir de résultats expérimentaux et de modèles. Pour certains noyaux, elles présentent des lacunes ou des désaccords. Pour compléter ces bases de données, n...

  9. Ceramic Hosts for Fission Products Immobilization

    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

  10. Ceramic Hosts for Fission Products Immobilization

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

  11. Rapid Separation of Fission Product 141La

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

    2013-01-01

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

  12. Study of actinides fission induced by multi-nucleon transfer reactions in inverse kinematics; Etude de la fission d'actinides produits par reactions de transfert multinucleon en cinematique inverse

    Derkx, X.

    2010-10-15

    The study of actinide fission encounters two major issues. On one hand, measurements of the fission fragment distributions and the fission probabilities allow a better understanding of the fission process itself and the discrimination among the models of nuclear structure and dynamics. On the other hand, new measurements are required to improve nuclear data bases, which are a key component for the design of new generation reactors and radio-toxic waste incinerators. This thesis is in line with different French and American experimental projects using the surrogate method, i.e. transfer reactions leading to the same compound nuclei as in neutron irradiation, allowing the study of fission of actinides which are inaccessible by conventional techniques, whereas they are important for applications. The experiment is based on multi-nucleon transfer reactions between a {sup 238}U beam and a {sup 12}C target, using the inverse kinematics technique to measure, for each transfer channel, the complete isotopic distributions of the fission fragments with the VAMOS spectrometer. The work presented in this dissertation is focused on the identification of the transfer channels and their properties, as their angular distributions and the distributions of the associated excitation energy, using the SPIDER telescope to identify the target recoil nuclei. This work of an exploratory nature aims to generalize the surrogate method to heavy transfers and to measure, for the first time, the fission probabilities in inverse kinematics. The obtained results are compared with available direct kinematics and neutron irradiation measurements. (author)

  13. Comparison of actinide production in traveling wave and pressurized water reactors

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of 239Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  14. Comparison of actinide production in traveling wave and pressurized water reactors

    Osborne, A.G.; Smith, T.A.; Deinert, M.R. [Department of Mechanical Engineering, University of Texas at Austin, Austin, TX (United States)

    2013-07-01

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

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

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

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

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

    1997-12-31

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

  17. Production of fission 131I

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

  18. Fission product retention in HTGR fuels

    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

  19. Fragment properties from fission of actinide nuclei induced by 6-10 MeV bremsstrahlungI

    Gook, A.; Eckardt, C.; Enders, J.; Freudenberger, M.; von Neumann-Cosel, P.; Oberstedt, A.; Oberstedt, S.; Richter, A.

    Experiments to investigate the photon-induced fission of actinide nuclei at excitation energies in the vicinity of the fission barrier are carried out at the superconducting Darmstadt linear electron accelerator S-DALINAC. A twin-Frisch-grid ionization chamber is used to deduce mass, total kinetic energy, and angular distributions of the fission fragments. First experiments on 238U and 234U have shown that the experimental setup provides excellent conditions for investigating low-energy bremsstrahlung induced fission. Further experiments on 234U and 232Th are currently in progress. In this contribution results from the first experiment on fission fragment mass and total kinetic energy distributions from 234,238U are presented along with preliminary data from an on-going investigation of angular distributions from 234U(γ, f)

  20. Downstream behavior of fission products

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

  1. Aerosols and fission product transport

    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

  2. Calculation code of the fission products activity

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

  3. Progress in fission product nuclear data

    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

  4. Progress in fission product nuclear data

    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

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

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

    2002-02-01

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

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

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

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

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

  8. Measurement of cross-sections of fission reactions induced by neutrons on actinides from the thorium cycle at n-TOF facility

    In the frame of innovating energy source system studies, thorium fuel cycle reactors are considered. Neutron induced fission cross section on such cycle involved actinides play a role in scenario studies. To feed them, data bases are built with experimental results and nuclear models. For some nuclei, they are not complete or in disagreement. In order to complete these data bases, we have built an original set up, consisting in an alternation of PPACs (Parallel Plate Avalanche Chamber) and ultra - thin targets, which we installed on n-TOF facility. We describe detectors, set up, and the particular care brought to target making and characterization. Fission products in coincidence are detected with precise time measurement and localization with delay line read out method. We contributed, within the n-TOF collaboration, to the CERN brand new intense spallation neutron source characterization, based on time of flight measurement, and we describe its characteristics and performances. We were able to measure such actinide fission cross sections as 232Th, 234U, 233U, 237Np, 209Bi, and natPb relative to 235U et 238U standards, using an innovative acquisition system. We took advantage of the lame accessible energy field, from 0.7 eV to 1 GeV, combined with the excellent energy resolution in this field. Data treatment and analysis advancement are described to enlighten performance and limits of the obtained results. (author)

  9. Recent progress in analysis for fission products

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

  10. Thermodynamic analysis of volatile organometallic fission products

    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)

  11. The effect of corrosion product colloids on actinide transport

    The near field of the proposed UK repository for ILW/LLW will contain containers of conditioned waste in contact with a cementious backfill. It will contain significant quantities of iron and steel, Magnox and Zircaloy. Colloids deriving from their corrosion products may possess significant sorption capacity for radioelements. If the colloids are mobile in the groundwater flow, they could act as a significant vector for activity transport into the far field. The desorption of plutonium and americium from colloidal corrosion products of iron and zirconium has been studied under chemical conditions representing the transition from the near field to the far field. Desorption Rd values of ≥ 5 x 106 ml g-1 were measured for both actinides on these oxides and hydroxides when actinide sorption took place under the near-field conditions and desorption took place under the far-field conditions. Desorption of the actinides occurred slowly from the colloids under far-field conditions when the colloids had low loadings of actinide and more quickly at high loadings of actinide. Desorbed actinide was lost to the walls of the experimental vessel. (author)

  12. Development of fission Mo-99 production technology

    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

  13. Development of fission Mo-99 production technology

    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.

  14. The chemistry of the fission products

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

  15. Adsorption of fission products on mediterranean mud

    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)

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

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

  17. Systematics of Fission-Product Yields

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

  18. Modeling Fission Product Sorption in Graphite Structures

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

    2013-04-08

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

  19. Electron spectra from decay of fission products

    Dickens, J K

    1982-09-01

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

  20. Electrolytic extraction of fission noble metals for waste minimizing in advanced actinide separation system

    Electrochemistry for recovering fission platinum group elements (Pd. Ru) and Re (simulator of Te) from HNO3 solutions, Electrolytic Extraction (EE) method was applied as a basic technology in the new actinides separation process. The deposition yields of these elements increased by the decrease of the initial HNO3, concentration. Co-existence of Pd2+ ion accelerates the deposition of RuNO3- and ReO4- ions, and especially Ru deposition yield was over 99% when Pd2+ ion was added during the electrolysis in 2.SN HNO3 solution. Addition of reducing reagents (hydrazine nitrate and HAN) increased Pd2+ deposition rate, however, these and the other complexing reagents (e.g., oxalic acid and EDTA) decreased the Ru deposition because of preferential Pd2+ deposition as well as those complexation with RuNO3-, respectively. On the electrolytic extraction from the simulated HLLW, the elements which had nobler standard redox potential (E0) tended to show higher deposition yields; the elements which had E0 over 0.7 V (Ru, Te, Se, Rh, Pd) can be quantitatively recovered by 3 hr. electrolysis without dilution of HLLW. (author)

  1. Fission products in glasses. Pt. 2

    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)

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

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

  3. Fission cross-section measurements on 233U and minor actinides at the CERN n-TOF facility

    Neutron-induced fission cross-sections of minor actinides have been measured at the white neutron source n-TOF at CERN, Geneva. The studied isotopes include 233U, interesting for Th/U based nuclear fuel cycles, 241,243Am and 245Cm, relevant for transmutation and waste reduction studies in new generation fast reactors (Gen-IV) or Accelerator Driven Systems. The measurements take advantage of the unique features of the n-TOF facility, namely the wide energy range, the high instantaneous neutron flux and the low background. Results for the involved isotopes are reported from ∼30 meV to around 1 MeV neutron energy. The measurements have been performed with a dedicated Fission Ionization Chamber (FIC), relative to the standard cross-section of the 235U fission reaction, measured simultaneously with the same detector. Results are here reported. (authors)

  4. JNDC nuclear data library of fission products

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

  5. Fission product release and thermal behaviour

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

  6. Attenuation mechanisms in the transport of in-vessel radiological source term fission products in an LMFBR

    Quantifying the release of radiological source term fission products from an LMFBR reactor vessel (RV) is a necessary input to the containment analysis. To estimate this initial source term value, the distribution of the fission products and actinides inside the RV, prior to release, must be known. The in-vessel source term fission product distribution and transport behavior is also essential in assessing and mitigating the plant contamination and cleanup problems which occur from any significant core disruption. This paper attempts to summarize the current knowledge on the behavior of several radioisotopes in different environments created by the accident, without dealing with the modeling of the transport process itself

  7. Measurement of cross-sections of fission reactions induced by neutrons on actinides from the thorium cycle at n-TOF facility; Mesures de sections efficaces de fission induite par neutrons sur des actinides du cycle du thorium a n-TOF

    Ferrant, L

    2005-09-01

    In the frame of innovating energy source system studies, thorium fuel cycle reactors are considered. Neutron induced fission cross section on such cycle involved actinides play a role in scenario studies. To feed them, data bases are built with experimental results and nuclear models. For some nuclei, they are not complete or in disagreement. In order to complete these data bases, we have built an original set up, consisting in an alternation of PPACs (Parallel Plate Avalanche Chamber) and ultra - thin targets, which we installed on n-TOF facility. We describe detectors, set up, and the particular care brought to target making and characterization. Fission products in coincidence are detected with precise time measurement and localization with delay line read out method. We contributed, within the n-TOF collaboration, to the CERN brand new intense spallation neutron source characterization, based on time of flight measurement, and we describe its characteristics and performances. We were able to measure such actinide fission cross sections as {sup 232}Th, {sup 234}U, {sup 233}U, {sup 237}Np, {sup 209}Bi, and {sup nat}Pb relative to {sup 235}U et {sup 238}U standards, using an innovative acquisition system. We took advantage of the lame accessible energy field, from 0.7 eV to 1 GeV, combined with the excellent energy resolution in this field. Data treatment and analysis advancement are described to enlighten performance and limits of the obtained results. (author)

  8. Correlation of recent fission product release data

    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

  9. A device for trapping fission products

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

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

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

  11. Fission Product Sorptivity in Graphite

    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

  12. Actinide Foil Production for MPACT Research

    Beller, Denis

    2012-10-30

    Sensitive fast-neutron detectors are required for use in lead slowing down spectrometry (LSDS), an active interrogation technique for used nuclear fuel assay for Materials Protection, Accounting, and Controls Technologies (MPACT). During the past several years UNLV sponsored a research project at RPI to investigate LSDS; began development of fission chamber detectors for use in LSDS experiments in collaboration with INL, LANL, and Oregon State U.; and participated in a LSDS experiment at LANL. In the LSDS technique, research has demonstrated that these fission chamber detectors must be sensitive to fission energy neutrons but insensitive to thermal-energy neutrons. Because most systems are highly sensitive to large thermal neutron populations due to the well-known large thermal cross section of 235U, even a miniscule amount of this isotope in a fission chamber will overwhelm the small population of higher-energy neutrons. Thus, fast-fission chamber detectors must be fabricated with highly depleted uranium (DU) or ultra-pure thorium (Th), which is about half as efficient as DU. Previous research conducted at RPI demonstrated that the required purity of DU for assay of used nuclear fuel using LSDS is less than 4 ppm 235U, material that until recently was not available in the U.S. In 2009 the PI purchased 3 grams of ultra-depleted uranium (uDU, 99.99998% 238U with just 0.2 ± 0.1 ppm 235U) from VNIIEF in Sarov, Russia. We received the material in the form of U3O8 powder in August of 2009, and verified its purity and depletion in a FY10 MPACT collaboration project. In addition, chemical processing for use in FC R&D was initiated, fission chamber detectors and a scanning alpha-particle spectrometer were developed, and foils were used in a preliminary LSDS experiment at a LANL/LANSCE in Sept. of 2010. The as-received U3O8 powder must be chemically processed to convert it to another chemical form while maintaining its purity, which then must be used to electro-deposit U

  13. Reactor with very low fission product inventory

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

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

    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.

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

    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

  16. Chemistry of fission products for accident analysis

    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

  17. Ex-vessel fission product release modeling

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

  18. Fission product release mechanisms and groupings

    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)

  19. The release of semi- and low-volatile fission products from bare UO2 samples during post-irradiation annealing

    The release data of semi-volatiles (Ru, Nb, Ba, Sb and Tc) and low-volatiles (Zr, La, Ce, Eu, Pr, Nd, Pm and actinides Np) measured from 154 tests performed at the Chalk River Laboratories are summarized. The release of semi-volatiles in general are controlled by the environment and temperature, although the release is different from element to element. Most of the low-volatile fission products are released only because of the matrix stripping (volatilization of UO2) in combination with high temperatures. Promethium is released in air at 1600 degree C, but no release was detected at 1600 degree C in steam or in Ar. No release of actinides Np was detected in air at 500∼1050 degree C. Matrix stripping affects both semi- and low-volatile fission product release only slightly. The release rates for the fission products can be used to directly calculate the release from bare UO2 under accident conditions

  20. Fission product decay heat for thermal reactors

    Dickens, J. K.

    1979-01-01

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

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

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

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

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

    2005-12-01

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

  3. Model for fission-product calculations

    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

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

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

    2016-06-01

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

  5. Development of fission Mo-99 production technology

    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.

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

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

  7. The effects of corrosion product colloids on actinide transport

    This report assesses the possible effects of colloidal corrosion products on the transport of actinides from the near field of radioactive waste repositories. The desorption of plutonium and americium from colloidal corrosion products of iron and zirconium was studied under conditions simulating a transition from near-field to far-field environmental conditions. Desorption of actinides occurred slowly from the colloids under far-field conditions. Measurements of particle stability showed all the colloids to be unstable in the near field. Stability increased under far-field conditions or as a result of the evolution of the near field. Migration of colloids from the near field is unlikely except in the presence of organic materials. (Author)

  8. Delayed Neutrons and Photoneutrons from Fission Products

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

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

    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

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

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

  11. Behaviour of Fission Products in Liquid Sodium

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

  12. Preparative electrophoresis of industrial fission product solutions

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

  13. HAMCIND, Cell Burnup with Fission Products Poisoning

    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

  14. Fission product measurement methods. Present state of knowledge

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

  15. Fission product release mechanisms and pathways

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

  16. Gas-phase transport of fission products

    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

  17. Development of fission Mo production technology

    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

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

    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

  19. Neutron-induced fission cross sections of short-lived actinides via the surrogate reaction method

    A brief discussion of surrogate reaction methods has been made and some of the recent results on neutron induced fission cross section measurements have been presented. The validation of the EMPIRE-3.1. predictions on neutron induced cross sections corresponding to fission barriers used from Barrier Formula (BF) and RIPL-1 libraries have been discussed

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

    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

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

    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

  2. Interplay of fission modes in mass distribution of light actinide nuclei 225,227Pa

    Dubey, R; Jhingan, A; Kaur, Gurpreet; Mukul, Ish; Mohanto, G; Siwal, D; Saneesh, N; Banerjee, T; Thakur, Meenu; Mahajan, Ruchi; Kumar, N; Chatterjee, M B

    2015-01-01

    Fission-fragment mass distributions were measured for 225,227Pa nuclei formed in fusion reactions of 19F + 206, 208Pb around fusion barrier energies. Mass-angle correlations do not indicate any quasi-fission like events in this bombarding energy range. Mass distributions were fitted by Gaussian distribution and mass variance extracted. At below-barrier energies, the mass variance was found to increase with decrease in energy for both nuclei. Results from present work were compared with existing data for induced fission of 224, 226Th and 228U around barrier energies. Enhancement in mass variance of 225, 227Pa nuclei at below-barrier energies shows evidence for presence of asymmetric fission events mixed with symmetric fission events. This is in agreement with the results of mass distributions of nearby nuclei 224, 226Th and 228U where two-mode fission process was observed. Two-mode feature of fission arises due to the shell effects changing the landscape of the potential energy surfaces at low excitation energ...

  3. Production techniques of fission 99Mo

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

  4. Energy Dependence of Plutonium Fission-Product Yields

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

  5. Energy Dependence of Plutonium Fission-Product Yields

    Lestone, J. P.

    2011-12-01

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

  6. Dynamical approach to isotopic-distribution of fission fragments from actinide nuclei

    Ishizuka, Chikako; Chiba, Satoshi; Karpov, Alexander V.; Aritomo, Yoshihiro

    2016-06-01

    Measurements of the isotope distribution of fission fragments, often denoted as the primary fission yield (pre-neutron yield) or independent fission yield (post-neutron yield) are still challenging at low excitation energies, so that it is important to investigate it within a theory. Such quantities are vital for applications as well. In this study, fragment distributions from the fission of U isotopes at low excitation energies are studied using a dynamical model. The potential energy surface is derived from the two center shell model including the shell and pairing corrections. In order to calculate the charge distribution of fission fragments, we introduce a new parameter ηZ as the charge asymmetry, in addition to three parameters describing a nuclear shape, z as the distance between two centers of mass, δ as fragment deformation, and ηA as the mass asymmetry. Using this model, we calculated the isotopic distribution of 236U for the n-induced process 235U + n → 236U at low excitation energies. As a result, we found that the current model can well reproduce isotopic fission-fragment distribution which can be compared favorably with major libraries.

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

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

  8. Decay Chain Deduction of Uranium Fission Products.

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

    2016-07-01

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

  9. Fission product transport in the atmosphere

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

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

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

    2015-06-01

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

  11. The role of fission products in whole core accidents

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

  12. Energy production using fission fragment rockets

    Chapline, G.; Matsuda, Y.

    1991-08-01

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

  13. Energy production using fission fragment rockets

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

  14. Fission product and aerosol behaviour within the containment

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

  15. Migration of fission products in UO2. Final report

    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

  16. Particular features of ternary fission induced by polarized neutrons in the major actinides U,235233 and Pu,241239

    Gagarski, A.; Gönnenwein, F.; Guseva, I.; Jesinger, P.; Kopatch, Yu.; Kuzmina, T.; Lelièvre-Berna, E.; Mutterer, M.; Nesvizhevsky, V.; Petrov, G.; Soldner, T.; Tiourine, G.; Trzaska, W. H.; Zavarukhina, T.

    2016-05-01

    Ternary fission in (n ,f ) reactions was studied with polarized neutrons for the isotopes U,235233 and Pu,241239. A cold longitudinally polarized neutron beam was available at the High Flux Reactor of the Institut Laue-Langevin in Grenoble, France. The beam was hitting the fissile targets mounted at the center of a reaction chamber. Detectors for fission fragments and ternary particles were installed in a plane perpendicular to the beam. In earlier work it was discovered that the angular correlations between neutron spin and the momenta of fragments and ternary particles were very different for 233U or 235U. These correlations could now be shown to be simultaneously present in all of the above major actinides though with different weights. For one of the correlations it was observed that up to scission the compound nucleus is rotating with the axis of rotation parallel to the neutron beam polarization. Entrained by the fragments also the trajectories of ternary particles are turned away albeit by a smaller angle. The difference in turning angles becomes observable upon reversing the sense of rotation by flipping neutron spin. All turning angles are smaller than 1∘. The phenomenon was called the ROT effect. As a distinct second phenomenon it was found that for fission induced by polarized neutrons an asymmetry in the emission probability of ternary particles relative to a plane formed by fragment momentum and neutron spin appears. The asymmetry is attributed to the Coriolis force present in the nucleus while it is rotating up to scission. The size of the asymmetry is typically 10-3. This asymmetry was termed the TRI effect. The interpretation of both effects is based on the transition state model. Both effects are shown to be steered by the properties of the collective (J ,K ) transition states which are specific for any of the reactions studied. The study of asymmetries of ternary particle emission in fission induced by slow polarized neutrons provides a new

  17. Interplay of fission modes in mass distribution of light actinide nuclei 225,227Pa

    R Dubey; Sugathan, P; Jhingan, A.; Gurpreet Kaur; Ish Mukul; G. Mohanto; Siwal, D.; Saneesh, N.; T Banerjee; Meenu Thakur; Ruchi Mahajan; Kumar, N; Chatterjee, M. B.

    2016-01-01

    Fission-fragment mass distributions were measured for 225,227 Pa nuclei formed in fusion reactions of 19 F + 206,208 Pb around fusion barrier energies. Mass-angle correlations do not indicate any quasi-fission like events in this bombarding energy range. Mass distributions were fitted by Gaussian distribution and mass variance extracted. At below-barrier energies, the mass variance was found to increase with decrease in energy for both nuclei. Results from present work were compared with exis...

  18. Neutron-induced fission cross sections of short-lived actinides with the surrogate reaction method

    Neutron-induced fission cross sections for 242,243Cm and 241Am have been obtained with the surrogate reaction method. Recent results for the neutron-induced cross section of 243Cm are questioned by the present data. For the first time, the 242Cm cross section has been determined up to the onset of second-chance fission. The good agreement at the lowest excitation energies between the present results and the existing neutron-induced data indicates that the distributions in spin and parity of states populated with both techniques are similar.

  19. Resuspension of fission products from sump water

    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

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

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

  1. Fuel rod internal chemistry and fission products behaviour

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

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

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

  3. Production and measurement of minor actinides in the commercial fuel cycle

    Stanbro, W.D. [comp.

    1997-03-01

    The minor actinide elements, particularly neptunium and americium, are produced as a normal byproduct of the operation of thermal power reactors. Because of the existence of long-lived isotopes of these elements, they constitute the major sources of the residual radiation in spent fuel or in wastes resulting from reprocessing. This has led to examinations by some countries of the possibility of separating the minor actinides from waste products. The papers found in this report address the production of minor actinides in common thermal power reactors as well as approaches to measure these materials in various media. The first paper in this volume, {open_quotes}Production of Minor Actinides in the Commercial Fuel Cycle,{close_quotes} uses calculations with the ORIGEN2 reactor and decay code to estimate the amounts of minor actinides in spent fuel and separated plutonium as a function of reactor irradiation and the time after discharge. The second paper, {open_quotes}Destructive Assay of Minor Actinides,{close_quotes} describes a number of promising approaches for the chemical analysis of minor actinides in the various forms in which they are found at reprocessing plants. The next paper, {open_quotes}Hybrid KED/XRF Measurement of Minor Actinides in Reprocessing Plants,{close_quotes} uses the results of a simulation model to examine the possible applications of the hybrid KED/XRF instrument to the determination of minor actinides in some of the solutions found in reprocessing plants. In {open_quotes}Calorimetric Assay of Minor Actinides,{close_quotes} the authors show some possible extensions of this powerful technique beyond the normal plutonium assays to include the minor actinides. Finally, the last paper in this volume, {open_quotes}Environment Measurements of Transuranic Nuclides,{close_quotes} discusses what is known about the levels of the minor actinides in the environment and ways to analyze for these materials in environmental matrices.

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

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

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

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

    2014-09-01

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

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

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

  7. Neutron-induced fission cross sections of short-lived actinides with the surrogate reaction method

    We present a review of the fission cross section measurements made by the CENBG collaboration over the last years using the surrogate reaction method. For example the neutron-induced fission cross sections of 233Pa (T1/2=27 d), 242Cm (T1/2=162.8 d) and 243Cm (T1/2=29.1 y) have been obtained by our group with this technique. The advantages and the difficulties of the surrogate method are discussed. Special attention is paid to the comparison between cross sections measured with the surrogate method and those obtained directly with neutrons at low energies. This comparison provides information on possible differences between the spin-parity distributions achieved in the two methods. We measured for the first time the fission cross section of 233Pa. Our results for 231Pa(n,f) revealed that the existing neutron-induced data overestimated the fission cross section above 1.5 MeV. The deduced 241Am(n,f) and 242Cm(n,f) cross sections agree with the available data obtained via neutron-induced reactions. The good agreement observed at the lowest neutron energies between the present results and the neutron-induced data for 242Cm(n,f) and 243Cm(n,f) indicates that the population of excited states generated by the transfer reactions used in this work is similar to the distribution fed in neutron induced reactions. This agreement illustrates the potential of the surrogate reaction method to provide neutron-induced fission cross sections for short-lived nuclei

  8. ENDF/B fission product decay data

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

  9. Library of data for fission products

    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

  10. Core degradation and fission product release

    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

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

    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

  12. Immobilization of fission products in phosphate ceramic waste forms

    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

  13. Some 235U reference fission product yield data evaluation

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

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

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

  15. Immobilization of fission products in phosphate ceramic waste forms

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

    1996-10-01

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

  16. ESF system fission product retention effectiveness

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

  17. Building the European Research Area in nuclear fission pioneering steps in actinide science

    The concept of the European Research Area (ERA) aims at closer development of research policies in Europe and closer networking of research capacities, to reduce fragmentation of research in Europe. The goal is to make European research more effective and competitive. Several approaches are made to create ERA. The European Research Framework Programme is one tool in this context, with the introduction of the new instruments, Integrated Projects, Networks of Excellence and Integrated Infrastructure Initiatives. Actinide science is one area that could benefit from better coordination and more effective use of the research capacities, both human and physical. The European Commission is thus funding a Network of Excellence (ACTINET-6) and an Integrated Project (EUROPART) in this area within the sixth EURATOM Framework Programme. (author)

  18. New molecules to separate actinides: the picolinamides

    The reprocessing of spent fuel is made with the Purex process, funded on liquid-liquid extraction of uranium nitrates(VI) and plutonium nitrates(IV) by the BTP (tributyl phosphate). To improve this proceeding, we look for extractants which allow, beyond U and Pu extractions, these of actinides (II) and allow separation of the whole actinides from the fission products, which have an important fraction of lanthanides. A new family seems to give good results: the picolinamides

  19. Effects of organic degradation products on the sorption of actinides

    Previous work has shown that products from the chemical degradation of cellulosic matter can significantly reduce sorption of uranium(VI) and plutonium(IV) on geological materials. Uranium(IV) batch sorption experiments have now been performed to study the effect of organic degradation products in a reducing environment. Thorium(IV) sorption has also been studied since thorium is an important radioelement in its own right and has potential use as a simulant for other tetravalent actinides. Sorption onto London clay, Caithness flagstones and St. Bees sandstone was investigated. Experimental conditions were chosen to simulate both those expected close to cementitious repository (pH ∝ 11) and at the edge of the zone of migration of the alkaline plume (pH ∝ 8). Work was carried out with both authentic degradation products and with gluconate, acting as a well-characterized simulant for cellulosic degradation products. The results show that the presence of organic species can cause a reduction in sorption. This is especially so in the presence of a high concentration of gluconate ions, but the reduction is significantly less with authentic degradation products. (orig.)

  20. Fission product and aerosol behaviour within the containment

    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

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

    2001-07-01

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

  2. Fission-product burn-up in fast reactors

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

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

    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)

  4. Measurements of fast neutron capture and fission cross sections of minor actinide isotopes

    The neutron capture cross sections of 240Pu, 242Pu and 241Am were measured in the energy range from 10 to 250 keV, with 197Au and 238U as standards. The subthreshold fission cross sections of 240Pu and 241Am were determined relative to 235U in the energy range from 10 to 250 keV and 10 keV to 1 MeV, respectively. Continuous neutron spectra and in one case monoenergetic neutrons were produced by means of the Li(p,n) and T(p,n) reactions with the Karlsruhe 3-MV pulsed Van de Graaff accelerator. Capture events were detected by a Moxon Rae detector, and fission events, observed with a NE213 liquid scintillator. The high neutron flux available at flight paths as short as 50 to 135 mm allowed a statistical accuracy of 1 to 3% for most of the measured data together with a moderate energy resolution of 10 to 30 ns/m. An overall uncertainty between 5 and 10% was obtained in most of the measurements. A comparison is made to recent data of other authors and to evaluated files. 8 figures, 1 table

  5. Fission barriers and half-lives of actinides in the quasimolecular shape valley

    Royer, G.; Jaffré, M.; Moreau, D.

    2012-10-01

    The energy of actinide nuclei in the fusionlike deformation valley has been determined from a liquid-drop model, taking into account the proximity energy, the mass and charge asymmetries, and the shell and pairing energies. Double-humped potential barriers appear. The saddle point corresponds to the second maximum and to the transition from compact one-body shapes with a deep neck to two touching ellipsoids. The scission point, where the effects of the nuclear attractive forces between the fragments vanish, lies at the end of an energy plateau below the saddle point and corresponds to two well-separated fragments. The kinetic and excitation energies of the fragments come from the energy on this plateau. The shell and pairing effects play a main role to decide the most probable decay path. The heights of the potential barriers roughly agree with the experimental data and the calculated half-lives follow the trend of the experimental values. A shallow third minimum and a third peak appear in specific asymmetric exit channels where one fragment is close to a double magic quasispherical nucleus, while the other one evolves from oblate to prolate shapes.

  6. Separation of actinides (III) from fission lanthanides by non-dispersive liquid-liquid extraction

    Both 2,6-bis (5, 6-dipropyl-1, 2, 4-triazin-3yl) pyridine (n-Pr-BTP), and a synergistic mixture of bis (chlorophenyl) dithio-phosphinic acid ((ClPh)2PSSH) and tri-n-octyl phosphine oxide (TOPO) are able to selectively extract actinides (III) over lanthanides (III) from nitric acid solutions. We performed counter-current extraction tests in hollow fiber modules (HFM) using these extractants. With n-Pr-BTP, up to 99.95% of americium could be separated from the aqueous feed phase. Lanthanide co-extraction was in the range of 1%. With the synergistic mixture, up to 99.99% of americium were removed, with approx. 30% of lanthanide co-extraction. However, with the latter system, specific flow rates more than five times higher were possible. This is due to different mass transfer kinetics in the two systems: Kinetic investigations with n-Pr-BTP performed in a stirred cell showed that the rate of extraction is controlled by a slow chemical reaction. On the other hand, the chemical reaction with (ClPh)2 PSSH + TOPO is fast, and the extraction rate is controlled by diffusion. (author)

  7. Protected Plutonium Production (P3) by transmutation of minor actinides for peace and sustainable prosperity

    'Protected Plutonium Production (P3)' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MAs). Doping the small amount of MAs such as 237Np or 241Am with large neutron capture cross-section into the uranium fuel to enhance the production of 238Pu or 242Pu, which have high spontaneous fission neutron source or also high decay heat to makes the process of the nuclear weapon manufacture and maintenance technologically difficult,can be effective for improving the isotopic barrier of proliferation resistance of the plutonium in thermal reactors. Super weapon grade plutonium could be produced in the blanket of a conventional FBR. However, by increasing the 238Pu or 242Pu ratio in the total plutonium by MAs doping into the fresh blanket, the protected plutonium with high proliferation-resistance can be bred. A new evaluation function, 'attractiveness', defined as a ratio of potential of fission yield to the technological difficulties of nuclear explosive device, has been proposed to evaluate the proliferation resistance of Pu based on the nuclear material property for Plutonium Categorization. The new evaluation function of attractiveness is applied for assessing the existing plutonium criteria as summarized in the following, (a) weapon grade plutonium (b) plutonium with 30% fraction of 240Pu (c) plutonium with 6% fraction of 238Pu (d) plutonium exempt from safeguards. Since both proliferation resistant plutonium compositions (b) and (c) give almost the same value of attractiveness, plutonium is categorized by following well accepted terminology, weapon grade, usable, practically unusable and exempt as shown. It is concluded based on the new evaluation function 'Attractiveness' that P3 mechanism by the transmutation of MA is very effective to improve the proliferation resistance of plutonium. In the conference, the fundamentals of P3 mechanism by transmutation of MA, and the comparison of the 'attractiveness' of

  8. Etude de la fission d'actinides produits par réactions de transfert multinucléon en cinématique inverse

    Derkx, X.

    2010-01-01

    L'étude de la fission des actinides répond à un double enjeu. Les mesures de distributions de fragments de fission et des probabilités de fission permettent une meilleure compréhension du phénomène en lui même et une discrimination des modèles de structure et de dynamique nucléaires. De plus, dans le contexte de la conception de réacteurs nucléaires de nouvelle génération et d'incinérateurs de déchets radiotoxiques, de nouvelles mesures sont indispensables pour améliorer les bases de données ...

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

    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

  10. Fission product revaporization in the reactor cooling system

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

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

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

  12. Neutron production from (α,n) reactions and spontaneous fission in ThO2, UO2, and (U,Pu)O2 fuels

    Available alpha-particle stopping cross-section and 1718O(α,n) cross-section data were adjusted, fitted, and used in calculating the thick-target neutron production function for alpha particles below 10 MeV in oxide fuels. The spent UO2 function produced was folded with actinide decay spectra to determine (α,n) neutron production by each of 89 actinides. Spontaneous-fission (SF) neutron production for 40 actinides was calculated as the product of anti ν(SF) and SF branching-fraction values accumulated or estimated from available data. These contributions and total neutron production in spent UO2 fuel are tabulated and, when combined with any calculated inventory, describe the spent UO2 neutron source. All data are tabulated and methodology is described to permit easy extension to specialized problems

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

    Alexandra Navrotsky

    2010-07-30

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

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

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

  15. Estimate of exposure impact to fission product transformation

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

  16. Fission product behavior in the Molten Salt Reactor Experiment

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

  17. Fission Product Transmutation in Mixed Radiation Fields

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

    2015-01-15

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

  18. 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

    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)

  19. Actinides analysis by accelerator mass spectrometry

    At the ANTARES accelerator at ANSTO a new beamline has been commissioned, incorporating new magnetic and electrostatic analysers, to optimise the efficiency for Actinides detection by Accelerator Mass Spectrometry (AMS). The detection of Actinides, particularly the isotopic ratios of uranium and plutonium, provide unique signatures for nuclear safeguards purposes. We are currently engaged in a project to evaluate the application of AMS to the measurement of Actinides in environmental samples for nuclear safeguards. Levels of certain fission products, Actinides and other radioactive species can be used as indicators of undeclared nuclear facilities or activities, either on-going or in the past Other applications of ultra-sensitive detection of Actinides are also under consideration. neutron-attenuation images of a porous reservoir rock

  20. Evaluation of prompt neutron spectra for minor actinide nuclei

    Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.

    1997-03-01

    Measurement data on fission prompt neutron spectra of minor actinide (MA) is much little, and its accuracy is also unsufficient. Therefore, conventional evaluation value of fission spectra of MA was assumed for its nuclear temperature by using a method of determining from its systemicity owing to assumption of the Maxwell type distribution, but it can be said that this method consider fully to features of MA isotopes. In this paper, some evaluation calculation results are shown by adopting an evaluation method developed by authors and based on modified Madland Nix model and are conducted by concept of physical properties on target nuclei. As a result, by adopting the level density parameter of fission fragments, the inverse process cross section, the fission product yield distribution and the total release energy, effect of inverse process cross section, mass distribution of fission product, calculation results of Cm isotope and systemicity of fission spectra of actinide isotope were investigated. (G.K.)

  1. Chemical separation and nuclear transmutation of by-product actinides

    The paper presents the most important results and conclusions of the assessment studies carried out by the Joint Research Centre-Ispra and by other organizations on the advanced waste disposal strategy based on chemical separation of By-product Actinides (BPA's) from high level liquid waste (HLLW) and their transmutation in nuclear reactors. The technological developments required for the implementation of this strategy have been identified: they concern mainly fuel reprocessing, BPA recovery from all important waste streams and fuel refabrication. After consideration of different strategies for BPA transmutation, the homogeneous recycling in FBR's appears to be most suitable due to its transmutation rate and the compatibility of BPA's with its fuel cycle. The fuel cycle with transmutation has been compared with an advanced reference fuel cycle on the basis of costs and risks. The large effort required for the development and implementation of this new fuel cycle, the increased costs operating the fuel cycle compared with the marginal benefits in the long-term risk of geological disposal, make this strategy not very attractive

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

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

  3. Fission product release from highly irradiated LWR fuel

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

  4. Transport of fission products in matrix and graphite

    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)

  5. Delayed-neutron branching ratios of fission products

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

  6. Feasibility studies of actinide recycle in LMFBRs as a waste management alternative

    A strategy of actinide burnup in LMFBRs is being investigated as a waste management alternative to long term storage of high level nuclear waste. This strategy is being evaluated because many of the actinides in the waste from spent-fuel reprocessing have half-lives of thousands of years and an alternative to geological storage may be desired. From a radiological viewpoint, the actinides and their daughters dominate the waste hazard for decay times beyond about 400 years. Actinide burnup in LMFBRs may be an attractive alternative to geological storage because the actinides can be effectively transmuted to fission products which have significantly shorter half-lives. Actinide burnup in LMFBRs rather than LWRs is preferred because the ratio of fission reaction rate to capture reaction rate for the actinides is higher in an LMFBR, and an LMFBR is not so sensitive to the addition of the actinide isotopes. An actinide target assembly recycle scheme is evaluated to determine the effects of the actinides on the LMFBR performance, including local power peaking, breeding ratio, and fissile material requirements. Several schemes are evaluated to identify any major problems associated with reprocessing and fabrication of recycle actinide-containing assemblies. The overall efficiency of actinide burnout in LMFBRs is evaluated, and equilibrium cycle conditions are determined. It is concluded that actinide recycle in LMFBRs offers an attractive alternative to long term storage of the actinides, and does not significantly affect the performance of the host LMFBR. Assuming a 0.1 percent or less actinide loss during reprocessing, a 0.1 percent loss of less during fabrication, and proper recycle schemes, virtually all of the actinides produced by a fission reactor economy could be transmuted in fast reactors

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

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

  8. Spray removal of fission products in PWR containments

    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)

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

    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)

  10. Atomic masses of fission product nuclei far from stability

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

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

    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

  12. Studies on the Separation of Cesium From Fission Products

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

    2003-01-01

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

  13. Interactions of fission product vapours with aerosols

    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.

  14. Organophosphorus reagents in actinide separations: Unique tools for production, cleanup and disposal

    Interactions of actinide ions with phosphate and organophosphorus reagents have figured prominently in nuclear science and technology, particularly in the hydrometallurgical processing of irradiated nuclear fuel. Actinide interactions with phosphorus-containing species impact all aspects from the stability of naturally occurring actinides in phosphate mineral phases through the application of the bismuth phosphate and PUREX processes for large-scale production of transuranic elements to the development of analytical separation and environment restoration processes based on new organophosphorus reagents. In this report, an overview of the unique role of organophosphorus compounds in actinide production, disposal, and environment restoration is presented. The broad utility of these reagents and their unique chemical properties is emphasized

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

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

  16. Uncertainties in fission-product decay-heat calculations

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

    1997-03-01

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

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

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

  18. Etude structurale et propriétés des verres peralumineux de conditionnement des produits de fission et actinides mineurs"

    Gasnier, Estelle

    2013-01-01

    Ce travail de thèse s’inscrit dans le cadre de la recherche de nouvelles formulations verrières pour le conditionnement des produits de fission et actinides mineurs (PFA). Il s’agit d’étudier une composition de verre dans le domaine peralumineux (défaut de compensateurs de charge en alcalins et alcalino-terreux par rapport à l’aluminium) présentant un taux de charge au moins équivalent à celui du verre R7T7 (18,5 % mass. PFA) et de statuer sur la potentialité de ces matrices vitreuses comme m...

  19. Radiochemical separation of actinides for their determination in environmental samples and waste products

    Gleisberg, B. [Nuclear Engineering and Analytics Rossendorf, Inc. (VKTA), Dresden (Germany)

    1997-03-01

    The determination of low level activities of actinides in environmental samples and waste products makes high demands on radiochemical separation methods. Artificial and natural actinides were analyzed in samples form the surrounding areas of NPP and of uranium mines, incorporation samples, solutions containing radioactive fuel, solutions and solids resutling from the process, and in wastes. The activities are measured by {alpha}-spectrometry and {gamma}-spectrometry. (DG)

  20. Long-Lived Fission Product Transmutation Studies

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

  1. On the quest of production of superheavy nuclei in reactions of 48Ca with the heaviest actinide targets

    The sequence of radioactive decays of an unknown isotope produced in a rare fusion reaction to known lighter isotopes is used to identify mass and atomic number of the mother isotope, which has been separated before from the bulk of other reaction products by an in-flight recoil separator. By this technique the elements 107 to 112 were produced by single atom decay-chain analysis. Such a correlation technique reaches its limit by the occurrence of accidental sequences and it collapses beyond a maximum possible correlation time, at which a true event cannot be distinguished anymore from a random event. 48Ca-induced fusion reactions with actinides are discussed. In 1983 at GSI, Darmstadt and LBL, Berkeley, 48Ca/248Cm-experiments (II) were performed, which are compared to recent 48Ca-experiments at FLNR-Dubna (I) irradiating 244Pu, 242Pu, and 238U. In these experiments production of isotopes of superheavy elements 112 and 114 is claimed. Our analysis of accidental sequences in 48Ca-induced reactions is presented, which is at variance with the published analysis from FLNR-Dubna. We find that the maximum correlation time using continuous beams at today existing separation systems is not in the one-hour regime, but in the few-minute regime. The five spontaneous fission events observed in the FLNR experiments are preceded by signals in the (1-16)-minute range. These times are shown to be longer than the maximum possible correlation times. The preceding signals are decoupled from the spontaneous fission signal and carry no information on the spontaneous fission events observed. Moreover, random probabilities of 0.2 to 0.6 for the signals preceding the fission events indicate that the correlations are of random origin. The evidence to have discovered element 114 in the reported experiments is classified ''very weak''. (orig.)

  2. Time Dependent Radio-toxicology of Fission Products

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

  3. Interpretation of fission product release from overheated fuel

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

  4. Recent progress in fission product separation

    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

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

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

  6. Compilation of fission product yields Vallecitos Nuclear Center

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

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

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

  8. Combining extractant systems for the simultaneous extraction of transuranic elements and selected fission products

    The popularity of solvent extraction (SX) stems from its ability to operate in a continuous mode, to achieve high throughputs and high decontamination factors of product streams, and to utilize relatively small quantities of very selective chemical compounds as metal ion complexants. The chemical pretreatment of nuclear waste for the purpose of waste minimization will probably utilize one or more SX processes. Because of the diversity and complexity of nuclear waste, perhaps the greatest difficulty for the separation chemist is to develop processes that remove not only actinides but also selected fission products in a single process. A stand alone acid-side SX process (TRUEX) for removal of uranium and transuranic elements (Np, Pu, Am) from nuclear waste has been widely reported. Recently, an acid-side SX process (SREX) to extract and recover 90Sr from high-level nuclear waste has also been reported. Both the TRUEX and SREX processes extract Tc to a significant extent although not as efficiently as they extract transuranics and Sr. Ideally one would like to have a process that can extract and recover all actinides as well as 99Tc, 90Sr, and 137Cs. A possible solution to multielement extraction is to mix two extractants with totally different properties into a single process solvent formulation. For this approach to be successful, both extractants must be essentially the same type, either neutral, liquid cationic, or liquid anionic. Experimental work has been carried out on mixed TRUEX and SREX processes, for synthetically created waste, and demonstrates the combined solvent formulation is effective at extracting both the actinides and Tc, as well as Sr. There is no evidence for the presence of either synergistic or antagonistic effects between the two extractants. This demonstates the feasibility of at least part of a combined solvent extraction scheme

  9. Fission properties of actinide nuclei from proton-induced fission at 26.5 and 62.9 MeV incident proton energies

    Demetriou, P.; Keutgen, Thomas; Prieels, René; El Masri, Youssef

    2010-01-01

    Fission properties of proton-induced fission on Th232, Np237, U238, Pu239, and Am241 targets, measured at the Louvain-la-Neuve cyclotron facility at proton energies of 26.5 and 62.9 MeV, are compared with the predictions of the state-of-the-art nuclear reaction code talys. The code couples the multimodal random neck-rupture model with the pre-equilibrium exciton and statistical models to predict fission fragment mass yields, pre- and post-scission neutron multiplicities, and total fission cro...

  10. Trapping technology for gaseous fission products from voloxidation process

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

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

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

  12. Trapping technology for gaseous fission products from voloxidation process

    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.

  13. Kinetics of fission product release prior to fuel slumping

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

  14. Integral test of fission-product cross sections

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

  15. Evaluation and compilation of fission product yields 1993

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

  16. Evaluation and compilation of fission product yields 1993

    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.

  17. Characterization of wastes from fission 99 Mo production

    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)

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

    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

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

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

  20. Waste treatment of fission product solutions containing aluminium nitrate

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

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

    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

  2. Fission product removal from molten salt using zeolite

    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

  3. Comparison of Fission Product Yields and Their Impact

    S. Harrison

    2006-02-01

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

  4. Evaluation of actinide partitioning and transmutation

    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

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

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

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

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

  7. The chemistry of fission products for accident analysis

    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

  8. Fission product chemistry in severe nuclear reactor accidents

    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)

  9. Gaseous fission products monitoring in an irradiation circuit

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

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

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

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

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

  12. Actinide burning and waste disposal

    for the second repository would be emplaced in the first repository. Reprocessing would now include separation of the fission products strontium and cesium. After interim storage for 20-300 years, the remaining cesium would also be emplaced in the first repository. One DOE laboratory proposes an accelerator to destroy actinides and long-lived fission products. The time required for geologic or managed storage is said to be reduced to only one to several centuries

  13. Experiments With Mass-Spectroscopically Separated Fission Product Ions

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

  14. Advanced Aqueous Separation Systems for Actinide Partitioning

    Nash, Kenneth L.; Clark, Sue; Meier, G Patrick; Alexandratos, Spiro; Paine, Robert; Hancock, Robert; Ensor, Dale

    2012-03-21

    One of the most challenging aspects of advanced processing of spent nuclear fuel is the need to isolate transuranium elements from fission product lanthanides. This project expanded the scope of earlier investigations of americium (Am) partitioning from the lanthanides with the synthesis of new separations materials and a centralized focus on radiochemical characterization of the separation systems that could be developed based on these new materials. The primary objective of this program was to explore alternative materials for actinide separations and to link the design of new reagents for actinide separations to characterizations based on actinide chemistry. In the predominant trivalent oxidation state, the chemistry of lanthanides overlaps substantially with that of the trivalent actinides and their mutual separation is quite challenging.

  15. Intermediate energy nuclear fission

    Nuclear fission has been investigated with the double-kinetic-energy method using silicon surface barrier detectors. Fragment energy correlation measurements have been made for U, Th and Bi with bremsstrahlung of 600 MeV maximum energy. Distributions of kinetic energy as a function of fragment mass are presented. The results are compared with earlier photofission data and in the case of bismuth, with calculations based on the liquid drop model. The binary fission process in U, Yb, Tb, Ce, La, Sb, Ag and Y induced by 600 MeV protons has been investigated yielding fission cross sections, fragment kinetic energies, angular correlations and mass distributions. Fission-spallation competition calculations are used to deduce values of macroscopic fission barrier heights and nuclear level density parameter values at deformations corresponding to the saddle point shapes. We find macroscopic fission barriers lower than those predicted by macroscopic theories. No indication is found of the Businaro Gallone limit expected to occur somewhere in the mass range A = 100 to A = 140. For Ce and La asymmetric mass distributions similar to those in the actinide region are found. A method is described for the analysis of angular correlations between complementary fission products. The description is mainly concerned with fission induced by medium-energy protons but is applicable also to other projectiles and energies. It is shown that the momentum and excitation energy distributions of cascade residuals leading to fission can be extracted. (Author)

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

    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

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

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

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

    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

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

    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)

  20. Products of fission, fusion and deep inelastic reactions

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

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

    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.)

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

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

  3. Fission product filter for hot reactor cooling gas

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

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

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

  5. Results of recent ORNL fission product release tests

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

  6. Mo-99 production by fission and future projections

    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)

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

    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

  8. A model for fission-product calculations, 1

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

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

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

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

    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

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

    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

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

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

    2012-01-01

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

  13. Nuclear fission

    V.M. STRUTINSKY's semi-classical method is the most precise to determine the energy of the different states along the fission way. The double-humped fission barrier explains fission isomerism. V.M. STRUTINSKY's barrier explains the ''intermediate structure'' observed in the cross section under the threshold; it provides also the observed effect of ''vibrational resonances'' with an interpretation. Taking an asymmetry parameter in consideration, a triple-humped fission barrier seems to be essential now for the light actinides. There is still a microscopic fission barrier to be explained

  14. A Covariance Generation Methodology for Fission Product Yields

    Terranova N.

    2016-01-01

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

  15. A Covariance Generation Methodology for Fission Product Yields

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

    2016-03-01

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

  16. Reducing uncertainties for short lived cumulative fission product yields

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

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

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

    2010-12-01

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

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

    Chadwick, Mark B.

    2009-10-01

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

  19. Separating the Minor Actinides Through Advances in Selective Coordination Chemistry

    Lumetta, Gregg J.; Braley, Jenifer C.; Sinkov, Sergey I.; Carter, Jennifer C.

    2012-08-22

    This report describes work conducted at the Pacific Northwest National Laboratory (PNNL) in Fiscal Year (FY) 2012 under the auspices of the Sigma Team for Minor Actinide Separation, funded by the U.S. Department of Energy Office of Nuclear Energy. Researchers at PNNL and Argonne National Laboratory (ANL) are investigating a simplified solvent extraction system for providing a single-step process to separate the minor actinide elements from acidic high-level liquid waste (HLW), including separating the minor actinides from the lanthanide fission products.

  20. Ab initio modelling of the behaviour of point defects and fission products in nuclear fuel; Modelisation par le calcul ab initio du comportement des defauts ponctuels et des produits de fission dans le combustible nucleaire

    Freyss, M.; Dorado, B.; Durinck, J. [CEA Cadarache (DEN/DEC/SESC/LLCC), 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Combustibles

    2008-07-01

    The aim of this work is to determine precisely the mechanisms of formation and migration of defects and fission products as well as the associated energies. Examples on uranium dioxide UO{sub 2} (standard nuclear fuel) and on uranium carbide UC (potential fuel for new generation reactors) are given. The obtained results are discussed and compared with the experimental results carried out. The ab initio method used is the Projector Augmented-Wave (PAW) method based on the density functional theory. The particular electronic properties of actinides are especially studied because, on account of their 5f orbitals more or less localized around the nucleus, it is difficult to model the actinide compounds by the DFT method. In particular, the modelling of the exchange-correlation interaction of the 5f electrons of UO{sub 2} requires approximations (as GGA+U) beyond those more currently used in ab initio calculations (LDA or GGA). (O.M.)

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

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

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

    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

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

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

  4. Measurement of fission product gases in the atmosphere

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

    1997-01-01

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

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

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

    2016-07-22

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

  6. Use of fast reactors for actinide transmutation

    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

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

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

  8. Rapid quantitation of uranium from mixed fission product samples

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

  9. Behaviour and retention of fission products in the containment

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

  10. Fission-product release from irradiated LWR fuel

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

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

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

  12. Quantitative analysis of fission products by γ spectrography

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

  13. Recoil release of fission products from nuclear fuel

    Wise, C.

    1985-10-01

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

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

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

  15. Advancing the scientific basis of trivalent actinide-lanthanide separations

    For advanced fuel cycles designed to support transmutation of transplutonium actinides, several options have been demonstrated for process-scale aqueous separations for U, Np, Pu management and for partitioning of trivalent actinides and fission product lanthanides away from other fission products. The more difficult mutual separation of Am/Cm from La-Tb remains the subject of considerable fundamental and applied research. The chemical separations literature teaches that the most productive alternatives to pursue are those based on ligand donor atoms less electronegative than O, specifically N- and S-containing complexants and chloride ion (Cl-). These 'soft-donor' atoms have exhibited usable selectivity in their bonding interactions with trivalent actinides relative to lanthanides. In this report, selected features of soft donor reagent design, characterization and application development will be discussed. The roles of thiocyanate, aminopoly-carboxylic acids and lactate in separation processes are detailed. (authors)

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

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

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

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

  18. Irradiation effects upon activities of fission product iodine

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

  19. Forced decontamination of fission products deposited on urban areas

    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)

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

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

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

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

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

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

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

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

    1994-03-01

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

  4. Determination of 90Sr in uranium fission products

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

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

    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)

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

    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

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

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

  8. Fission product release from fuel under LWR accident conditions

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

  9. Neutron cross section calculations for fission-product nuclei

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

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

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

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

    Herrmann, Steven Douglas

    2014-05-27

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

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

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

  13. A new hybrid surrogate ratio method for neutron-induced fission cross section measurements of short-lived actinides

    We will present a brief review of various surrogate methods employed for compound nuclear cross-section measurements along with our recent results using the hybrid surrogate ratio approach for determination of neutron induced fission cross sections of 233Pa and 234Pa isotopes

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

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

  15. Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

    One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

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

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

  17. Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

    Barber, Duncan Henry

    benchmark calculation demonstrates the improvement in agreement of the total inventory of those chemical elements included in the RMC fuel model to an ORIGEN-S calculation. ORIGEN-S is the Oak Ridge isotope generation and depletion computer program. The Gibbs energy minimizer requires a chemical database containing coefficients from which the Gibbs energy of pure compounds, gas and liquid mixtures, and solid solutions can be calculated. The RMC model of irradiated uranium dioxide fuel has been converted into the required format. The Gibbs energy minimizer has been incorporated into a new model of fission-product vaporization from the fuel surface. Calculated release fractions using the new code have been compared to results calculated with SOURCE IST 2.0P11 and to results of tests used in the validation of SOURCE 2.0. The new code shows improvements in agreement with experimental releases for a number of nuclides. Of particular significance is the better agreement between experimental and calculated release fractions for 140La. The improved agreement reflects the inclusion in the RMC model of the solubility of lanthanum (III) oxide (La2O3) in the fuel matrix. Calculated lanthanide release fractions from earlier computer programs were a challenge to environmental qualification analysis of equipment for some accident scenarios. The new prototype computer program would alleviate this concern. Keywords: Nuclear Engineering; Material Science; Thermodynamics; Radioactive Material, Gibbs Energy Minimization, Actinide Generation and Depletion, FissionProduct Generation and Depletion.

  18. RAPID QUANTITATION OF URANIUM FROM MIXED FISSION PRODUCT SAMPLES

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

    2016-03-09

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

  19. Kinetic studies on the removal of fission products from molten salt using Zeolite-4A. Contributed Paper RD-15

    Molten salt electrorefining process is one of the nonaqueous processes, being developed for reprocessing metallic spent fuel. This process uses liquid metals and molten salts and is operated at elevated temperatures. In the electro-refining process, the spent fuel is used as the anode of the electro-refiner and the actinide elements in the spent fuel are electrotransported from the anode through the molten salt electrolyte onto a suitable cathode where they are collected as metals in pure form. After some batches are processed, chlorides of fission products such as alkali, alkaline earth and rare earth metals accumulate in the electrolyte salt. The accumulated FPs in the salt will be removed by adsorption/ion-exchange by using zeolite columns. Hence, kinetic studies on the adsorption of Cs, Ba which are some of the major FP products in LiCI-KCI eutectic, have been carried out

  20. Sequential separation of actinides and lanthanides by extraction chromatography using a CMPO-TBP/XAD7 column

    CMPO/TBP sorbed on Amberlite XAD7 resin was used for the separation of actinides and lanthanides from nitric acid solutions by extraction chromatography. The distribution ratios of actinides and lanthanide fission products (Ce, Eu) as a function of acid concentration and some complexing agents were determined. In strong HNO3 medium (> 1 mol/l) the tri-, tetra- and hexavalent actinides as well as the lanthanides have shown great affinity for the CMPO/TBP/XAD7 sorbent. The same behavior was found in HCl medium except for trivalent actinides and lanthanides which show lower distribution values in the same acid range. The effect of some complexing agents as DTPA and ammonium oxalate were also investigated. In DTPA only hexavalent actinides showed higher distribution value. On the basis of these differences, an alternative procedure for actinide-lanthanide separation and actinides from each other is proposed. (author)

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

    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.

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

    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

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

    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)

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

    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)

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

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

    2016-05-01

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

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

    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.

  7. Coordination chemistry for new actinide separation processes

    The amount of wastes and the number of chemical steps can be decreased by replacing the PUREX process extractant (TBP) by, N.N- dialkylamides (RCONR'2). Large amounts of deep underground storable wastes can be stored into sub-surface disposals if the long lived actinide isotopes are removed. Spent nuclear fuels reprocessing including the partitioning of the minor actinides Np, Am, Cm and their transmutation into short half lives fission products is appealing to the public who is not favorable to the deep underground storage of large amounts of long half lived actinide isotopes. In this paper coordination chemistry problems related to improved chemical separations by solvent extraction are presented. 2 tabs.; 4 refs

  8. Partitioning of fission products from irradiated nitride fuel using inductive vaporization

    Irradiated nitride fuel (Pu0.3Zr0.7)N fabricated at PSI in frame of the CONFIRM project and having a burn-up of 10.4 % FIMA (Fission per Initial Metal Atom) has been investigated by means of inductive vaporization. The study of thermal stability and release behavior of Pu, Am, Zr and fission products (FPs) was performed in a wide temperature range (up to 2300 C. degrees) and on different redox conditions. On-line monitoring by ICP-MS detected low nitride stability and significant loss of Pu and Am at T>1900 C. degrees during annealing under inert atmosphere (Ar). The oxidative pre-treatment of nitride fuel on air at 1000 C. degrees resulted in strong retention of Pu and Am in the solid, as well as of most FPs. Thermodynamic modelling of elemental speciation using GEM-Selektor v.3 code (Gibbs Energy Minimization Selektor), supported by a comprehensive literature review on thermodynamics of actinides and FPs, revealed a number of binary compounds of Cs, Mo, Te, Sr and Ba to occur in the solid. Speciation of some FPs in the fuel is discussed and compared to earlier results of electron probe microanalysis (EPMA). Predominant vapor species predicted by GEM-Selektor calculations were Pu(g), Am(g) and N2. Nitrogen can be completely released from the fuel after complete oxidation at 1000 C. degrees. With regard to the irradiated nitride reprocessing technology, this result can have an important practical application as an alternative way for 15N recovery. (authors)

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

    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

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

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

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

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

    2007-02-28

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

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

    The charge distributions of fission products in proton-induced fission of 238U and 232Th were measured in a wide mass range. The most probable charges lay on the proton-rich side in the light fragment region and on the proton-deficient side in the heavy one compared with the unchanged charge distribution hypothesis. This result implies that the charge polarization occurs in the fission process. The charge polarization was examined with respect to the ground-state Q values. The estimations by the Q values fairly well reproduced the experimental most probable charges. These results suggest that the fission path to the most favorable charge division may go through the most energetically favorable path at scission point. (author)

  13. Composition of high fission product wastes resulting from future reprocessing of commercial nuclear fuels

    Swanson, J.L

    1986-07-01

    Pacific Northwest Laboratory studies, aimed at defining appropriate glass compositions for future disposal of high-level wastes, have developed composition ranges for the waste that will likely result during reprocessing of Light Water Reactor (LWR) and Liquid Metal Reactor (LMR) fuels. The purpose of these studies was to provide baseline waste characterizations for possible future commercial high-level waste so that waste immobilization technologies (e.g., vitrification) can be studied. Ranges in waste composition are emphasized because the waste will vary with time as different fuels are reprocesses, because choice of process chemicals is nuclear, and because fuel burnups will vary. Consequently, composition ranges are based on trends in fuel reprocessing procedures and on achievable burnups in operating reactors. In addition to the fission product and actinide elements, which are the primary hazardous materials in the waste, likely composition ranges are given for inert elements that may be present in the waste. These other elements may be present because of being present in the fuel, because of being added as process chemical during reprocessing, because of being added during equipment decontamination, or because of corrosion of plant equipment and/or fuel element cladding. This report includes a discussion of the chemicals added in variation of the PUREX process, which is likely to remain the favored reprocessing technique for commercial nuclear fuels. Consideration is also given to a pyrochemical process proposed for the reprocessing of some LMR fuels.

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

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

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

    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

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

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

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

    Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 2% at 14000C to >50% at 20000C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag

  18. Transient fission product release during reactor shutdown and startup

    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)

  19. Fission product source terms and engineered safety features

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

  20. Behavior of Nb fission product during nuclear fuel reprocessing

    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

  1. Spectroscopy of neutron rich nuclei using cold neutron induced fission of actinide targets at the ILL: the EXILL campaign

    de France G.; Blanc A.; Drouet F.; Jentschel M.; Köster U.; Mutti P.; Régis J.M.; Simpson G.; Soldner T.; Stezowski O.; Ur C.A.; Urban W.; Vancrayenest A.

    2014-01-01

    A combination of germanium detectors has been installed at the PF1B neutron guide of the ILL to perform the prompt spectroscopy of neutron-rich nuclei produced in the neutron-capture induced-fission of 235U and 241Pu. In addition LaBr3 detectors from the FATIMA collaboration have been installed in complement with the EXOGAM clovers to measure lifetimes of low-lying excited states. The measured characteristics and online spectra indicate very good performances of the overall setup.

  2. Spectroscopy of neutron rich nuclei using cold neutron induced fission of actinide targets at the ILL: the EXILL campaign

    de France G.

    2014-03-01

    Full Text Available A combination of germanium detectors has been installed at the PF1B neutron guide of the ILL to perform the prompt spectroscopy of neutron-rich nuclei produced in the neutron-capture induced-fission of 235U and 241Pu. In addition LaBr3 detectors from the FATIMA collaboration have been installed in complement with the EXOGAM clovers to measure lifetimes of low-lying excited states. The measured characteristics and online spectra indicate very good performances of the overall setup.

  3. Assessing the role of the (n, γ f) process in the low-energy fission of actinides

    Talou, Patrick; Lynn, J. E.; Kawano, T.; Mosby, S.; Couture, A.; Bouland, O.

    2016-06-01

    We review the role of the (n, γ f) process in the low-energy neutron-induced fission reaction of 239Pu. Recent measurements of the average total γ-ray energy released in this reaction were performed with the Detector for Advanced Neutron Capture Experiments (DANCE) at Los Alamos. Significant fluctuations of this quantity in the resonance region below 100 eV can be interpreted by invoking the presence of the indirect (n, γ f) process. Modern calculations of the probability for such an event to occur are presented.

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

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

  5. Catalytic electrolytic extraction of long-lived fission products

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

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

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

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

    Keshaw, Jeetesh Bhana

    2007-01-01

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

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

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

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

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

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

  11. Comparative assessment of the transmutation efficiency of plutonium and minor actinides in fusion/fission hybrids and ads

    A preliminary comparative assessment relevant to the transmutation efficiency of plutonium and minor actinides has been performed in the case of ANSALDO's Energy Amplifier Demonstration Facility based on molten lead-bismuth eutectic cooling, classical MOX-fuel technology and operating at 80 MWth. The neutronic calculations presented in this paper are a result of a state-of-the-art computer code package, EA-MC, developed by C. Rubbia and his group at CERN. Both high-energy particle interactions and low-energy neutron transport are treated with a sophisticated method based on a full Monte Carlo simulation, together with modern nuclear data libraries. Detailed Monte Carlo transport calculations were performed for different types of external neutron sources: D-D and D-T fusion sources and proton induced spallation neutron sources. The fuel core was described on a pin-by- pin basis allowing for detailed scans of the main neutronic properties, e.g. neutron flux spectra and power density distributions. (author)

  12. Immobilization of fission products in phosphate ceramic waste forms

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

  13. Application of inorganic exchangers in fission product separation

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

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

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

  15. Airborne measurements of fission product fall-out

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

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

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

  17. Immobilization of fission products in phosphate ceramic waste forms

    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.

  18. Fission Product Release from Spent Nuclear Fuel During Melting

    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

  19. Metabolism of fission products in man: Marshallese experience

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

  20. Fission Product Fast Reactor Constants System of JNDC

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

  1. Production of actinide nuclei by multi-nucleon transfer

    Lauritsen, T.; Ahmad, I.; Carpenter, M.P. [and others

    1995-08-01

    Multi-nucleon transfers have increasingly allowed us to reach parts of the nuclear chart where regular compound nuclear reactions are prohibited. The interesting region of Ra and Rn, where a rich tapestry of nuclear structure manifests itself, is now accessible using this technique of deep inelastic scattering. In particular, these nuclei are predicted to lie at the onset of octupole deformation and the region is rich in examples of shape coexistence. There are several theoretical predictions of nuclear structure of these nuclei that have not been experimentally tested. Moreover, there is serious disagreement among these theories. We used a beam of {sup 136}Xe at 720 MeV from ATLAS on a target of {sup 232}Th to produce a range of Rn isotopes, with a mass from 220 to 224, and Ra isotopes with masses greater than 222. The beam energy, target and beam were selected carefully to enhance the cross-section for production of these nuclei and reduce the Doppler broadening of the gamma rays that were observed in the Argonne Notre Dame gamma-ray facility. The 12 germanium detectors of this array allowed the observation of gamma-gamma coincidences. The inner ball of 50 BGO detectors allowed us to record the multiplicity and sum-energy information for each event. The latter should permit us to determine the entry region in the products of the transfer reaction. We had four successful days of beam-time, when we collected in excess of 8 x 10{sup 7} events. Data analysis is in progress at the University of Liverpool. A complete set of spectroscopic information on the yrast structure of the many nuclei produced in this reaction is being extracted.

  2. Project research on nuclear physical and chemical characteristics of actinide nuclides

    The chemical and nuclear physical characteristics of actinide elements have been investigated using the experimental methods and instruments of this laboratory. This laboratory has a facility in which the transuranium elements (TRU) and the long-lived fission products (LLFP) can be dealt with. The utility of this facility has been expected. The investigation on the actinide elements and its fission products have been carried out as a project research from both view points of science and technology. The research reports during three years (2005-07) are described here. (M.H.)

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

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

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

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

    2013-01-01

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

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

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

  6. Gas core reactors for actinide transmutation. [uranium hexafluoride

    Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

    1979-01-01

    The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

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

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

  8. Effects of radiation and fission product incorporation in a yttria-stabilized zirconia based inert matrix fuel

    Zhu, Sha

    This work has investigated the irradiation and incorporation effects of fission products in a yttria-stabilized zirconia (YSZ) based inert matrix fuel (IMF). The concept of inert matrix fuel is based on a new strategy for disposition of plutonium generated from the reprocessing of commercial nuclear fuel and the dismantling of nuclear weapons, i.e. using uranium-free oxides to "burn" plutonium and other actinides (Np, Cm, and Am) in reactors. This approach allows direct disposal, without reprocessing, after once-through burn-up. YSZ and MgAl2O4-YSZ composites are among the potential ceramics for IMF due to their high chemical durability and radiation resistance. The research involved investigating the production, nature, and accumulation of irradiation-induced defects, the behavior of the fission products in the ceramics, the structural stability and amorphization resistance of the YSZ during implantation. Ion implantations were conducted with 200--400 keV Cs+, Sr+, I+, Xe+ and Ti+ up to fluences of 1 x 1017/cm 2 at both room temperature and temperatures of 600--700°C. Thermal annealing was subsequently completed after room temperature ion implantations. In situ and ex situ transmission electron microscopy (TEM), optical absorption spectroscopy, photo-luminescence spectroscopy, and electron paramagnetic resonance (EPR) spectroscopy were employed to characterize the irradiation induced defect evolution and analyze the defect structures. Various irradiation effects were observed and determined in the experiments, such as point defects (F type and V type color centers), defect clusters (dislocation loops), cavities (voids and bubbles), the crystalline-to-amorphous transition, and the phase transformation from fluorite to pyrochlore structure. The ion irradiation-induced amorphization mechanism, the retention ability of the fission products, and structural stability of YSZ are discussed in terms of ion incorporation effects, implanted ion radii, and the solubility

  9. Uncertainty analysis on fission Mo production in HANARO

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

  10. Actinide production in the reaction of heavy ions with curium-248

    Chemical experiments were performed to examine the usefulness of heavy ion transfer reactions in producing new, neutron-rich actinide nuclides. A general quasi-elastic to deep-inelastic mechanism is proposed, and the utility of this method as opposed to other methods (e.g. complete fusion) is discussed. The relative merits of various techniques of actinide target synthesis are discussed. A description is given of a target system designed to remove the large amounts of heat generated by the passage of a heavy ion beam through matter, thereby maximizing the beam intensity which can be safely used in an experiment. Also described is a general separation scheme for the actinide elements from protactinium (Z=91) to mendelevium (Z=101), and fast specific procedures for plutonium, americium and berkelium. The cross sections for the production of several nuclides from the bombardment of 248Cm with 18O, 86Kr and 136Xe projectiles at several energies near and below the Coulomb barrier were determined. The results are compared with yields from 48Ca and 238U bombardments of 248Cm. Simple extrapolation of the product yields into unknown regions of charge and mass indicates that the use of heavy ion transfer reactions to produce new, neutron-rich above-target species is limited. The substantial production of neutron-rich below-target species, however, indicates that with very heavy ions like 136Xe and 238U the new species 248Am, 249Am and 247Pu should be produced with large cross sections from a 248Cm target. A preliminary, unsuccessful attempt to isolate 247Pu is outlined. The failure is probably due to the half life of the decay, which is calculated to be less than 3 minutes. The absolute gamma ray intensities from 251Bk decay, necessary for calculating the 251Bk cross section, are also determined

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

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

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

    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)

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

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

  14. High temperature behaviour of fission products in irradiated ZrO2 and MgAl2O4

    Zirconia and spinel are promising inert matrices for the incineration of Pu and other actinides in nuclear reactors. In the scenario of direct storage of the burnt fuel in deep geological repositories, the study of the diffusion and retention of fission products in the matrix is of prime interest. Specific fission products (Cs and I) were introduced into ZrO2 and MgAl2O4 single crystals by ion implantation. The samples were then either annealed or irradiated with Ar ions at temperatures ranging from 200 deg C up to 1 000 deg C. The structural properties of the host matrix, the diffusion and release of implanted species due to thermal annealing or high-temperature irradiation were investigated by RBS spectrometry. No change of the depth profile of implanted ions occurs up to 800 deg C upon annealing. Conversely, a clear increase of the profile and a strong loss of implanted ions are observed at 650 deg C during irradiation with Ar ions. This radiation-enhanced diffusion effect has to be taken into account for future applications. (authors)

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

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

    2013-03-01

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

  16. Extraction of actinides and lanthanides by calixarenes CMPO. Possibility to separate actinides from lanthanides (Calixpart project)

    The CALIXPART project accepted by the European Community within the framework of the 5 PCRD, relates to the 'selective extraction of minor actinides from H.A. liquid waste by organized matrices'. The objective of this new project is the selective extraction in only one step of minor actinides from a solution of fission products including lanthanides. This separation will be investigated through two strategies: - In the first one, macrocycles will be grafted with ligands containing nitrogen or sulphur which are able to discriminate actinides from lanthanides, but generally present very low distribution coefficients in strongly acidic solutions. Following the example of calixarenes CMPO, the grafting of these ligands on macrocyclic supports should increase the distribution coefficients, and thus allow to use these extractants at nitric acid concentrations up to 3 M. The nitrogen or sulphur ligands are not necessarily selective with respect to the other fission products, and the macrocyclic structure should also afford this necessary selectivity if one wishes to operate in a single step. Once americium and curium separated, the difference in size between both cations is undoubtedly sufficient to make it possible to separate them at the stripping stage. - The second strategy considered is the introduction of two types of ligands (hard and soft) on a macrocyclic structure, the first ensuring the extraction of lanthanides and trivalent actinides, the seconds bringing discrimination between these two groups of cations. (author)

  17. 1: Mass asymmetric fission barriers for 98Mo; 2: Synthesis and characterization of actinide-specific chelating

    Excitation functions have been measured for complex fragment emission from the compound nucleus 98Mo, produced by the reaction of 86Kr with 12C. Mass asymmetric fission barriers have been obtained by fitting the excitation functions with a transition state formalism. The extracted barriers are ∼ 5.7 MeV higher, on average, than the calculations of the Rotating Finite Range Model (RFRM). These data clearly show an isospin dependence of the conditional barriers when compared with the extracted barriers from 90Mo and 94Mo. Eleven different liquid/liquid extractants were synthesized based upon the chelating moieties 3,2-HOPO and 3,4-HOPO; additionally, two liquid/liquid extractants based upon the 1,2-HOPO chelating moiety were obtained for extraction studies. The Pu(IV) extractions, quite surprisingly, yielded results that were very different from the Fe(III) extractions. The first trend remained the same: the 1,2-HOPOs were the best extractants, followed closely by the 3,2-HOPOs, followed by the 3,4-HOPOs; but in these Pu(IV) extractions the 3,4-HOPOs performed much better than in the Fe(III) extractions. 129 refs

  18. 1: Mass asymmetric fission barriers for {sup 98}Mo; 2: Synthesis and characterization of actinide-specific chelating agents

    Veeck, A.C. [Univ. of California, Berkeley, CA (United States). Dept. of Chemistry]|[Lawrence Livermore National Lab., CA (United States). Glenn T. Seaborg Inst. for Transactinium Science]|[Lawrence Berkeley National Lab., CA (United States). Nuclear Science Div.

    1996-08-01

    Excitation functions have been measured for complex fragment emission from the compound nucleus {sup 98}Mo, produced by the reaction of {sup 86}Kr with {sup 12}C. Mass asymmetric fission barriers have been obtained by fitting the excitation functions with a transition state formalism. The extracted barriers are {approximately} 5.7 MeV higher, on average, than the calculations of the Rotating Finite Range Model (RFRM). These data clearly show an isospin dependence of the conditional barriers when compared with the extracted barriers from {sup 90}Mo and {sup 94}Mo. Eleven different liquid/liquid extractants were synthesized based upon the chelating moieties 3,2-HOPO and 3,4-HOPO; additionally, two liquid/liquid extractants based upon the 1,2-HOPO chelating moiety were obtained for extraction studies. The Pu(IV) extractions, quite surprisingly, yielded results that were very different from the Fe(III) extractions. The first trend remained the same: the 1,2-HOPOs were the best extractants, followed closely by the 3,2-HOPOs, followed by the 3,4-HOPOs; but in these Pu(IV) extractions the 3,4-HOPOs performed much better than in the Fe(III) extractions. 129 refs.

  19. Adaptation of ICP-AES in lead cell facility in Chemistry Group, IGCAR and analysis of simulated high level waste as a part of the studies on minor actinide partitioning

    The spent fuel discharged from the nuclear reactor contains unused uranium and plutonium, and Np, Am, Cm called as minor actinides and fission products. Spent fuel is dissolved in nitric acid. U and Pu are recovered by a solvent extraction process known as PUREX process using 1.1 M TBP as extractant. The raffinate rejected is known as High Level Liquid Waste which is a complex mixture of minor actinides, corrosion products, and fission products. Partitioning of minor actinides (MA) and its transmutation is a viable strategy for the safe management of high level liquid waste (HLLW)

  20. Behavior of fission products in sulfide reprocessing process

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

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

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

  2. Diffusion of Fission Product Elements in Compacted Bentonite

    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)

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

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

  4. A review of libraries of fission product yields

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

  5. A model for fission product distribution in CANDU fuel

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

  6. Review of fission product plateout investigations at General Atomic

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

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

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

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

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

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

    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

  10. Separation of actinides from spent nuclear fuel: A review.

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials. PMID:27427893

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

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

  12. The Phebus Fission Product and Source Term International Programmes

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

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

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

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

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

  15. Burning of actinides: A complementary waste management option?

    The TRU actinide are building up at a rate of about 90 tHM per year. Approximately 45 tHM will remain occluded in the spent fuel structures, leaving about 45 tHM available; 92% as recycled plutonium and 8% as minor actinides (neptunium, americium, curium) immobilized in vitrified waste. There is renewed interest in partitioning and transmutation (P and T), largely because of difficulties encountered throughout the world in finding suitable geologic formations in locations which are acceptable to the public. In 1988, the Japanese Atomic Energy Commission launched a very important and comprehensive R and D program. The general strategy of introducing Partitioning and Transmutation (P and T) as an alternative waste management option is based on the radiological benefit which is expected from such a venture. The selection of the actinides and long-lived fission products which are beneficial to eliminate by transmutation depends upon a number of technical factors, including hazard and decontamination factors, and the effect of geological confinement. There are two ways to approach the separation of minor actinides and long-lived fission products from reprocessing streams: by modifying the current processes in order to reroute the critical nuclides into a single solution, for example high-level liquid waste, and use this as a source for partitioning processes; and by extension of the conventional PUREX process to all minor actinides and long-lived fission products in second generation reprocessing plants. Prior to the implementation of one of these schemes, it seems obvious to improve the separation yield of plutonium from HLW within the presently running plants. Actinide P and T is not an alternative long-term waste management option. Rather, it is a complementary technique to geologic disposal capable of further decreasing the radiological impact of the fuel cycle over the very long term. 1 tab

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

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

  17. Selective extraction of actinides from high level liquid wastes. Study of the possibilities offered by the Redox properties of actinides

    Partitioning of high level liquid wastes coming from nuclear fuel reprocessing by the PUREX process, consists in the elimination of minor actinides (Np, Am, and traces of Pu and U). Among the possible processes, the selective extraction of actinides with oxidation states higher than three is studied. First part of this work deals with a preliminary step; the elimination of the ruthenium from fission products solutions using the electrovolatilization of the RuO4 compound. The second part of this work concerns the complexation and oxidation reactions of the elements U, Np, Pu and Am in presence of a compound belonging to the insaturated polyanions family: the potassium phosphotungstate. For actinide ions with oxidation state (IV) complexed with phosphotungstate anion the extraction mechanism by dioctylamine was studied and the use of a chromatographic extraction technic permitted successful separations between tetravalents actinides and trivalents actinides. Finally, in accordance with the obtained results, the basis of a separation scheme for the management of fission products solutions is proposed

  18. Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity [O1] - Fundamentals of P3 Mechanism and Methodology Development for Plutonium Categorization

    'Protected Plutonium Production (P3)' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MAs). Doping the small amount of MAs such as 237Np or 241Am with large neutron capture cross-section into the uranium fuel to enhance the production of 238Pu or 242Pu, which have high spontaneous fission neutron source or also high decay heat to makes the process of the nuclear weapon manufacture and maintenance technologically difficult, can be effective for improving the isotopic barrier of proliferation resistance of the plutonium in thermal reactors. Super weapon grade plutonium could be produced in the blanket of a conventional FBR. However, by increasing the 238Pu or 242Pu ratio in the total plutonium by MAs doping into the fresh blanket, the protected plutonium with high proliferation-resistance can be bred. A new evaluation function, 'attractiveness', defined as a ratio of potential of fission yield to the technological difficulties of nuclear explosive device, has been proposed to evaluate the proliferation resistance of Pu based on the nuclear material property for Plutonium Categorization. In the conference, the fundamentals of P3 mechanism by transmutation of MA, and the comparison of the 'attractiveness' of the Pu produced in advanced reactors based on P3 mechanism and in the conventional reactors will be presented. Instead of the geological disposal or just their burning of MAs by the fission reaction, they should be treated as valuable fertile materials to enhance the proliferation resistance of plutonium produced in the thermal and fast breeder reactors for peace and sustainable prosperity in future. Acknowledgement: Some parts of this work have been supported by the Ministry of Education, Culture, Sports, Science and Technology in Japan. (authors)

  19. The selective extraction of oxidized minor actinides: a possible route for the Actinex program

    In the SPIN programme, defined by CEA to improve the management of high level nuclear waste, a part called ACTINEX is specially devoted to the extraction of long-lived alpha emitters and fission products from high level liquid waste issuing from the PUREX process. Concerning the actinides elements, as U and Pu are already recovered, the main objective to reach is now the quantitative extraction of Np and Am. The transmutation of these recovered actinides into short-lived radionuclides is then forecast. This paper deals with the possibilities to define a minor actinides partitioning process based on the selective extraction of actinides oxidized to their oxidation states higher than three. It essentially focuses on americium chemistry. Finally, two general separation scheme for minor actinide partitioning are proposed and discussed. (authors). 5 figs., 13 refs

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

    Rubchenya V. A.

    2013-12-01

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