WorldWideScience

Sample records for actinides fission products

  1. Actinide and fission product separation and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-07-01

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

  2. Actinide and fission product partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

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

  3. Actinide and fission product partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

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

  4. Actinide and fission product separation and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-07-01

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

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

    Science.gov (United States)

    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.

  6. Radioanalytical determination of actinides and fission products in Belarus soils.

    Science.gov (United States)

    Michel, H; Gasparro, J; Barci-Funel, G; Dalmasso, J; Ardisson, G; Sharovarov, G

    1999-04-01

    Alpha emitting actinides such as plutonium, americium or curium were measured by alpha-spectrometry after radiochemical separation. The short range of alpha-particles within matter requires, after a pre-concentration process, a succession of isolation and purification steps based on the valence states modification of the researched elements. For counting, actinides were electrodeposited in view to obtain the mass-less source necessary to avoid self-absorption of the emitted radiations. Activity concentrations of gamma-emitting fission products were calculated after measurement with high purity germanium detectors (HPGe). These different methods were used to analyse soils sampled in the Republic of Belarus, not far from the Chernobyl nuclear plant.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Salahuddin Asif

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-01

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

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

    CERN Document Server

    Sors, M

    1999-01-01

    Research and development work is currently in progress in France on the feasibility of transmutation of very long-lived radionuclides such as americium, blended with an inert medium such as magnesium oxide and pelletized for irradiation in a fast neutron reactor. The process is primarily designed to produce ceramics for nuclear reactors, but could also be used to produce targets for accelerators. The Actinide Development Laboratory is part of the ATALANTE complex at Marcoule, where the CEA investigates reprocessing, liquid and solid waste treatment and vitrification processes. The laboratory produces radioactive sources; after use, their constituents are recycled, notably through R and D programs requiring such materials. Recovered americium is purified, characterized and transformed for an experiment known as ECRIX, designed to demonstrate the feasibility of fabricating americium-based ceramics and to determine the reactor transmutation coefficients.

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

    CERN Document Server

    Herrera-Martínez, Adonai

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

  12. Application of ionic liquids in actinide and fission product separations: progress and prospects.

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, D. C.; Young, B. A.; Jensen, M. P.; Rickert, P. G.; Dzielawa, J. A.; Dilger, A. A.; Rausch, D. J.; Dietz, M. L.; Chemistry

    2006-01-01

    Ionic liquids (ILs), particularly those that are liquid at room temperature, have attracted intense interest as alternatives to conventional organic solvents in a host of synthetic, catalytic, and electrochemical applications. Recently, growing attention has been devoted to their use in separations, typically as replacements for the organic diluents employed in traditional liquid-liquid extraction or membrane-based separations of organic solutes or metal ions. Although studies of the extraction of metal ions into various ILs indicate that these solvents frequently provide extraction efficiencies far greater than those obtained with conventional solvents, other work suggests that they suffer from various drawbacks that could limit their utility as extraction solvents. In this chapter, we examine the viability of ionic liquids as the basis for extraction systems for the separation of actinides and fission products from acidic media and consider approaches by which their limitations may be overcome.

  13. Prompt fission neutron spectrum of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Capote, R. [International Atomic Energy Agency, Vienna (Austria); Chen, Y. -J. [China Institute of Atomic Energy, Beijing (China); Hambsch, F. J. [European Commission, Joint Research Centre - IRRM, Geel (Belgium); Jurado, B. [CENBG, CNRS/IN2P3, Gradignan (France); Kornilov, N. [Ohio Univ., Athens, OH (United States); Lestone, J. P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Litaize, O. [CEA, DEN, DER, SPRC, Saint-Paul-Lez-Durance (France); Morillon, B. [CEA, DAM, DIF, Arpajon (France); Neudecker, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Oberstedt, S. [European Commission, Joint Research Centre - IRRM, Geel (Belgium); Ohsawa, T. [Kinki Univ., Osaka-fu (Japan); Otuka, N. [International Atomic Energy Agency, Vienna (Austria); Pronyaev, V. G. [Institute of Physics and Power Engineering, Obninsk (Russian Federation); Saxena, A. [Bhabha Atomic Research Centre, Mumbai (India); Schmidt, K. H. [CENBG, CNRS/IN2P3, Gradignan (France); Serot, O. [CEA, DEN, DER, SPRC, Saint-Paul-Lez-Durance (France); Shcherbakov, O. A. [Petersburg Nuclear Physics Institute of NRC " Kurchatov Institute" , Gatchina (Russian Federation); Shu, N. -C. [China Institute of Atomic Energy, Beijing (China); Smith, D. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Talou, P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trkov, A. [International Atomic Energy Agency, Vienna (Austria); Tudora, A. C. [Univ. of Bucharest, Magurele (Romania); Vogt, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Univ. of California, Davis, CA (United States); Vorobyev, A. S. [Petersburg Nuclear Physics Institute of NRC " Kurchatov Institute" , Gatchina (Russian Federation)

    2016-01-06

    Here, the energy spectrum of prompt neutron 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

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

    Indian Academy of Sciences (India)

    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.

  16. Studies on Neutron, Photon (Bremsstrahlung and Proton Induced Fission of Actinides and Pre-Actinides

    Directory of Open Access Journals (Sweden)

    H. Naik

    2015-08-01

    Full Text Available We present the yields of various fission products determined in the reactor neutron, 3.7-18.1 MeV quasi-mono energetic neutron, 8-80 MeV bremsstrahlung and 20-45 MeV proton induced fission of 232Th and 238U using radiochemical and off-line beta or gamma ray counting. The yields of the fission products in the bremsstrahlung induced fission natPb and 209Bi with 50- 70 MeV and 2.5 GeV based on off-line gamma ray spectrometric technique were also presented. From the yields of fission products, the mass chains yields were obtained using charge distribution correction. From the mass yield distribution, the peak-to-valley (P/V ratio was obtained. The role of excitation energy on the peak-to-valley ratio and fine structure such as effect of shell closure proximity and even-odd effect of mass yield distribution were examined. The higher yields of the fission products around A=133-134, 138-140 and 143-144 and their complementary products explained from the nuclear structure effect and role of standard I and II mode of asymmetric fission. In the neutron, photon (bremsstrahlung and proton induced fission, the asymmetric mass distribution for actinides (Th, U and symmetric distribution for pre-actinides (Pb, Bi were explained from different type of potential fission barrier

  17. Fission cross section measurements for minor actinides

    Energy Technology Data Exchange (ETDEWEB)

    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)

  18. Fission cross section measurements for minor actinides

    Energy Technology Data Exchange (ETDEWEB)

    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)

  19. Fission of actinides using a table-top laser

    CERN Document Server

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    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.

  2. Prompt Fission Neutron Spectra of Actinides

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. Recovery and use of fission product noble metals

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

  4. Experimental approach to fission process of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Hiroshi [Osaka Univ., Toyonaka (Japan). Faculty of Science

    1997-07-01

    From experimental views, it seems likely that the mechanism of nuclear fission process remains unsolved even after the Bohr and Weeler`s study in 1939. Especially, it is marked in respect of mass distribution in unsymmetric nuclear fission. The energy dependency of mass distribution can be explained with an assumption of 2-mode nuclear fission. Further, it was demonstrated that the symmetrical fission components and the unsymmetrical ones have different saddle and fission points. Thus, the presence of the 2-mode fission mechanism was confirmed. Here, transition in the nuclear fission mechanism and its cause were investigated here. As the cause of such transition, plausible four causes; a contribution of multiple-chance fission, disappearance of shell effects, beginning of fission following collective excitation due to GDR and nuclear phase transition were examined in the condition of excitation energy of 14.0 MeV. And it was suggested that the transition in the nuclear fission concerned might be related to phase transition. In addition, the mechanism of nuclear fission at a low energy and multi-mode hypothesis were examined by determination of the energy for thermal neutron fission ({sup 233,235}U and {sup 239}Pu) and spontaneous nuclear fission ({sup 252}Cf). (M.N.)

  5. Fission fragment angular distributions in pre-actinide nuclei

    Science.gov (United States)

    Banerjee, Tathagata; Nath, S.; Jhingan, A.; Kaur, Gurpreet; Dubey, R.; Yadav, Abhishek; Laveen, P. V.; Shamlath, A.; Shareef, M.; Gehlot, J.; Saneesh, N.; Prasad, E.; Sugathan, P.; Pal, Santanu

    2016-10-01

    Background: Complete fusion of two nuclei leading to formation of a heavy compound nucleus (CN) is known to be hindered by various fission-like processes, in which the composite system reseparates after capture of the target and the projectile inside the potential barrier. As a consequence of these non-CN fission (NCNF) processes, fusion probability (PCN) starts deviating from unity. Despite substantial progress in understanding, the onset and the experimental signatures of NCNF and the degree of its influence on fusion have not yet been unambiguously identified. Purpose: This work aims to investigate the presence of NCNF, if any, in pre-actinide nuclei by systematic study of fission angular anisotropies and fission cross sections (σfis) in a number of nuclear reactions carried out at and above the Coulomb barrier (VB) . Method: Fission fragment angular distributions were measured for six 28Si-induced reactions involving isotopically enriched targets of 169Tm,176Yb,175Lu,180Hf,181Ta, and 182W leading to probable formation of CN in the pre-actinide region, at a laboratory energy (Elab) range of 129-146 MeV. Measurements were performed with large angular coverage (θlab=41∘ -170∘) in which fission fragments (FFs) were detected by nine hybrid telescope (E -Δ E ) detectors. Extracted fission angular anisotropies and σfis were compared with statistical model (SM) predictions. Results: Barring two reactions involving targets with large non-zero ground state spin (J ) , viz., 175Lu(7/2+) and 181Ta(7/2+) , experimental fission angular anisotropies were found to be higher in comparison with predictions of the statistical saddle point model (SSPM), at Ec .m . near VB. Comparison of present results with those from neighboring systems revealed that experimental anisotropies increasingly deviated from SSPM predictions as one moved from pre-actinide to actinide nuclei. For reactions involving targets with large nonzero J , this deviation was subdued. Comparison between

  6. Measurements of Fission Cross Sections of Actinides

    CERN Multimedia

    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.

  7. Study of extraterrestrial disposal of radioactive wastes. Part 3: Preliminary feasibility screening study of space disposal of the actinide radioactive wastes with 1 percent and 0.1 percent fission product contamination

    Science.gov (United States)

    Hyland, R. E.; Wohl, M. L.; Finnegan, P. M.

    1973-01-01

    A preliminary study was conducted of the feasibility of space disposal of the actinide class of radioactive waste material. This waste was assumed to contain 1 and 0.1 percent residual fission products, since it may not be feasible to completely separate the actinides. The actinides are a small fraction of the total waste but they remain radioactive much longer than the other wastes and must be isolated from human encounter for tens of thousands of years. Results indicate that space disposal is promising but more study is required, particularly in the area of safety. The minimum cost of space transportation would increase the consumer electric utility bill by the order of 1 percent for earth escape and 3 percent for solar escape. The waste package in this phase of the study was designed for normal operating conditions only; the design of next phase of the study will include provisions for accident safety. The number of shuttle launches per year required to dispose of all U.S. generated actinide waste with 0.1 percent residual fission products varies between 3 and 15 in 1985 and between 25 and 110 by 2000. The lower values assume earth escape (solar orbit) and the higher values are for escape from the solar system.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  10. Energy-Dependent Fission Q Values Generalized for All Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Vogt, R

    2008-09-25

    We generalize Madland's parameterization of the energy release in fission to obtain the dependence of the fission Q values on incident neutron energy, E{sub n}, for all major and minor actinides. These Q(E{sub n}) parameterizations are included in the ENDL2008 release. This paper describes calculations of energy-dependent fission Q values based on parameterizations of the prompt energy release in fission [1], developed by Madland [1] to describe the prompt energy release in neutron-induced fission of {sup 235}U, {sup 238}U, and {sup 239}Pu. The energy release is then related to the energy deposited during fission so that experimentally measurable quantities can be used to obtain the Q values. A discussion of these specific parameterizations and their implementation in the processing code for Monte Carlo neutron transport, MCFGEN, [2] is described in Ref. [3]. We extend this model to describe Q(E) for all actinides, major and minor, in the Evaluated Nuclear Data Library (ENDL) 2008 release, ENDL2008.

  11. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    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.

  13. Fission of actinides through quasimolecular shapes

    Science.gov (United States)

    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.

  14. Validation of minor actinides fission neutron cross-sections

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2015-01-01

    Full Text Available Verification of neutron fission cross-sections of minor actinides from some recently available evaluated nuclear data libraries was carried out by comparison of the reaction rates calculated by the MCNP6.1 computer code to the experimental values. The experimental samples, containing thin layers of 235U, 237Np, 238,239,240,241Pu, 242mAm, 243Cm, 245Cm, and 247Cm, deposited on metal support and foils of 235U (pseudo-alloy 27Al + 235U, 238U, natIn, 64Zn, 27Al, and multi-component sample alloy 27Al + 55Mn + natCu + natLu + 197Au, were irradiated in the channels of the tank containing fluorine salts 0.52NaF + 0.48ZrF4, labelled as the Micromodel Salt Blanket, inserted in the lattice centre of the MAKET heavy water critical assembly at the Institute for Theoretical and Experimental Physics, Moscow. This paper is a continuation of earlier initiated scientific-research activities carried out for validation of the evaluated fission cross-sections of actinides that were supposed to be used for the quality examination of the fuel design of the accelerator driven systems or fast reactors, and consequently, determination of transmutation rates of actinides, and therefore, determination of operation parameters of these reactor facilities. These scientific-research activities were carried out within a frame of scientific projects supported by the International Science and Technology Center and the International Atomic Energy Agency co-ordinated research activities, from 1999 to 2010. Obtained results confirm that further research is needed in evaluations in order to establish better neutron cross-section data for the minor actinides and selected nuclides which could be used in the accelerator driven systems or fast reactors.

  15. Systematic Characteristics of Fast Neutron Fission Cross Sections for Actinide Nuclei

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The neutron fission cross sections of actinide nuclei are important data for the design of nuclear reactor and nuclear engineering, and so on. So far, there has been a certain amount of experimental data for the fission cross sections of actinide nuclei. However,

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    CERN Multimedia

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

    Directory of Open Access Journals (Sweden)

    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.

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

    NARCIS (Netherlands)

    Damen, P.M.G.

    2003-01-01

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

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

    NARCIS (Netherlands)

    Damen, P.M.G.

    2003-01-01

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

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

    Directory of Open Access Journals (Sweden)

    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.

  2. Studies on the properties of hard-spectrum, actinide fissioning reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, J.B.; Prichard, A.W.; Schofield, P.E.; Robinson, A.H.; Spinrad, B.I.

    1980-01-01

    It is technically feasible to construct an operable (e.g., safe and stable) reactor to burn waste actinides rapidly. The heart of the concept is a driver core of EBR-II type, with a central radial target zone in which fuel elements, made entirely of waste actinides are exposed. This target fuel undergoes fission, as a result of which actinides are rapidly destroyed. Although the same result could be achieved in more conventionally designed LWR or LMFBR systems, the fast spectrum reactor does a much more efficient job, by virtue of the fact that in both LWR and LMFBR reactors, actinide fission is preceded by several captures before a fissile nuclide is formed. In the fast spectrum reactor that is called ABR (actinide burning reactor), these neutron captures are short-circuited.

  3. Actinide incineration in fusion-fission hybrid-A model nuclear synergy

    Science.gov (United States)

    Taczanowski, Stefan

    2012-06-01

    The alliance of fusion with fission is a cause worthy of great efforts, as being able to ease (if not even to solve) serious problems that both these forms of nuclear energy are facing. Very high investment costs caused by tokamak enormous size, material consumption and difficult technology put in doubt whether alone the minute demand for fuel raw material (Li) and lack of danger of uncontrolled supercriticality prove sufficient for making it competitive. Preliminary evaluations demonstrated that a radical shift of energy production i.e. the energy gain from plasma to fission blanket is feasible [1]. A reduction in the fusion component to about 2% at given system power allows for a radical drop in plasma Q down to the values of ˜0.2-0.3 achievable in small systems [2] (e.g. mirrors) of sizes comparable to fission reactors. As a result in a Fusion-Driven Actinide Incinerator (FDI) both radiations from the plasma: corpuscular (i.e. neutrons and ions) and photons are drastically reduced. Thus are too, first of all - the neutron induced radiation damage: DPA and gas production, then plasma-wall interactions. The fundamental safety of the system has been proved by simulation of its collapse that has shown preservation its subcriticality. Summarizing, all the above problems may be solved with synergic union of fusion with fission embodied in the concept of FDI - small and less expensive.

  4. Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H. [Cogema, 78 - Saint Quentin en Yvelines (France); Guillou, E. [Cogema Etablissement de la Hague, D/SQ/SMT, 50 - Beaumont Hague (France); Cousinou, P. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, 92 (France); Barbry, F. [CEA Valduc, Inst. de Protection et de Surete Nucleaire, 21 - Is sur Tille (France); Grouiller, J.P.; Bignan, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2001-07-01

    Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis.

  5. Fission Product Library and Resource

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-29

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

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

    CERN Document Server

    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...

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

    CERN Document Server

    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 ...

  8. Yields of fission products from various uranium and thorium targets

    Energy Technology Data Exchange (ETDEWEB)

    Kronenberg, A.; Spejewski, E.H.; Mervin, B.; Jost, C.; Carter, H.K. [Oak Ridge Associated Universities, Oak Ridge, TN 37831 (United States); Stracener, D.W. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Greene, J.P. [Argonne National Laboratory, Argonne, IL 60439 (United States)], E-mail: greene@anl.gov; Nolen, J.A. [Argonne National Laboratory, Argonne, IL 60439 (United States); Talbert, W.L. [TechSource, Inc., Santa Fe, NM 87501 (United States)

    2008-10-15

    Yield measurements from proton-induced fission have been performed on a number of actinide targets, both Th and U, at the on-line test facility at Oak Ridge National Laboratory. The results are discussed with a focus on the production process and physical and chemical properties of the targets.

  9. Yields of Fission Products from Various Uranium and Thorium Targets

    Energy Technology Data Exchange (ETDEWEB)

    Kronenberg, Andreas [Oak Ridge Associated Universities (ORAU); Spejewski, Eugene H. [Oak Ridge Associated Universities (ORAU); Mervin, Brenden T. [Oak Ridge Associated Universities (ORAU); Jost, Cara [Oak Ridge Associated Universities (ORAU); Carter, H Kennon [Oak Ridge Associated Universities (ORAU); Stracener, Daniel W [ORNL; Greene, John P. [Argonne National Laboratory (ANL); Nolen, Jerry A. [Argonne National Laboratory (ANL); Talbert, Willard L. [TechSource, Inc.

    2008-01-01

    Yield measurements from proton-induced fission have been performed on a number of actinide targets, both Th and U, at the on-line test facility at Oak Ridge National Laboratory. The results are discussed with a focus on the production process and physical and chemical properties of the targets.

  10. Yields of fission products from various uranium and thorium targets.

    Energy Technology Data Exchange (ETDEWEB)

    Kronenberg, A.; Spejewski, E. H.; Mervin, B.; Jost, C.; Carter, H. K.; Stracener, D. W.; Greene, J. P.; Nolen, J. A.; Talbert, W. L.; Physics; Oak Ridge Associated Univ.; ORNL; TechSource, Inc.

    2008-10-31

    Yield measurements from proton-induced fission have been performed on a number of actinide targets, both Th and U, at the on-line test facility at Oak Ridge National Laboratory. The results are discussed with a focus on the production process and physical and chemical properties of the targets.

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

    Indian Academy of Sciences (India)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    Directory of Open Access Journals (Sweden)

    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.

  14. Binary scission configurations in fission of light actinides

    Energy Technology Data Exchange (ETDEWEB)

    Ohtsuki, Tsutomu [Tohoku Univ., Sendai (Japan). Lab. of Nuclear Science; Nagame, Y.; Nishinaka, I.; Tsukada, K.; Ikezoe, H.; Tanikawa, M.; Zhao, Y.L.; Sueki, K.; Nakahara, H.

    1997-07-01

    Mass and kinetic energy distributions of fission fragments have been accurately measured by a double velocity time-of-flight technique in the 13 MeV proton-induced fissions of {sup 232}Th and {sup 238}U. A binary structure is observed in total kinetic energy distributions in the fragments with mass number around A=130 for both the fissions, indicating that there are at least two kinds of scission configurations. A correlation between the scission configurations and mass yield distributions reveals that elongated scission configurations are associated with the symmetric mass distribution and compact scission configurations with the asymmetric mass distribution. (author)

  15. Developing an in-situ Detector of Neutron-Induced Fission for Actinide Sputtering Characterization

    Science.gov (United States)

    Fellers, Deion

    2016-09-01

    The physical mechanism describing the transfer of large amounts of energy due to fission in a material is not well understood and represents one of the modern challenges facing nuclear scientists, with applications including nuclear energy and national defense. Fission fragments cause damage to the material from sputtering of matter as they pass through or near the material's surface. We have developed a new technique at the Los Alamos Neutron Science Center for characterizing the ejecta by using ultracold neutrons (neutrons with kinetic energy less than 300 neV) to induce fission at finely controlled depths in an actinide. This program will ultimately provide a detailed description of the properties of the sputtered particles as a function of the depth of the fission in the material. A key component of this project is accurately quantifying the number of neutron induced fissions in the sample. This poster depicts the development of an in-situ detector of neutron-induced fission for the AShES (Actinide Sputtering from ultracold neutron Exposure at the Surface) experiment.

  16. Actinide fission and capture cross section measurements at ILL: the Mini-INCA project

    Energy Technology Data Exchange (ETDEWEB)

    Letourneau, A.; Bringer, O.; Chabod, S.; Fioni, G.; Foucher, Y.; Marie, F.; Veyssiere, Ch. [CEA/Saclay/DSM/DAPNIA - Gif-sur-Yvette (France); Mahamid, I. Al. [Lawrence Berkeley National Laboratory, E.H. and S division, Berkeley CA (United States); Blandin, Ch. [CEA/Cadarache/DEN/DER/SPEX - Saint-Paul-lez-Durances (France); Chartier, F. [CEA/Saclay/DEN/DPC/SECR - Gif-sur-Yvette (France); Faust, H.; Mutti, P. [Institut Laue-Langevin, BP 156, F-38042 Grenoble Cedex 9 (France)

    2005-07-01

    Fission cross section of short-lived minor actinides is of prime importance for the incineration of minor actinides in high and thermal neutron fluxes. But due to the shortness of their half-lives, measurements are difficult to handle on these isotopes and the existing data present some large discrepancies. An original method has been developed, in the framework of the Mini-INCA project at ILL, to measure the fission and capture cross sections of minor actinides with low error bars associated even for short-lived isotopes. This method lies on a quasi on-line alpha- and gamma-spectroscopy of irradiated samples and on the use of fission micro-chambers. Coupled to a very powerful Monte-Carlo simulation, both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials could be gathered. In this paper, the method is explained in its originality and some recent results are given and compared with existing measurements and evaluated data libraries. (authors)

  17. Fission cross section calculations of actinides with EMPIRE code

    Energy Technology Data Exchange (ETDEWEB)

    Sin, M.; Oblozinsky, P.; Herman,M.; Capote,R.

    2010-04-30

    The cross sections of the neutron induced reactions on {sup 233,234,236}U, {sup 237}Np, {sup 238,242}Pu, {sup 241,243}Am, {sup 242,246}Cm carried out in the energy range 1 keV-20 MeV with EMPIRE code are presented, emphasizing the fission channel. Beside a consistent, accurate set of evaluations, the paper contains arguments supporting the choice of the reaction models and input parameters. A special attention is paid to the fission parameters and their uncertainties.

  18. Actinide production from xenon bombardments of curium-248

    Energy Technology Data Exchange (ETDEWEB)

    Welch, R.B.

    1985-01-01

    Production cross sections for many actinide nuclides formed in the reaction of /sup 129/Xe and /sup 132/Xe with /sup 248/Cm 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 /sup 136/Xe + /sup 248/Cm reaction at a similar energy. When compared to the reaction with /sup 136/Xe, 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, /sup 129/Xe, /sup 132/Xe, and /sup 136/Xe with /sup 197/Au, 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.

  19. Actinides: How well do we know their stellar production?

    Science.gov (United States)

    Goriely, S.; Arnould, M.

    2001-12-01

    The reliable evaluation of the r-process production of the actinides and careful estimates of the uncertainties affecting these predictions are key ingredients especially in nucleo-cosmochronology studies based on the analysis of very metal-poor stars or on the composition of meteorites. This type of information is also required in order to make the best possible use of future high precision data on the actinide composition of galactic cosmic rays, of the local interstellar medium, or of meteoritic grains of presumed circumstellar origin. This paper provides the practitioners in these various fields with the most detailed and careful analysis of the r-process actinide production available to-date. This study is based on a version of the multi-event canonical model of the r-process which discards the largely used waiting point approximation. It considers also different combinations of models for the calculation of nuclear masses, beta -decay and fission rates. Two variants of the model used to predict nuclear reaction rates are adopted. In addition, the influence of the level of Pb and Bi production by the r-process on the estimated actinide production is evaluated by relying on the solar abundances of these two elements. In total, thirty-two different cases are presented, and are considered to give a fair picture of the level of reliability of the predictions of the actinide production, at least in the framework of a simple r-process model. This simplicity is imposed by our inability to identify the proper astrophysical sites for the r-process. As a guide to the practitioners, constraints on the actinide yield predictions and associated uncertainties are suggested on grounds of the measured abundances of r-nuclides, including Th and U, in the star CS 31082-001, and under the critical and questionable assumption of the ``universality'' of the r-process. We also define alternative constraints based on the nucleo-cosmochronological results derived from the present

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-11

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

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

    Science.gov (United States)

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

    2014-09-01

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

  5. Reducing Actinide Production Using Inert Matrix Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, Mark [Colorado School of Mines, Golden, CO (United States)

    2017-08-23

    The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  7. Ceramic Hosts for Fission Products Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Peter C Kong

    2010-07-01

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

  8. Rapid Separation of Fission Product 141La

    Institute of Scientific and Technical Information of China (English)

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

    2013-01-01

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

  9. Fission-product retention in HTGR fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-17

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

  11. Bruyères-le-Châtel Neutron Evaluations of Actinides with the TALYS Code: The Fission Channel

    Science.gov (United States)

    Romain, P.; Morillon, B.; Duarte, H.

    2016-01-01

    For several years, various neutron evaluations of plutonium and uranium isotopes have been performed at Bruyères-le-Châtel (BRC), from 1 keV up to 30 MeV. Since only nuclear reaction models have been used to produce these evaluations, our approach was named the "Full Model" approach. Total, shape elastic and direct inelastic cross sections were obtained from the coupled channels model using a dispersive optical potential developed for actinides, with a large enough coupling scheme including the lowest octupolar band. All other cross sections were calculated using the Hauser-Feshbach theory (TALYS code) with a pre-equilibrium component above 8-10 MeV. In this paper, we focus our attention on the fission channel. More precisely, we will present the BRC contribution to fission modeling and the philosophy adopted in our "Full Model" approach. Performing evaluations with the "Full Model" approach implies the optimization of a large number of model parameters. With increasing neutron incident energy, many residual nuclei produced by nucleon emission also lead to fission. All available experimental data assigned to various fission mechanisms of the same nucleus were used to determine fission barrier parameters. For uranium isotopes, triple-humped fission barriers were required in order to reproduce accurately variations of the experimental fission cross sections. Our BRC fission modeling has shown that the effects of the class II or class III states located in the wells of the fission barrier sometimes provide an anti-resonant transmission rather than a resonant one. Consistent evaluations were produced for a large series of U and Pu isotopes. Resulting files were tested against integral data.

  12. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  13. Fission Product Yields from 232Th, 238U, and 235U Using 14 MeV Neutrons

    Science.gov (United States)

    Pierson, B. D.; Greenwood, L. R.; Flaska, M.; Pozzi, S. A.

    2017-01-01

    Neutron-induced fission yield studies using deuterium-tritium fusion-produced 14 MeV neutrons have not yet directly measured fission yields from fission products with half-lives on the order of seconds (far from the line of nuclear stability). Fundamental data of this nature are important for improving and validating the current models of the nuclear fission process. Cyclic neutron activation analysis (CNAA) was performed on three actinide targets-thorium-oxide, depleted uranium metal, and highly enriched uranium metal-at the University of Michigan's Neutron Science Laboratory (UM-NSL) using a pneumatic system and Thermo-Scientific D711 accelerator-based fusion neutron generator. This was done to measure the fission yields of short-lived fission products and to examine the differences between the delayed fission product signatures of the three actinides. The measured data were compared against previously published results for 89Kr, -90, and -92 and 138Xe, -139, and -140. The average percent deviation of the measured values from the Evaluated Nuclear Data Files VII.1 (ENDF/B-VII.1) for thorium, depleted-uranium, and highly-enriched uranium were -10.2%, 4.5%, and -12.9%, respectively. In addition to the measurements of the six known fission products, 23 new fission yield measurements from 84As to 146La are presented.

  14. Development of fission Mo-99 production technology

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  15. Measurements of thermal fission and capture cross sections of minor actinides within the Mini-INCA project

    Energy Technology Data Exchange (ETDEWEB)

    Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A.; Panebianco, S.; Veyssiere, Ch. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee, 91- Gif sur Yvette (France); Oriol, L. [CEA Cadarache, Dept. d' Etudes des Reacteurs, 13 - Saint Paul lez Durance (France); Chartier, F. [CEA Saclay, Dept. de Physico-Chimie, 91 - Gif sur Yvette (France); Mutti, P. [Institut Laue Langevin, 38 - Grenoble, (France); AlMahamid, I. [Wadsworth Center, New York State Dept. of Health, Albany, NY (United States)

    2008-07-01

    In the framework of nuclear waste transmutation studies, the Mini-INCA project has been initiated at Cea/DSM to determine optimal conditions for transmutation and incineration of Minor Actinides in high intensity neutron fluxes in the thermal region. Our experimental tool is based on alpha- and gamma-spectroscopy of irradiated samples and microscopic fission-chambers. It can provide both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials. The {sup 232}Th, {sup 237}Np, {sup 241}Am, and {sup 244}Cm transmutation chains have been explored in details, showing some discrepancies in comparison with evaluated data libraries but in overall good agreement with recent experimental data. (authors)

  16. Measuring neutron-induced fission cross-section of shortlived actinides using a lead neutron-slowing-down spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Granier, T. E-mail: thierry.granier@cea.fr; Pangault, L.; Ethvignot, T.; Haight, R.C.; Ledoux, X.; Meot, V.; Patin, Y.; Pras, P.; Szmigiel, M.; Rundberg, R.S.; Wilhelmy, J.B

    2003-06-21

    The 'lead-slowing-down-spectrometer' method is an established technique for measuring neutron-induced reaction cross-sections. It is known to provide high neutron fluences below 100 keV. In this work, the possibility of applying this method to the measurement of the neutron-induced fission cross-section of shortlived actinides and in particular of the 77 eV isomer of {sup 235}U is investigated. Numerical simulations and a test-experiment using a photovaltaic cell fission detector demonstrate the feasibility of such a measurement at the Los Alamos Neutron Science Center using 800 MeV proton-induced spallation to provide source neutrons.

  17. Modeling Fission Product Sorption in Graphite Structures

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-04-08

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

  18. Actinides How well do we know their stellar production?

    CERN Document Server

    Goriely, S

    2001-01-01

    The reliable evaluation of the r-process production of the actinides and careful estimates of the uncertainties affecting these predictions are key ingredients especially in nucleo-cosmochronology studies based on the analysis of very metal-poor stars or on the composition of meteorites. This type of information is also required in order to make the best possible use of future high precision data on the actinide composition of galactic cosmic rays, of the local interstellar medium, or of meteoritic grains of presumed circumstellar origin. This paper provides the practitioners in these various fields with the most detailed and careful analysis of the r-process actinide production available to-date. In total, thirty-two different multi-event canonical calculations using different nuclear ingredients or astrophysics conditions are presented, and are considered to give a fair picture of the level of reliability of the predictions of the actinide production, at least in the framework of a simple r-process model. T...

  19. Actinide Foil Production for MPACT Research

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James

    2012-12-19

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

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

    Science.gov (United States)

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

    2005-12-01

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

  2. Fission Product Sorptivity in Graphite

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

  3. Fission fragment mass energy correlation for 13 actinides ranging from Th to Cf

    Energy Technology Data Exchange (ETDEWEB)

    Caitucoli, F.; Barreau, G.; Leroux, B.; Doan, T.P.; Sicre, A.; Asghar, M.; Perrin, P.; Maurel, M.

    1986-01-01

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

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

    Science.gov (United States)

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

    2016-06-01

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

  5. Research on Nuclear Reaction Network Equation for Fission Product Nuclides

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Nuclear Reaction Network Equation calculation system for fission product nuclides was developed. With the system, the number of the fission product nuclides at different time can be calculated in the different neutron field intensity and neutron energy spectra

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

    Directory of Open Access Journals (Sweden)

    Ishizuka Chikako

    2016-01-01

    Full Text Available 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. Dynamical approach to isotopic-distribution of fission fragments from actinide nuclei

    Science.gov (United States)

    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.

  8. Development of fission Mo-99 production technology

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-01

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

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

    Science.gov (United States)

    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

  10. REMOVAL OF FISSION PRODUCTS FROM WATER

    Science.gov (United States)

    Rosinski, J.

    1961-12-19

    A process is given for precipitating fission products from a body of water having a pH of above 6.5. Calcium permanganate and ferrous sulfate are added in a molar ratio of l: 3, whereby a mixed precipitate of manganese dioxide, ferric hydroxide and calcium sulfate is formed; the precipitate carries the fisston products and settles to the bottom of the body of water. (AEC)

  11. SPIDER Progress Towards High Resolution Correlated Fission Product Data

    Science.gov (United States)

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

    2014-09-01

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

  12. Decay spectroscopy of exotic fission products

    Science.gov (United States)

    Rykaczewski, Krzysztof

    2014-09-01

    The beta decay studies of exotic fission products have been performed at the Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Laboratory. The scientific program was focused on the beta-strength function measurements and resulting new half-lives and beta-delayed neutron properties. These observables are important for nuclear structure analysis and modeling of the nucleosynthesis within rapid neutron capture process. The highlights include ten new beta half-lives and several Pn branching ratios including an observation of beta-delayed two-neutron emitter 86Ga. In addition, the measurements of the beta-strength within beta-gamma emission window were performed using a Modular Total Absorption Spectrometer for 22 fission products. These MTAS results are also important for the analysis of reactor anti-neutrino anomaly. The beta decay studies of exotic fission products have been performed at the Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Laboratory. The scientific program was focused on the beta-strength function measurements and resulting new half-lives and beta-delayed neutron properties. These observables are important for nuclear structure analysis and modeling of the nucleosynthesis within rapid neutron capture process. The highlights include ten new beta half-lives and several Pn branching ratios including an observation of beta-delayed two-neutron emitter 86Ga. In addition, the measurements of the beta-strength within beta-gamma emission window were performed using a Modular Total Absorption Spectrometer for 22 fission products. These MTAS results are also important for the analysis of reactor anti-neutrino anomaly. Supported by the U.S. DOE Office of Nuclear Physics under Contracts DE-AC05-00R22725 (ORNL), DE-FG02-96ER40983 (UTK).

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

    Science.gov (United States)

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

    2015-06-01

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

  14. Fission barriers of two odd-neutron actinide nuclei taking into account the time-reversal symmetry breaking at the mean-field level

    Science.gov (United States)

    Koh, Meng-Hock; Bonneau, L.; Quentin, P.; Hao, T. V. Nhan; Wagiran, Husin

    2017-01-01

    Background: For a long time, fission barriers of actinide nuclei have been mostly microscopically calculated for even-even fissioning systems. Calculations in the case of odd nuclei have been performed merely within a so-called equal-filling approximation (EFA) as opposed to an approach taking explicitly into account the time-reversal-breaking properties at the mean-field level—and for only one single-particle configuration. Purpose: We study the dependence of the fission barriers on various relevant configurations (e.g., to evaluate the so-called specialization energy). In addition, we want to assess the relevance of the EFA approach as a function of the deformation, which has been already found for the ground-state deformation. Methods: Calculations within the Hartree-Fock plus BCS approach with self-consistent particle blocking have been performed by using the SkM* Skyrme effective interaction in the particle-hole channel and a seniority force in the particle-particle channel. Axial symmetry has been imposed throughout the whole fission path while the intrinsic parity symmetry has been allowed to be broken in the outer fission barrier region. Results: Potential-energy curves have been determined for six different configurations in 235U and four in 239Pu. Inner and outer fission barriers have been calculated along with some spectroscopic properties in the fission isomeric well. These results have been compared with available data. The influence of time-reversal-breaking mean fields on the solutions has been investigated. Conclusions: A sizable configuration dependence of the fission barrier (width and height) has been demonstrated. A reasonable agreement with available systematic evaluations of fission-barrier heights has been found. The EFA approach has been validated at the large elongations occurring at the outer-barrier region.

  15. Photo-fission Product Yield Measurements at Eγ=13 MeV on 235U, 238U, and 239Pu

    Science.gov (United States)

    Tornow, W.; Bhike, M.; Finch, S. W.; Krishichayan, Fnu; Tonchev, A. P.

    2016-09-01

    We have measured Fission Product Yields (FPYs) in photo-fission of 235U, 238U, and 239Pu at TUNL's High-Intensity Gamma-ray Source (HI γS) using mono-energetic photons of Eγ = 13 MeV. Details of the experimental setup and analysis procedures will be discussed. Yields for approximately 20 fission products were determined. They are compared to neutron-induced FPYs of the same actinides at the equivalent excitation energies of the compound nuclear systems. In the future photo-fission data will be taken at Eγ = 8 . 0 and 10.5 MeV to find out whether photo-fission exhibits the same so far unexplained dependence of certain FPYs on the energy of the incident probe, as recently observed in neutron-induced fission, for example, for the important fission product 147Nd. Work supported by the U. S. Dept. of Energy, under Grant No. DE-FG02-97ER41033, and by the NNSA, Stewardship Science Academic Alliances Program, Grant No. DE-NA0001838 and the Lawrence Livermore, National Security, LLC under Contract No. DE-AC52-07NA27344.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  18. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  19. Fission Product Neutron Cross Section Library and Its Reliability Assessment

    Institute of Scientific and Technical Information of China (English)

    QIAN; Jing; SUN; Zheng-jun; LIU; Ting-jin; SHU; Neng-chuan

    2013-01-01

    A complete library of neutron cross section data has been developed for fission product nuclides.It contains data for 1 121 fission product nuclides of mass number A from 66 to 172 and atomic numbers Z from 22 to 72,where involves a lot of very short-lived radioactive ones.The data were taken from better

  20. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Energy Technology Data Exchange (ETDEWEB)

    G. Palmiotti

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 418 nuclides; (2) Covariance uncertainty data for 185 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions at higher energies for isotopes of F, Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new Decay Data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide

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

    Energy Technology Data Exchange (ETDEWEB)

    Stansfield, O.M.

    1990-07-25

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gentils, A

    2003-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Nash, K. L.

    2000-01-12

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Alexandra Navrotsky

    2010-07-30

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

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

    Science.gov (United States)

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

    2014-09-01

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

  7. Advanced Aqueous Separation Systems for Actinide Partitioning

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Fission-product SiC reaction in HTGR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Montgomery, F.

    1981-07-13

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

  9. Production of heavy actinides in incomplete fusion reactions

    Science.gov (United States)

    Antonenko, N. V.; Cherepanov, E. A.; Iljinov, A. S.; Mebel, M. V.

    1994-10-01

    We present preliminary results of calculations by the phenomenological model of the estimated yield of some heavy actinide isotopes. It is assumed that these isotopes are produced as a result of multinucleon transfers followed by neutrons and charged particle emission A.S. Iljinov and E.A. Cherepanov (1980). The yield P(sub Z, N)(E*) of primary excited actinides is found using the model of N.V. Antonenko and R.V. Jolos (1991). Absolute cross-sections for different binary reaction channels are obtained by summing the cross-sections for all subchannels with an appreciable yield according to J. Wilczynski et al. (1980).

  10. Thermodynamic treatment of noble metal fission products in nuclear fuel

    Science.gov (United States)

    Kaye, M. H.; Lewis, B. J.; Thompson, W. T.

    2007-06-01

    Based on a critical evaluation of the literature, a comprehensive thermodynamic model has been developed for the complete quinary system involving the noble metal fission products in nuclear fuel: Mo-Pd-Rh-Ru-Tc. This treatment was based on the foundation of ten binary systems and an interpolation scheme. The thermodynamic model has been demonstrated to fit the available experimental data for the ternary sub-systems. This work can be used with other models for potentially non-stoichiometric UO 2+ x containing fission products, as well as data for other phases, to assess the chemical form of fission products in irradiated fuel material.

  11. Chemical state of fission products in irradiated uranium carbide fuel

    Science.gov (United States)

    Arai, Yasuo; Iwai, Takashi; Ohmichi, Toshihiko

    1987-12-01

    The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.

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

    Science.gov (United States)

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

    1980-10-01

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

  13. Uncertainties in fission-product decay-heat calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  14. Thermodynamics of fission products in UO2+-x

    Energy Technology Data Exchange (ETDEWEB)

    Nerikar, Pankaj V [Los Alamos National Laboratory

    2009-01-01

    The stabilities of selected fission products - Xe, Cs, and Sr - are investigated as a function of non-stoichiometry x in UO{sub 2{+-}x}. In particular, density functional theory (OFT) is used to calculate the incorporation and solution energies of these fission products at the anion and cation vacancy sites, at the divacancy, and at the bound Schottky defect. In order to reproduce the correct insulating state of UO{sub 2}, the DFT calculations are performed using spin polarization and with the Hubbard U tenn. In general, higher charge defects are more soluble in the fuel matrix and the solubility of fission products increases as the hyperstoichiometry increases. The solubility of fission product oxides is also explored. CS{sub 2}O is observed as a second stable phase and SrO is found to be soluble in the UO{sub 2} matrix for all stoichiometries. These observations mirror experimentally observed phenomena.

  15. B-Plant fission product flowsheets: Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Rey, G.

    1961-09-29

    The technical bases for Phase I of the Fission Product Program have previously been presented for Phase I design. This report represents its sequel for Phase II of the program. Phase I provides the means of segregating, concentrating, and aging crude fission product fractions with market value. Phase II is to provide the means for single-line purification and packaging of megacurie quantities of fission products (principally strontium-90). The technical bases for Phase II project scoping studies are presented herein in the form of process flowsheets and tabulated data. Equipment needs are also described. Conceptual processes and flow diagrams for Phase III are presented for the Waste Management Program and the Fission Product Program.

  16. Studies on the Separation of Cesium From Fission Products

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Attachment behavior of fission products to solution aerosol

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  19. Interactions of fission product vapours with aerosols

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

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

    Science.gov (United States)

    Shcherbina, Natalia; Kivel, Niko; Günther-Leopold, Ines

    2013-06-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Shcherbina, Natalia, E-mail: natalia.shcherbina@psi.ch [Department of Nuclear Energy and Safety, Paul Scherrer Institut (PSI), Villigen 5232 (Switzerland); Kivel, Niko; Günther-Leopold, Ines [Department of Nuclear Energy and Safety, Paul Scherrer Institut (PSI), Villigen 5232 (Switzerland)

    2013-06-15

    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% H{sub 2}/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.

  3. Evaluation and compilation of fission product yields 1993

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  4. Trapping technology for gaseous fission products from voloxidation process

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-05-15

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

  5. Comparison of Fission Product Yields and Their Impact

    Energy Technology Data Exchange (ETDEWEB)

    S. Harrison

    2006-02-01

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

  6. Fission products behaviour during a power transient: Their inventory in an intragranular bubble

    Science.gov (United States)

    Desgranges, L.; Blay, Th.; Lamontagne, J.; Roure, I.; Bienvenu, Ph.

    2017-09-01

    The behaviour of fission products is a key issue during Anticipated Operational Occurrences (AOOs) or Condition II transients or accidental sequence for nuclear fuel. Here we characterized how fission products behaved inside chromium doped UO2 pellet during a power ramp. At the pellet centre fission products have left the UO2 lattice and can be found in bubbles. The composition of the bubbles was determined using an original experimental methodology. The existence of separated precipitates made of metallic fission products for the one, and volatile fission products for the other, was evidenced. This result is discussed with regards to the behaviour of fission products during a power ramp.

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

    Science.gov (United States)

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

    2012-11-16

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

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

    CERN Document Server

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

    2012-01-01

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

  9. Energy dependence of fission product yields from 235U, 238U, and 239Pu with monoenergetic neutrons between thermal and 14.8 MeV

    Science.gov (United States)

    Gooden, Matthew; Arnold, Charles; Bhike, Megha; Bredeweg, Todd; Fowler, Malcolm; Krishichayan; Tonchev, Anton; Tornow, Werner; Stoyer, Mark; Vieira, David; Wilhelmy, Jerry

    2017-09-01

    Under a joint collaboration between TUNL-LANL-LLNL, a set of absolute fission product yield measurements has been performed. The energy dependence of a number of cumulative fission product yields (FPY) have been measured using quasi-monoenergetic neutron beams for three actinide targets, 235U, 238U and 239Pu, between 0.5 and 14.8 MeV. The FPYs were measured by a combination of fission counting using specially designed dual-fission chambers and γ-ray counting. Each dual-fission chamber is a back-to-back ionization chamber encasing an activation target in the center with thin deposits of the same target isotope in each chamber. This method allows for the direct measurement of the total number of fissions in the activation target with no reference to the fission cross-section, thus reducing uncertainties. γ-ray counting of the activation target was performed on well-shielded HPGe detectors over a period of two months post irradiation to properly identify fission products. Reported are absolute cumulative fission product yields for incident neutron energies of 0.5, 1.37, 2.4, 3.6, 4.6, 5.5, 7.5, 8.9 and 14.8 MeV. Preliminary results from thermal irradiations at the MIT research reactor will also be presented and compared to present data and evaluations. This work was performed under the auspices of the U.S. Department of Energy by Los Alamos National Security, LLC under contract DE-AC52-06NA25396, Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344 and by Duke University and Triangle Universities Nuclear Laboratory through NNSA Stewardship Science Academic Alliance grant No. DE-FG52-09NA29465, DE-FG52-09NA29448 and Office of Nuclear Physics Grant No. DE-FG02-97ER41033.

  10. Development of Calculation Code for Fission Product and Corrosion Product in PWR’s Primary Loop

    Institute of Scientific and Technical Information of China (English)

    XU; Zhi-long; WAN; Hai-xia; SHAO; Jing; WU; Xiao-chun; LI; Long; LIU; Xing-min; KE; Guo-tu

    2015-01-01

    With the basis of study on generation,release and migration of fission product,calculation model for each of the above processes was developed,and calculation method for source term of PWR fission products was established.Study on source term of corrosion product in primary loop was been done.Based on the study of corrosion,

  11. Monazite as a suitable actinide waste form

    Energy Technology Data Exchange (ETDEWEB)

    Schlenz, Hartmut; Heuser, Julia; Schmitz, Stephan; Bosbach, Dirk [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung (IEK), Nukleare Entsorgung und Reaktorsicherheit (IEK-6); Neumann, Andreas [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung (IEK), Nukleare Entsorgung und Reaktorsicherheit (IEK-6); RWTH Aachen Univ. (Germany). Inst. for Crystallography

    2013-03-01

    The conditioning of radioactive waste from nuclear power plants and in some countries even of weapons plutonium is an important issue for science and society. Therefore the research on appropriate matrices for the immobilization of fission products and actinides is of great interest. Beyond the widely used borosilicate glasses, ceramics are promising materials for the conditioning of actinides like U, Np, Pu, Am, and Cm. Monazite-type ceramics with general composition LnPO{sub 4} (Ln = La to Gd) and solid solutions of monazite with cheralite or huttonite represent important materials in this field. Monazite appears to be a promising candidate material, especially because of its outstanding properties regarding radiation resistance and chemical durability. This article summarizes the most recent results concerning the characterization of monazite and respective solid solutions and the study of their chemical, thermal, physical and structural properties. The aim is to demonstrate the suitability of monazite as a secure and reliable waste form for actinides. (orig.)

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

    Science.gov (United States)

    Chadwick, Mark B.

    2009-10-01

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

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

    Science.gov (United States)

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

    2010-12-01

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

  14. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M.B.; Herman, M.; Author(s): Chadwick,M.B.; Herman,M.; Oblozinsky,P.; Dunn,M.E.; Danon,Y.; Kahler,A.C.; Smith,D.L.; Pritychenko,B.; Arbanas,G.; Arcilla,R.; Brewer,R.; Brown,D.A.; Capote,R.; Carlson,A.D.; Cho,Y.S.; Derrien,H.; Guber,K.; Hale,G.M.; Hoblit,S.; Holloway,S.: Johnson,T.D.; Kawano,T.; Kiedrowski,B.C.; Kim,H.; Kunieda,S.; Larson,N.M.; Leal,L.; Lestone,J.P.; Little,R.C.; McCutchan,E.A.; MacFarlane,R.E.; MacInnes,M.; Mattoon,C.M.; McKnight,R.D.; Mughabghab,S.F.; Nobre,G.P.A.; Palmiotti,G.; Palumbo,A.; Pigni,M.T.; Pronyaev,V.G.; Sayer,R.O.; Sonzogni,A.A.; Summers,N.C.; Talou,P.; Thompson,I.J.; Trkov,A.; Vogt,R.L.; van der Marck,S.C.; Wallner,A.; White,M.C.; Wiarda,D.; Young,P.G.

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides {sup 235,238}U and {sup 239}Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on {sup 239}Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0

  15. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M. B. [Los Alamos National Laboratory (LANL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Oblozinsky, Pavel [Brookhaven National Laboratory (BNL); Dunn, Michael E [ORNL; Danon, Y. [Rensselaer Polytechnic Institute (RPI); Kahler, A. [Los Alamos National Laboratory (LANL); Smith, Donald L. [Argonne National Laboratory (ANL); Pritychenko, B [Brookhaven National Laboratory (BNL); Arbanas, Goran [ORNL; Arcilla, r [Brookhaven National Laboratory (BNL); Brewer, R [Los Alamos National Laboratory (LANL); Brown, D A [Brookhaven National Laboratory (BNL); Capote, R. [International Atomic Energy Agency (IAEA); Carlson, A. D. [National Institute of Standards and Technology (NIST); Cho, Y S [Korea Atomic Energy Research Institute; Derrien, Herve [ORNL; Guber, Klaus H [ORNL; Hale, G. M. [Los Alamos National Laboratory (LANL); Hoblit, S [Brookhaven National Laboratory (BNL); Holloway, Shannon T. [Los Alamos National Laboratory (LANL); Johnson, T D [Brookhaven National Laboratory (BNL); Kawano, T. [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Kim, H [Korea Atomic Energy Research Institute; Kunieda, S [Los Alamos National Laboratory (LANL); Larson, Nancy M [ORNL; Leal, Luiz C [ORNL; Lestone, J P [Los Alamos National Laboratory (LANL); Little, R C [Los Alamos National Laboratory (LANL); Mccutchan, E A [Brookhaven National Laboratory (BNL); Macfarlane, R E [Los Alamos National Laboratory (LANL); MacInnes, M [Los Alamos National Laboratory (LANL); Matton, C M [Lawrence Livermore National Laboratory (LLNL); Mcknight, R D [Argonne National Laboratory (ANL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Nobre, G P [Brookhaven National Laboratory (BNL); Palmiotti, G [Idaho National Laboratory (INL); Palumbo, A [Brookhaven National Laboratory (BNL); Pigni, Marco T [ORNL; Pronyaev, V. G. [Institute of Physics and Power Engineering (IPPE), Obninsk, Russia; Sayer, Royce O [ORNL; Sonzogni, A A [Brookhaven National Laboratory (BNL); Summers, N C [Lawrence Livermore National Laboratory (LLNL); Talou, P [Los Alamos National Laboratory (LANL); Thompson, I J [Lawrence Livermore National Laboratory (LLNL); Trkov, A. [Jozef Stefan Institute, Slovenia; Vogt, R L [Lawrence Livermore National Laboratory (LLNL); Van der Marck, S S [Nucl Res & Consultancy Grp, Petten, Netherlands; Wallner, A [University of Vienna, Austria; White, M C [Los Alamos National Laboratory (LANL); Wiarda, Dorothea [ORNL; Young, P C [Los Alamos National Laboratory (LANL)

    2011-01-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He; Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl; K; Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides (235,238)U and (239)Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es; Fm; and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on (239)Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide

  16. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Science.gov (United States)

    Chadwick, M. B.; Herman, M.; Obložinský, P.; Dunn, M. E.; Danon, Y.; Kahler, A. C.; Smith, D. L.; Pritychenko, B.; Arbanas, G.; Arcilla, R.; Brewer, R.; Brown, D. A.; Capote, R.; Carlson, A. D.; Cho, Y. S.; Derrien, H.; Guber, K.; Hale, G. M.; Hoblit, S.; Holloway, S.; Johnson, T. D.; Kawano, T.; Kiedrowski, B. C.; Kim, H.; Kunieda, S.; Larson, N. M.; Leal, L.; Lestone, J. P.; Little, R. C.; McCutchan, E. A.; MacFarlane, R. E.; MacInnes, M.; Mattoon, C. M.; McKnight, R. D.; Mughabghab, S. F.; Nobre, G. P. A.; Palmiotti, G.; Palumbo, A.; Pigni, M. T.; Pronyaev, V. G.; Sayer, R. O.; Sonzogni, A. A.; Summers, N. C.; Talou, P.; Thompson, I. J.; Trkov, A.; Vogt, R. L.; van der Marck, S. C.; Wallner, A.; White, M. C.; Wiarda, D.; Young, P. G.

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-22

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-22

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

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

    Science.gov (United States)

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

    2014-05-01

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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Directory of Open Access Journals (Sweden)

    Talou Patrick

    2016-01-01

    Full Text Available 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-01-01

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

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

    Science.gov (United States)

    Herrmann, Steven Douglas

    2014-05-27

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

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

  6. Chemical state of fission products in irradiated UO 2

    Science.gov (United States)

    Imoto, S.

    1986-08-01

    The chemical state of fission products in irradiated UO 2 fuel has been estimated for FBR as well as LWR on the basis of equilibrium calculation with the SOLGASMIX-PV code. The system considered for the calculation is composed of a gas phase, a CaF 2 type oxide phase, three grey phases, a noble metal alloy, a mixed telluride phase and several other phases each consisting of single compound. The distribution of elements into these phases and the amount of chemical species in each phase at different temperatures are obtained as a function of oxygen potential for LWR and FBR. Changes of the chemical potential of the fuel-fission products system during burnup are also evaluated with particular attention to the difference between LWR and FBR. Some informations obtained by the calculation are compared with the results of post irradiation examination of UO 2 fuels.

  7. RAPID QUANTITATION OF URANIUM FROM MIXED FISSION PRODUCT SAMPLES

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-09

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

  8. 1: Mass asymmetric fission barriers for {sup 98}Mo; 2: Synthesis and characterization of actinide-specific chelating agents

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-28

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

  11. Actinide Production in the Reaction of Heavy Ions withCurium-248

    Energy Technology Data Exchange (ETDEWEB)

    Moody, K.J.

    1983-07-01

    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 {sup 248}Cm with {sup 18}O, {sup 86}Kr and {sup 136}Xe projectiles at several energies near and below the Coulomb barrier were determined. The results are compared with yields from {sup 48}Ca and {sup 238}U bombardments of {sup 248}Cm. 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 {sup 136}Xe and {sup 238}U the new species {sup 248}Am, {sup 249}Am and {sup 247}Pu should be produced with large cross sections from a {sup 248}Cm target. A preliminary, unsuccessful attempt to isolate {sup 247}Pu 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 {sup 251}Bk decay, necessary for calculating the {sup 251}Bk cross section, are also determined.

  12. Advances in fuel materials for the transmutation of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Prunier, C.

    1994-12-31

    The physical feasibility of actinides, spent fuels and fission products burning in fission reactors is well understood. In fast reactors, this operation is more favourable. The homogeneous recycling mode has had a preliminary validation in Phenix (the Super fact experiment). For the heterogenous recycling mode, past experience for {sup 238} Pu production in thermal spectrum was obtained with Np O{sub 2}-Mg O targets. An irradiation experiment in Phenix blanket is foreseen with the same type of target. The {sup 237} Np problem seems to be most conveniently treated, even in the short term, by homogeneous recycling with Pu in fast reactors. (author). 15 figs., 4 tabs.

  13. Production of actinide nuclei by multi-nucleon transfer

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Penning trap mass measurements and laser spectroscopy on neutron-rich fission products extracted from the research reactor TRIGA-Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Eibach, Martin; Ketelaer, Jens; Ketter, Jochen; Knuth, Konstantin [Institut fuer Physik, Universitaet Mainz (Germany); Blaum, Klaus; Nagy, Szilard [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); Eberhardt, Klaus; Noertershaeuser, Wilfried [Institut fuer Kernchemie, Universitaet Mainz (Germany); Herfurth, Frank [GSI, Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Smorra, Christian [Physikalisches Institut, Universitaet Heidelberg (Germany); Institut fuer Kernchemie, Universitaet Mainz (Germany)

    2009-07-01

    TRIGA-SPEC is a setup for Penning trap mass spectrometry and collinear laser spectroscopy on short-lived neutron-rich nuclides located at the research reactor TRIGA-Mainz. It is dedicated to the determination of nuclear ground-state properties like masses and charge-radii. The nuclides are produced by neutron-induced fission of an actinide target located in a target chamber near the reactor core. It is required to extract the nuclides fast and with high efficiency from the target chamber in order to make precision experiments on short-living species with half-lives in the order of 1s. To this end, they are flushed out with a helium gas jet containing carbon aerosols and transported through a skimmer region to an ECR ion source. The characterisation of the carbon aerosol generator and the verification of transported fission products are presented.

  15. Accelerator-Driven Production of Fission 99Mo

    Energy Technology Data Exchange (ETDEWEB)

    Youker, A. J.; Chemerisov, S. D.; Tkac, P.; Krebs, J. F.; Rotsch, D. A.; Kalensky, M.; Heltemes, T. A.; Alford, K.; Byrnes, J. P.; Gromov, R.; Hafenrichter, L.; Hebden, A. S.; Jerden, J. L.; Jonah, C. D.; Makarashvili, V.; Quigley, K. J.; Schneider, J. F.; Stepinski, D. C.; Wesolowski, K. A.; Vandegrift, G. F.

    2016-01-01

    I{esults al'e reportecl for the procluction of eeMo l'rom the accelerator-clriveu subcritical fission of a low enriohed uranir¡m (Ltju) aqì.reous solution. Phase I ol'these experiments used a 5 L r.rranyl sulfate solution with a eeMo encl-of-irracliation produotion limit of 2 Ci. The separation, recovery, and pulification of eeMo were demonstrated Lrsing the recyclecl solution. Fission product paltitioriing tl'ends will be shown for the recovery colutt'ttt, concentratiorl colurnn, and LE,U Modified Cintichem prooess. The results fi'om a 1.4 Ci oeMo production run, where the fìnal product was seut to GE Flealthcare for testing, will be highlightecl. The information gained cluring Phase ì lias signilìcantly irnpacted the clesign and implernentation of Phase ll. Phase II focuses on an end-of-irradiation ploduction of 20 Ci of eeMo and a fissior'ì power density similar to the production fàcility in a20 L LìlU uranyl sulfate solution.

  16. Fission product plateout/liftoff/washoff test plan. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Acharya, R.; Hanson, D.

    1988-05-01

    A test program is planned in the COMEDIE loop of the Commissariat a l`Energy Atomique (CEA), Grenoble, France, to generate integral test data for the validation of computer codes used to predict fission product transport and core corrosion in the Modular High Temperature Gas-Cooled Reactor (MHTGR). The inpile testing will be performed by the CEA under contract from the US Department of Energy (DOE); the contract will be administered by Oak Ridge National Laboratory (ORNL). The primary purpose of this test plan is to provide an overview of the proposed program in terms of the overall scope and schedule. 8 refs, 3 figs.

  17. Equilibrium Temperature Profiles within Fission Product Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  18. Decay characteristics of fission products and summation calculation

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

  19. Chemical and Ceramic Methods Toward Safe Storage of Actinides

    Energy Technology Data Exchange (ETDEWEB)

    P.E.D. Morgan; R.M. Housley; J.B. Davis; M.L. DeHaan

    2005-08-19

    A very import, extremely-long-term, use for monazite as a radwaste encapsulant has been proposed. THe use of ceramic La-monazite for sequestering actinides (isolating them from the environment), especially plutonium and some other radioactive elements )e.g., fission-product rare earths), had been especially championed by Lynn Boatner of ORNL. Monazite may be used alone or, copying its compatibility with many other minerals in nature, may be used in diverse composite combinations.

  20. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

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

  1. Airborne measurements of fission product fall-out

    Energy Technology Data Exchange (ETDEWEB)

    Hovgaard, J.; Korsbech, U.

    1992-12-01

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

  2. Biological removal of cationic fission products from nuclear wastewater.

    Science.gov (United States)

    Ngwenya, N; Chirwa, E M N

    2011-01-01

    Nuclear energy is becoming a preferred energy source amidst rising concerns over the impacts of fossil fuel based energy on global warming and climate change. However, the radioactive waste generated during nuclear power generation contains harmful long-lived fission products such as strontium (Sr). In this study, cationic strontium uptake from solution by microbial cultures obtained from mine wastewater is evaluated. A high strontium removal capacity (q(max)) with maximum loading of 444 mg/g biomass was achieved by a mixed sulphate reducing bacteria (SRB) culture. Sr removal in SRB was facilitated by cell surface based electrostatic interactions with the formation of weak ionic bonds, as 68% of the adsorbed Sr(2+) was easily desorbed from the biomass in an ion exchange reaction with MgCl₂. To a lesser extent, precipitation reactions were also found to account for the removal of Sr from aqueous solution as about 3% of the sorbed Sr was precipitated due to the presence of chemical ligands while the remainder occurred as an immobile fraction. Further analysis of the Sr-loaded SRB biomass by scanning electron microscopy (SEM) coupled to energy dispersive X-ray (EDX) confirmed extracellular Sr(2+) precipitation as a result of chemical interaction. In summary, the obtained results demonstrate the prospects of using biological technologies for the remediation of industrial wastewaters contaminated by fission products.

  3. Microbial Transformations of Actinides and Other Radionuclides

    Energy Technology Data Exchange (ETDEWEB)

    Francis,A.J.; Dodge, C. J.

    2009-01-07

    Microorganisms can affect the stability and mobility of the actinides and other radionuclides released from nuclear fuel cycle and from nuclear fuel reprocessing plants. Under appropriate conditions, microorganisms can alter the chemical speciation, solubility and sorption properties and thus could increase or decrease the concentrations of radionuclides in solution in the environment and the bioavailability. Dissolution or immobilization of radionuclides is brought about by direct enzymatic action or indirect non-enzymatic action of microorganisms. Although the physical, chemical, and geochemical processes affecting dissolution, precipitation, and mobilization of radionuclides have been extensively investigated, we have only limited information on the effects of microbial processes and biochemical mechanisms which affect the stability and mobility of radionuclides. The mechanisms of microbial transformations of the major and minor actinides U, Pu, Cm, Am, Np, the fission products and other radionuclides such as Ra, Tc, I, Cs, Sr, under aerobic and anaerobic conditions in the presence of electron donors and acceptors are reviewed.

  4. Actinides and Life's Origins.

    Science.gov (United States)

    Adam, Zachary

    2007-12-01

    There are growing indications that life began in a radioactive beach environment. A geologic framework for the origin or support of life in a Hadean heavy mineral placer beach has been developed, based on the unique chemical properties of the lower-electronic actinides, which act as nuclear fissile and fertile fuels, radiolytic energy sources, oligomer catalysts, and coordinating ions (along with mineralogically associated lanthanides) for prototypical prebiotic homonuclear and dinuclear metalloenzymes. A four-factor nuclear reactor model was constructed to estimate how much uranium would have been required to initiate a sustainable fission reaction within a placer beach sand 4.3 billion years ago. It was calculated that about 1-8 weight percent of the sand would have to have been uraninite, depending on the weight percent, uranium enrichment, and quantity of neutron poisons present within the remaining placer minerals. Radiolysis experiments were conducted with various solvents with the use of uraniumand thorium-rich minerals (metatorbernite and monazite, respectively) as proxies for radioactive beach sand in contact with different carbon, hydrogen, oxygen, and nitrogen reactants. Radiation bombardment ranged in duration of exposure from 3 weeks to 6 months. Low levels of acetonitrile (estimated to be on the order of parts per billion in concentration) were conclusively identified in 2 setups and tentatively indicated in a 3(rd) by gas chromatography/mass spectrometry. These low levels have been interpreted within the context of a Hadean placer beach prebiotic framework to demonstrate the promise of investigating natural nuclear reactors as power production sites that might have assisted the origins of life on young rocky planets with a sufficiently differentiated crust/mantle structure. Future investigations are recommended to better quantify the complex relationships between energy release, radioactive grain size, fissionability, reactant phase, phosphorus

  5. Actinides and Life's Origins

    Science.gov (United States)

    Adam, Zachary

    2007-12-01

    There are growing indications that life began in a radioactive beach environment. A geologic framework for the origin or support of life in a Hadean heavy mineral placer beach has been developed, based on the unique chemical properties of the lower-electronic actinides, which act as nuclear fissile and fertile fuels, radiolytic energy sources, oligomer catalysts, and coordinating ions (along with mineralogically associated lanthanides) for prototypical prebiotic homonuclear and dinuclear metalloenzymes. A four-factor nuclear reactor model was constructed to estimate how much uranium would have been required to initiate a sustainable fission reaction within a placer beach sand 4.3 billion years ago. It was calculated that about 1-8 weight percent of the sand would have to have been uraninite, depending on the weight percent, uranium enrichment, and quantity of neutron poisons present within the remaining placer minerals. Radiolysis experiments were conducted with various solvents with the use of uranium- and thorium-rich minerals (metatorbernite and monazite, respectively) as proxies for radioactive beach sand in contact with different carbon, hydrogen, oxygen, and nitrogen reactants. Radiation bombardment ranged in duration of exposure from 3 weeks to 6 months. Low levels of acetonitrile (estimated to be on the order of parts per billion in concentration) were conclusively identified in 2 setups and tentatively indicated in a 3rd by gas chromatography/mass spectrometry. These low levels have been interpreted within the context of a Hadean placer beach prebiotic framework to demonstrate the promise of investigating natural nuclear reactors as power production sites that might have assisted the origins of life on young rocky planets with a sufficiently differentiated crust/mantle structure. Future investigations are recommended to better quantify the complex relationships between energy release, radioactive grain size, fissionability, reactant phase, phosphorus

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Science.gov (United States)

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

    2012-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-15

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

  11. A systematic study of actinide production from the interactions of heavy ions with sup 248 Cm

    Energy Technology Data Exchange (ETDEWEB)

    Leyba, J.D.

    1990-09-07

    Production cross sections for heavy actinides produced from the interactions of {sup 12}C, {sup 31}P, {sup 40}Ar, and {sup 44}Ca ions with {sup 248}Cm were measured at energies ranging from 0.98 to 1.35 X Coulomb barrier. The recoiling reaction products were collected in copper or gold catcher foils located near the {sup 248}Cm target. Separate fractions of Bk, Cf, Es, Fm, and Md were obtained from a radiochemical separation procedure. For the {sup 12}C system, a He/KCl jet was used to transport the recoiling No activities of interest to a rotating wheel system. The isotopic distributions of the actinide products were found to be essentially symmetric about the maximum with full-widths-at-half-maximum of approximately 2.5 mass units. Isotopic distributions of the {sup 12}C, {sup 31}P, {sup 40}Ar, and {sup 44}Ca systems were found to be very similar to the {sup 40,48}Ca systems studied previously. The maxima of the isotopic distributions generally occurred for those reaction channels which involved the exchange of the fewest number of nucleons between the target and projectile for which the calculated excitation energy was a positive quantity. Additionally, the maxima of the excitation functions occurred at those projectile energies which were consistent with the calculated reaction barriers based upon a binary reaction mechanism. The experimental data from the four systems investigated were compared to several models of heavy ion interactions including a damped reaction mechanism, compound nucleus formation and subsequent particle evaporation, and classical partial wave calculations for binary systems.

  12. Solar r-process-constrained actinide production in neutrino-driven winds of supernovae

    Science.gov (United States)

    Goriely, S.; Janka, H.-Th.

    2016-07-01

    Long-lived radioactive nuclei play an important role as nucleo-cosmochronometers and as cosmic tracers of nucleosynthetic source activity. In particular, nuclei in the actinide region like thorium, uranium, and plutonium can testify to the enrichment of an environment by the still enigmatic astrophysical sources that are responsible for the production of neutron-rich nuclei by the rapid neutron-capture process (r-process). Supernovae and merging neutron-star (NS) or NS-black hole binaries are considered as most likely sources of the r-nuclei. But arguments in favour of one or the other or both are indirect and make use of assumptions; they are based on theoretical models with remaining simplifications and shortcomings. An unambiguous observational determination of a production event is still missing. In order to facilitate searches in this direction, e.g. by looking for radioactive tracers in stellar envelopes, the interstellar medium or terrestrial reservoirs, we provide improved theoretical estimates and corresponding uncertainty ranges for the actinide production (232Th, 235, 236, 238U, 237Np, 244Pu, and 247Cm) in neutrino-driven winds of core-collapse supernovae. Since state-of-the-art supernova models do not yield r-process viable conditions - but still lack, for example, the effects of strong magnetic fields - we base our investigation on a simple analytical, Newtonian, adiabatic and steady-state wind model and consider the superposition of a large number of contributing components, whose nucleosynthesis-relevant parameters (mass weight, entropy, expansion time-scale, and neutron excess) are constrained by the assumption that the integrated wind nucleosynthesis closely reproduces the Solar system distribution of r-process elements. We also test the influence of uncertain nuclear physics.

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

    Directory of Open Access Journals (Sweden)

    Rubchenya V. A.

    2013-12-01

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

  14. Neutron Cross Section Covariances for Structural Materials and Fission Products

    Science.gov (United States)

    Hoblit, S.; Cho, Y.-S.; Herman, M.; Mattoon, C. M.; Mughabghab, S. F.; Obložinský, P.; Pigni, M. T.; Sonzogni, A. A.

    2011-12-01

    We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from 10 eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also 23Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.

  15. Neutron Cross Section Covariances for Structural Materials and Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Hoblit, S.; Hoblit,S.; Cho,Y.-S.; Herman,M.; Mattoon,C.M.; Mughabghab,S.F.; Oblozinsky,P.; Pigni,M.T.; Sonzogni,A.A.

    2011-12-01

    We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from 10{sup -5} eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also {sup 23}Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.

  16. Cascade-exciton model analysis of level density parameter dependence in proton induced fission cross sections of some sub-actinide nuclei

    Institute of Scientific and Technical Information of China (English)

    Zafar Yasin; Warda Iram; Muhammad Asghar; M. Ikram Shahzad

    2011-01-01

    Fission cross sections strongly depend on the ratio of the level density parameter in fission to neutron emission,af/an.In this work,a cascade-exciton model implemented in the code CEM95 has been used to observe this effect for proton induced fission cross sections of tungsten,lead and bismuth.The method was employed using different level density parameter ratios for each fission cross section calculation.The calculated fission cross sections are compared with the available experimental data in the literature.It has been observed that a change of the ratio of the level density parameter,af/an,is necessary with the incident energy of the proton,to best estimate the fission cross sections in CEM95.

  17. Radiochemical studies of neutron deficient actinide isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Williams, K.E.

    1978-04-01

    The production of neutron deficient actinide isotopes in heavy ion reactions was studied using alpha, gamma, x-ray, and spontaneous fission detection systems. A new isotope of berkelium, /sup 242/Bk, was produced with a cross-section of approximately 10 ..mu..b in reactions of boron on uranium and nitrogen on thorium. It decays by electron capture with a half-life of 7.0 +- 1.3 minutes. The alpha-branching ratio for this isotope is less than 1% and the spontaneous fission ratio is less than 0.03%. Studies of (Heavy Ion, pxn) and (Heavy Ion, ..cap alpha..xn) transfer reactions in comparison with (Heavy ion, xn) compound nucleus reactions revealed transfer reaction cross-sections equal to or greater than the compound nucleus yields. The data show that in some cases the yield of an isotope produced via a (H.I.,pxn) or (H.I.,..cap alpha..xn) reaction may be higher than its production via an xn compound nucleus reaction. These results have dire consequences for proponents of the ''Z/sub 1/ + Z/sub 2/ = Z/sub 1+2/'' philosophy. It is no longer acceptable to assume that (H.I.,pxn) and (H.I.,..cap alpha..xn) product yields are of no consequence when studying compound nucleus reactions. No evidence for spontaneous fission decay of /sup 228/Pu, /sup 230/Pu, /sup 232/Cm, or /sup 238/Cf was observed indicating that strictly empirical extrapolations of spontaneous fission half-life data is inadequate for predictions of half-lives for unknown neutron deficient actinide isotopes.

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

    Science.gov (United States)

    Shields, Dan; Spider Team

    2015-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Schenter, R. E. [Smart Bullets Inc., 2521 SW Luradel Street, Portland, OR 97219 (United States); Brown, G. J. [Ozarks Medical Center, Cancer Treatment Center, Shaw Medical Building, 1111 Kentucky Avenue, West Plains, MO 65775 (United States); Holden, C. S. [Thorenco LLC, 369 Pine Street, San Francisco, CA 94104 (United States)

    2006-07-01

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

  1. Evaluation of Neutron Induced Reactions for 32 Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Il

    2007-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E.; Turnbull, Andrew J.

    2010-01-29

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

  4. Chemical thermodynamics of complex systems: fission product behavior in LWR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Kohli, R.

    1981-03-01

    A detailed thermodynamic assessment has been made of the chemical reactions of fission products in LWR fuel rods. Using recent thermodynamic data and the in-reactor oxygen potential and temperature range of LWRs, equilibrium thermodynamic calculations were performed for the most plausible reactions of the fission products. The emphasis in this model is on the chemistry of cesium and rubidium and their reactions with the fuel, other fission products, and the zircaloy cladding. The model predictions are discussed for their implications in fuel-cladding interactions.

  5. Radiation Damage and Fission Product Release in Zirconium Nitride

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-08-29

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

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

    Science.gov (United States)

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

    2013-04-01

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

  8. Fission products from the damaged Fukushima reactor observed in Hungary.

    Science.gov (United States)

    Bihari, Árpád; Dezső, Zoltán; Bujtás, Tibor; Manga, László; Lencsés, András; Dombóvári, Péter; Csige, István; Ranga, Tibor; Mogyorósi, Magdolna; Veres, Mihály

    2014-01-01

    Fission products, especially (131)I, (134)Cs and (137)Cs, from the damaged Fukushima Dai-ichi nuclear power plant (NPP) were detected in many places worldwide shortly after the accident caused by natural disaster. To observe the spatial and temporal variation of these isotopes in Hungary, aerosol samples were collected at five locations from late March to early May 2011: Institute of Nuclear Research, Hungarian Academy of Sciences (ATOMKI, Debrecen, East Hungary), Paks NPP (Paks, South-Central Hungary) as well as at the vicinity of Aggtelek (Northeast Hungary), Tapolca (West Hungary) and Bátaapáti (Southwest Hungary) settlements. In addition to the aerosol samples, dry/wet fallout samples were collected at ATOMKI, and airborne elemental iodine and organic iodide samples were collected at Paks NPP. The peak in the activity concentration of airborne (131)I was observed around 30 March (1-3 mBq m(-3) both in aerosol samples and gaseous iodine traps) with a slow decline afterwards. Aerosol samples of several hundred cubic metres of air showed (134)Cs and (137)Cs in detectable amounts along with (131)I. The decay-corrected inventory of (131)I fallout at ATOMKI was 2.1±0.1 Bq m(-2) at maximum in the observation period. Dose-rate contribution calculations show that the radiological impact of this event at Hungarian locations was of no considerable concern.

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

    Energy Technology Data Exchange (ETDEWEB)

    Myrick, T.E.; Schaich, R.W.; Williams, F.V.

    1983-08-01

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

  10. Actinides-1981

    Energy Technology Data Exchange (ETDEWEB)

    1981-09-01

    Abstracts of 134 papers which were presented at the Actinides-1981 conference are presented. Approximately half of these papers deal with electronic structure of the actinides. Others deal with solid state chemistry, nuclear physic, thermodynamic properties, solution chemistry, and applied chemistry.

  11. Determination of {sup 140}La fission product interference factor for INAA

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro Jr, Iberê S.; Genezini, Frederico A.; Saiki, Mitiko; Zahn, Guilherme S. [Instituto de Pesquisas Energéticas e Nucleares, IPEN/CNEN-SP, São Paulo - SP (Brazil)

    2014-11-11

    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 {sup 140}La, because this radioisotope used in INAA for the determination of concentration the La is also produced by the {sup −}β of {sup 140}Ba, an uranium fission product. The {sup 140}La interference factor was studied in this work and a factor to describe its time dependence was obtained.

  12. Solar r-process-constrained actinide production in neutrino-driven winds of supernovae

    CERN Document Server

    Goriely, S

    2016-01-01

    Long-lived radioactive nuclei play an important role as nucleo-cosmochronometers and as cosmic tracers of nucleosynthetic source activity. In particular nuclei in the actinide region like thorium, uranium, and plutonium can testify to the enrichment of an environment by the still enigmatic astrophysical sources that are responsible for the production of neutron-rich nuclei by the rapid neutron-capture process (r-process). Supernovae and merging neutron-star (NS) or NS-black hole binaries are considered as most likely sources of the r-nuclei. But arguments in favour of one or the other or both are indirect and make use of assumptions; they are based on theoretical models with remaining simplifications and shortcomings. An unambiguous observational determination of a production event is still missing. In order to facilitate searches in this direction, e.g.\\ by looking for radioactive tracers in stellar envelopes, the interstellar medium or terrestrial reservoirs, we provide improved theoretical estimates and co...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-21

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

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

    Directory of Open Access Journals (Sweden)

    Zakari-Issoufou A.-A.

    2014-03-01

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

  15. Actinide and lanthanide separation process (ALSEP)

    Science.gov (United States)

    Guelis, Artem V.

    2013-01-15

    The process of the invention is the separation of minor actinides from lanthanides in a fluid mixture comprising, fission products, lanthanides, minor actinides, rare earth elements, nitric acid and water by addition of an organic chelating aid to the fluid; extracting the fluid with a solvent comprising a first extractant, a second extractant and an organic diluent to form an organic extractant stream and an aqueous raffinate. Scrubbing the organic stream with a dicarboxylic acid and a chelating agent to form a scrubber discharge. The scrubber discharge is stripped with a simple buffering agent and a second chelating agent in the pH range of 2.5 to 6.1 to produce actinide and lanthanide streams and spent organic diluents. The first extractant is selected from bis(2-ethylhexyl)hydrogen phosphate (HDEHP) and mono(2-ethylhexyl)2-ethylhexyl phosphonate (HEH(EHP)) and the second extractant is selected from N,N,N,N-tetra-2-ethylhexyl diglycol amide (TEHDGA) and N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA).

  16. Advanced Aqueous Separation Systems for Actinide Partitioning

    Energy Technology Data Exchange (ETDEWEB)

    Nash, Ken [Washington State Univ., Pullman, WA (United States); Martin, Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lumetta, Gregg [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-04-02

    One of the most challenging aspects of advanced processing of used nuclear fuel is the separation of transplutonium actinides from fission product lanthanides. This separation is essential if actinide transmutation options are to be pursued in advanced fuel cycles, as lanthanides compete with actinides for neutrons in both thermal and fast reactors, thus limiting efficiency. The separation is difficult because the chemistry of Am3+ and Cm3+ is nearly identical to that of the trivalent lanthanides (Ln3+). The prior literature teaches that two approaches offer the greatest probability of devising a successful group separation process based on aqueous processes: 1) the application of complexing agents containing ligand donor atoms that are softer than oxygen (N, S, Cl-) or 2) changing the oxidation state of Am to the IV, V, or VI state to increase the essential differences between Am and lanthanide chemistry (an approach utilized in the PUREX process to selectively remove Pu4+ and UO22+ from fission products). The latter approach offers the additional benefit of enabling a separation of Am from Cm, as Cm(III) is resistant to oxidation and so can easily be made to follow the lanthanides. The fundamental limitations of these approaches are that 1) the soft(er) donor atoms that interact more strongly with actinide cations than lanthanides form substantially weaker bonds than oxygen atoms, thus necessitating modification of extraction conditions for adequate phase transfer efficiency, 2) soft donor reagents have been seen to suffer slow phase transfer kinetics and hydro-/radiolytic stability limitations and 3) the upper oxidation states of Am are all moderately strong oxidants, hence of only transient stability in media representative of conventional aqueous separations systems. There are examples in the literature of both approaches having been described. However, it is not clear at present that any extant process is sufficiently robust for application at the scale

  17. Fission Product Yields of 233U, 235U, 238U and 239Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

    Science.gov (United States)

    Laurec, J.; Adam, A.; de Bruyne, T.; Bauge, E.; Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G.; Authier, N.; Casoli, P.

    2010-12-01

    The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for 235U(n,f), 239Pu(n,f) in a thermal spectrum, for 233U(n,f), 235U(n,f), and 239Pu(n,f) reactions in a fission neutron spectrum, and for 233U(n,f), 235U(n,f), 238U(n,f), and 239Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.

  18. Nuclear data evaluation of long-lived fission products: Microscopic vs. phenomenological optical potentials

    Directory of Open Access Journals (Sweden)

    Minato Futoshi

    2017-01-01

    Full Text Available Neutron-nucleus cross sections calculated by macroscopic potentials are compared with a microscopic one to study the performance for long-lived fission products. The macroscopic potentials show a good agreement with the microscopic one at higher energies, where neutron experimental data are scarce. Besides it, analyses of differential elastic cross sections at low energies also suggest that the macroscopic potentials are still effective and applicable enough for the long-lived fission products.

  19. Towards a multiscale approach for assessing fission product behaviour in UN

    Energy Technology Data Exchange (ETDEWEB)

    Klipfel, M.; Di Marcello, V.; Schubert, A.; Laar, J. van de; Van Uffelen, P., E-mail: Paul.van-Uffelen@ec.europa.eu

    2013-11-15

    Ab initio modelling of fission products (i.e. Nb, Y, Gd, Nd, Zr, Sm, Eu, Ce, Ba, Mo, Sr, Rh, Pd, and Ru) in uranium nitride is carried out by assessing the incorporation, along with their contributions to local swelling of the fuel matrix. Fission products (FP's) in UN have shown to be preferably accommodated at U vacancies in bound [1 0 0]-Schottky defects, nevertheless, similar incorporation energies were found at a single U vacancy. From the investigated incorporation and migration mechanism, we found that FP's in UN predominately migrate along U–U vacancies, since the incorporation energies for all FP are lowest at single U vacancy or at the U vacancy in a Schottky defect. The energy required to induce a migration of a volatile FP from an N vacancy to U vacancy is about 4–5.5 eV. The local volume changes caused by the fission-product substitution have been assessed by means of DFT and combined with the fission-product concentrations obtained by means of neutron calculations (SCALE) to predict fission product swelling in UN. The linear swelling of nitride fuel resulting from these calculations, and the assumption that fission products do not interact and form secondary phases, leads to a reasonable estimation for the swelling rate as a function of burn-up (or time) when compared with empirical correlations in the open literature.

  20. Aqueous waste management for minor actinides and lanthanides separation process

    Energy Technology Data Exchange (ETDEWEB)

    Pochon, P.; Boyer, S.; Sans, D

    2004-07-01

    The French strategy of high level radioactive aqueous waste management is an incorporation in glassy fission products containers. Therefore, nitric acid soluble organic reagents needed for minor actinides and lanthanides selective separation from fission product solutions have to be sufficiently removed to reach carbon concentrations compatible with calcinator working. Thus, the ability of reagents to be oxidized under concentration conditions with or without denitration becomes a criteria of selection and have been studied. Further, if not working, other operations like hot hydrogen peroxide oxidation, catalyzed or not, are investigated. Reagents involved in this work are mainly complexing products (N-(2-Hydroxyethyl) Ethylene-diamine-tri-acetic Acid), pH keeping reagents (carboxylic acids like citric, glycolic, tartaric and lactic acid) and alkaline species (Tetramethylammonium hydroxide). Behaviour of acetic acid, which is often the main degradation product, has also been observed. In all cases, reaction products are characterized. (authors)

  1. Temperature dependent fission product removal efficiency due to pool scrubbing

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shunsuke, E-mail: suchida@iae.or.jp [Institute of Applied Energy, 1-14-2, Nishi-Shimbashi, Minato-ku, Tokyo 105-0003 (Japan); Itoh, Ayumi; Naitoh, Masanori; Okada, Hidetoshi; Suzuki, Hiroyuki [Institute of Applied Energy, 1-14-2, Nishi-Shimbashi, Minato-ku, Tokyo 105-0003 (Japan); Hanamoto, Yukio [KAKEN, Inc., 1044, Hori-machi, Mito 310-0903 (Japan); Osakabe, Masahiro [Tokyo University of Marine Science & Technology, Koutou-ku, Tokyo 135-8533 (Japan); Fujikawa, Masahiro [Japan Broadcasting Corporation, 2-2-1, Jinnan, Shibuya-ku, Tokyo 150-8001 (Japan)

    2016-03-15

    Highlights: • Pool temperature effects on the FP removal were not clearly concluded in the previous publications. • It was confirmed that the removal efficiency decreased with temperature around the boiling point. • A modified empirical formula for FP removal was proposed as a function of sub-cooling temperature. • DF could be predicted with an accuracy within a factor of 2 with the proposed formula. - Abstract: The wet-well of boiling water reactors plays important roles not only to suppress the pressure in the primary containment vessel due to steam scrubbing effects during severe accidents but also to mitigate release of radioactive fission products (FP), aerosols and particulates, into the environment. The effects of steam scrubbing in the wet-well on FP removal have been well studied and reported by changing major parameters determining the removal efficiencies, e.g., aerosol diameters, submergence (depth of scrubbing nozzles) and steam/non-condensable gas volume fraction. Unfortunately, the effects of pool temperature on the FP removal were not clearly concluded in the previous publications, though it would be easily expected that boiling in the pool resulted in reduced aerosol removal efficiency. In order to determine the temperature effects on FP removal efficiency, amounts of cesium in aerosols released from scrubbing pool were measured by changing pool temperature in mini and medium scale scrubbing experiments, and then, it was confirmed that the removal efficiency clearly decreased with temperature around the boiling point. Then, a modified empirical formula to express the FP removal around the boiling point temperature was proposed as a function of sub-cooling temperature by applying the effective steam volume fraction, which was designated as the volume ratio of condensed steam in the pool versus the sum of input steam and non-condensable gas. By comparing the measured removal efficiency with the calculated, it was validated that the

  2. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    Energy Technology Data Exchange (ETDEWEB)

    Leal, Luiz C [ORNL; Derrien, Herve [ORNL; Dunn, Michael E [ORNL; Mueller, Don [ORNL

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  3. A double-Bragg detector with digital signal processing for the event-by-event study of fission in actinide nuclei

    Science.gov (United States)

    Frost, R. J. W.; Smith, A. G.

    2016-09-01

    In the current paper, a windowless double-Bragg chamber incorporating full signal digitisation has been developed for the purpose of studying the energy (E), mass (A), charge (Z) and angular distributions (θ, Φ) of nuclei generated by fission. This device measures E for each fission fragment by collection of the charge produced during ionisation of the fill gas. Subsequent digitisation of the signals from each of two anodes yields information on dE/dx, as well as electron collection time, which can be further used for polar angle (θ) determination. Frisch-grid and cathode signals are also digitised and are used both for anode signal correction and to produce further information on θ. To verify the operation of this detector, three angular determination techniques from the literature were implemented, and the results were found to be consistent with the referenced paper. Current results from the spontaneous fission of 252Cf are presented.

  4. Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo [ed.

    1992-06-15

    Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10{sup {minus}5} eV to 20 MeV. Almost of the cross section data reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in order tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum.

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

    CERN Document Server

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Haag, Nils-Holger

    2013-10-09

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

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

    Science.gov (United States)

    Koehl, Michael A.

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

  9. Novel diglycolamide functionalized calix[4]arenes for actinide extraction and supported liquid membrane studies: Part II. Role of substituents in the pendent arms and mass transfer modeling I

    NARCIS (Netherlands)

    Ansari, S.A.; Mohapatra, P.K.; Iqbal, M.; Kandwal, P.; Huskens, J.; Verboom, W.

    2013-01-01

    Several calix[4]arene-functionalized diglycolamide (C4DGA) ligands were evaluated for the extraction as well as supported liquid membrane (SLM) transport of actinides and fission product elements from nitric acid feed solutions. The extraction efficiency of the C4DGA ligands for Am(III) was orders o

  10. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    Energy Technology Data Exchange (ETDEWEB)

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in

  11. Theoretical Description of the Fission Process

    Energy Technology Data Exchange (ETDEWEB)

    Witold Nazarewicz

    2009-10-25

    Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nation’s nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic

  12. Fission products, activity calculation of spent-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Souka, N.; El-Hakiem, M.N.

    1981-01-01

    This work is a scrutiny of the activity of burned up fuel elements of the ET-RR-1. A knowledge of this activity as well as its decay with time is quite helpful in shielding calculations related to construction purposes of hot facilities. The present treatment is based on a knowledge of: fuel composition, percentage burnup, and fission yields of produced isotopes. Cooling periods ranging from 1 hr to 10 years were considered.

  13. Fission Product Removal From Spent Oxide Fuel By Head-End Processing

    Energy Technology Data Exchange (ETDEWEB)

    B. R. Westphal; K. J. Bateman; R. P. Lind; K. L. Howden; G. D. Del Cul

    2005-10-01

    The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. Experiments have been performed with irradiated oxide fuel to evaluate the removal of fission products. During these experiments, operating parameters such as temperature and pressure have been varied to discern their effects on the behavior of specific fission products. In general, the extent of removal increases with increasing operating temperature and decreasing pressure. Removal efficiencies as high as 98% have been achieved during testing. Given the results of testing, an explanation of the likely fission product species being removed during the test program is also provided. In addition, experiments have been performed with other oxidative gases (steam and ozone) on surrogates to determine their potential benefit for removal of fission products.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Ha, Kwang Soon; Kim, Dong Ha; Kim, Tae Woon [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

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

  15. The rate of decay of fresh fission products from a nuclear reactor

    Science.gov (United States)

    Dolan, David J.

    Determining the rate of decay of fresh fission products from a nuclear reactor is complex because of the number of isotopes involved, different types of decay, half-lives of the isotopes, and some isotopes decay into other radioactive isotopes. Traditionally, a simplified rule of 7s and 10s is used to determine the dose rate from nuclear weapons and can be to estimate the dose rate from fresh fission products of a nuclear reactor. An experiment was designed to determine the dose rate with respect to time from fresh fission products of a nuclear reactor. The experiment exposed 0.5 grams of unenriched Uranium to a fast and thermal neutron flux from a TRIGA Research Reactor (Lakewood, CO) for ten minutes. The dose rate from the fission products was measured by four Mirion DMC 2000XB electronic personal dosimeters over a period of six days. The resulting dose rate following a rule of 10s: the dose rate of fresh fission products from a nuclear reactor decreases by a factor of 10 for every 10 units of time.

  16. Immobilization of noble metal fission products in a metallic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Frank, S. M.; Bateman, K.; Marsden, K. C.; Keiser, D. D.; O' Holleran, T. P.; Hahn, P. A. [Idaho National Laboratory, Boise (United States)

    2008-08-15

    Development of the metallic waste form for the consolidation of spent-fuel cladding and the immobilization of specific fission-product radionuclides occurred as part of the larger Electrometallurgical Treatment Research and Demonstration conducted by Argonne National Laboratory for the U.S. Department of Energy from 1996 to 1999. The Global Nuclear Energy Partnership (GNEP) proposal for advanced reprocessing of spent nuclear fuel also proposes to combine recovered fission-product technetium and other transition metal fission products, primarily the undissolved solid (UDS) residue from the dissolver vessels, into a metallic, high-level waste form for geological disposal. This approach is similar to the production of the MWF produced during the treatment of spent EBR-II fuel at the INL.

  17. Accuracy Improvement of Neutron Nuclear Data on Minor Actinides

    Directory of Open Access Journals (Sweden)

    Harada Hideo

    2015-01-01

    Full Text Available Improvement of accuracy of neutron nuclear data for minor actinides (MAs and long-lived fission products (LLFPs is required for developing innovative nuclear system transmuting these nuclei. In order to meet the requirement, the project entitled as “Research and development for Accuracy Improvement of neutron nuclear data on Minor ACtinides (AIMAC” has been started as one of the “Innovative Nuclear Research and Development Program” in Japan at October 2013. The AIMAC project team is composed of researchers in four different fields: differential nuclear data measurement, integral nuclear data measurement, nuclear chemistry, and nuclear data evaluation. By integrating all of the forefront knowledge and techniques in these fields, the team aims at improving the accuracy of the data. The background and research plan of the AIMAC project are presented.

  18. Statistical model calculations of pre-scission neutron multiplicity for the heavy ion induced fusion-fission reactions with actinide target 232Th

    Directory of Open Access Journals (Sweden)

    Thakur Meenu

    2015-01-01

    Full Text Available The reaction mechanism of 19F + 232Th and 28Si + 232Th systems populating the near-super-heavy compound nuclei 251Es and 260Rf respectively are investigated using neutron multiplicity as a probe. The prescission neutron multiplicities of these compound nuclei are calculated at different excitation energies using a statistical model code. These calculations are performed using the Bohr-Wheeler transition state fission width as well as the dissipative dynamical fission width based on the Kramers’ prescription. For 19F + 232Th system, the measured yield of pre-scission is compared with the statistical model calculations for the decay of a compound nucleus in the excitation energy range of 54-90 MeV. The comparison between the measured and the calculated values indicates that the Bohr-Wheeler fission width underestimates the pre-scission neutron yield and a large amount of dissipation strength is required to reproduce the experimental pre-scission neutron multiplicities. The excitation energy dependence of the fitted values of the dissipation coefficient is also discussed. In addition, exploratory statistical model calculations of pre-scission neutron multiplicity for the 28Si + 232Th system are presented in the above range of excitation energy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

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

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

    Directory of Open Access Journals (Sweden)

    Äystö J.

    2012-02-01

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

  1. Pyrometallurgical processes for recovery of actinide elements

    Energy Technology Data Exchange (ETDEWEB)

    Battles, J.E.; Laidler, J.J.; McPheeters, C.C.; Miller, W.E.

    1994-01-01

    A metallic fuel alloy, nominally U-20-Pu-lOZr, is the key element of the Integral Fast Reactor (IFR) fuel cycle. Metallic fuel permits the use of an innovative, simple pyrometallurgical process, known as pyroprocessing, (the subject of this report), which features fused salt electrorefining of the spent fuel. Electrorefining separates the actinide elements from fission products, without producing a separate stream of plutonium. The plutonium-bearing product is contaminated with higher actinides and with a minor amount of rare earth fission products, making it diversion resistant while still suitable as a fuel material in the fast spectrum of the IFR core. The engineering-scale demonstration of this process will be conducted in the refurbished EBR-II Fuel Cycle Facility, which has entered the start-up phase. An additional pyrometallurgical process is under development for extracting transuranic (TRU) elements from Light Water Reactor (LWR) spent fuel in a form suitable for use as a feed to the IFR fuel cycle. Four candidate extraction processes have been investigated and shown to be chemically feasible. The main steps in each process are oxide reduction with calcium or lithium, regeneration of the reductant and recycle of the salt, and separation of the TRU product from the bulk uranium. Two processes, referred to as the lithium and salt transport (calcium reductant) processes, have been selected for engineering-scale demonstration, which is expected to start in late 1993. An integral part of pyroprocessing development is the treatment and packaging of high-level waste materials arising from the operations, along with the qualification of these waste forms for disposal in a geologic repository.

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

    Energy Technology Data Exchange (ETDEWEB)

    Katakura, J. (Japan Atomic Energy Research Inst., Tokai-mura, Naka-gun, Ibaraki-ken (Japan)); England, T.R. (Los Alamos National Lab., NM (United States))

    1991-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-15

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

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

    Science.gov (United States)

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

    2013-03-01

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

  5. Neutron-induced fission cross sections of 233U and 243Am in the energy range 0.5 Mev En 20 MeV @ n_TOF

    CERN Document Server

    Belloni, F; Milazzo, P M; Calviani, M; Colonna, N; Mastinu, P; Abbondanno, U; Aerts, G; Álvarez, H; Álvarez-Velarde, F; Andriamonje, S; Andrzejewski, J; Assimakopoulos, P; Audouin, L; Badurek, G; Baumann, P; Becvár, F; Berthoumieux, E; Calviño, F; Cano-Ott, D; Capote, R; Carrapiço, C; Cennini, P; Chepel, V; Chiaveri, E; Cortes, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillmann, I; Domingo-Pardo, C; Dridi, W; Duran, I; Eleftheriadis, C; Embid-Segura, M; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Fujii, K; Furman, W; Goncalves, I; González-Romero, E; Gramegna, F; Guerrero, C; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martinez, A; Igashira, M; Jericha, E; Käppeler, F; Kadi, Y; Karadimos, D; Karamanis, D; Kerveno, M; Koehler, P; Kossionides, E; Krticka, M; Lampoudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marrone, S; Martínez, T; Massimi, C; Mengoni, A; Moreau, C; Mosconi, M; Neves, F; Oberhummer, H; O'Brien, S; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Pigni, M T; Plag, R; Plompen, A; Plukis, A; Poch, A; Praena, J; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rubbia, C; Rudolf, G; Rullhusen, P; Salgado, J; Santos, C; Sarchiapone, L; Savvidis, I; Stephan, C; Tagliente, G; Tain, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vazl, P; Ventura, A; Villamarin, D; Vincente, M C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wiescher, M; Wisshak, K

    2011-01-01

    Neutron-induced fission cross-sections of actinides have been recently measured at the neutron time of flight facility n_TOF at CERN in the frame of a research project involving isotopes relevant for nuclear astrophysics and nuclear technologies. Fission fragments are detected by a gas counter with good discrimination between nuclear fission products and background events. Neutron-induced fission cross-sections of 233U and 243Am were determined relative to 235U. The present paper reports the results obtained at neutron energies between 0.5 and 20 MeV.

  6. Detection of Actinides via Nuclear Isomer De-Excitation

    Energy Technology Data Exchange (ETDEWEB)

    Francy, Christopher J. [Oregon State Univ., Corvallis, OR (United States)

    2009-07-01

    This dissertation discusses a data collection experiment within the Actinide Isomer Identification project (AID). The AID project is the investigation of an active interrogation technique that utilizes nuclear isomer production, with the goal of assisting in the interdiction of illicit nuclear materials. In an attempt to find and characterize isomers belonging to 235U and its fission fragments, a 232Th target was bombarded with a monoenergetic 6Li ion beam, operating at 45 MeV.

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

    CERN Document Server

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

    2002-01-01

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

  8. Diffusion of fission products and radiation damage in SiC

    Science.gov (United States)

    Malherbe, Johan B.

    2013-11-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

  10. HTGR accident initiation and progression analysis status report. Volume V. AIPA fission product source terms

    Energy Technology Data Exchange (ETDEWEB)

    Alberstein, D.; Apperson, C.E. Jr.; Hanson, D.L.; Myers, B.F.; Pfeiffer, W.W.

    1976-02-01

    The primary objective of the Accident Initiation and Progression Analysis (AIPA) Program is to provide guidance for high-temperature gas-cooled reactor (HTGR) safety research and development. Among the parameters considered in estimating the uncertainties in site boundary doses are uncertainties in fission product source terms generated under normal operating conditions, i.e., fuel body inventories, circulating coolant activity, total plateout activity in the primary circuit, and plateout distributions. The volume presented documents the analyses of these source term uncertainties. The results are used for the detailed consequence evaluations, and they provide the basis for evaluation of fission products important for HTGR maintenance and shielding.

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

    CERN Document Server

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

    2013-01-01

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

  12. Chemical forms of solid fission products in the irradiated uranium—plutonium mixed nitride fuel

    Science.gov (United States)

    Arai, Yasuo; Maeda, Atsushi; Shiozawa, Ken-ichi; Ohmichi, Toshihiko

    1994-06-01

    Chemical forms of solid fission products in the irradiated (U, Pu)N fuel were estimated by both thermodynamic equilibrium calculation and electron microprobe analysis on burnup simulated samples prepared by carbothermic reduction. Besides the MX type matrix phase dissolving zirconium, niobium, yttrium and rare earth elements, the existence of two kinds of inclusion was recognized. One is URu 3 type intermetallic compound constituted by uranium and platinum group elements. The other is an alloy containing molybdenum as a principal constituent. Furthermore, the swelling rate due to solid fission products precipitation was evaluated to be about 0.5% per %FIMA.

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

    Energy Technology Data Exchange (ETDEWEB)

    Asner, D.M.; Burns, K.; Campbell, L.W.; Greenfield, B.; Kos, M.S., E-mail: markskos@gmail.com; Orrell, J.L.; Schram, M.; VanDevender, B.; Wood, L.S.; Wootan, D.W.

    2015-03-11

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

  15. Theoretical Description of the Fission Process

    Energy Technology Data Exchange (ETDEWEB)

    Witold Nazarewicz

    2003-07-01

    The main goals of the project can be summarized as follows: Development of effective energy functionals that are appropriate for the description of heavy nuclei. Our goal is to improve the existing energy density (Skyrme) functionals to develop a force that will be used in calculations of fission dynamics. Systematic self-consistent calculations of binding energies and fission barriers of actinide and trans-actinide nuclei using modern density functionals. This will be followed by calculations of spontaneous fission lifetimes and mass and charge divisions using dynamic adiabatic approaches based on the WKB approximation. Investigate novel microscopic (non-adiabatic) methods to study the fission process.

  16. Fission Measurements with Dance

    Science.gov (United States)

    Jandel, M.; Bredeweg, T. A.; Fowler, M. M.; Bond, E. M.; Chadwick, M. B.; Clement, R. R.; Couture, A.; O'Donnell, J. M.; Haight, R. C.; Keksis, A. L.; Reifarth, R.; Rundberg, R. S.; Ullmann, J. L.; Vieira, D. J.; Wilhelmy, J. B.; Wouters, J. M.; Agvaanluvsan, U.; Dashdorj, D.; Macri, R. A.; Parker, W. E.; Wilk, P. A.; Wu, C. Y.; Becker, J. A.; Angell, C. T.; Tonchev, A. P.; Baker, J. D.

    2008-08-01

    Neutron capture cross section measurements on actinides are complicated by the presence of neutron-induced fission. An efficient fission tagging detector used in coincidence with the Detector for Advanced Neutron Capture Experiments (DANCE) provides a powerful tool in undertaking simultaneous measurements of (n,γ) and (n,f) cross sections. Preliminary results on 235U(n,γ) and (n,f) and 242mAm(n,f) cross sections measured with DANCE and a custom fission-tagging parallel plate avalanche counter (PPAC) are presented. Additional measurements of γ-ray cluster multiplicity distributions for neutron-induced fission of 235U and 242mAm and spontaneous fission of 252Cf are shown, as well as γ-ray energy and average γ-ray energy distributions.

  17. Assessment of SFR fuel pin performance codes under advanced fuel for minor actinide transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Bouineau, V.; Lainet, M.; Chauvin, N.; Pelletier, M. [French Alternative Energies and Atomic Energy Commission - CEA, CEA Cadarache, DEN/DEC/SESC, 13108 Saint Paul lez Durance (France); Di Marcello, V.; Van Uffelen, P.; Walker, C. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D- 76344 Eggenstein-Leopoldshafen (Germany)

    2013-07-01

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. In the SUPERFACT Experiment four different oxide fuels containing high and low concentrations of {sup 237}Np and {sup 241}Am, representing the homogeneous and heterogeneous in-pile recycling concepts, were irradiated in the PHENIX reactor. The behavior of advanced fuel materials with minor actinide needs to be fully characterized, understood and modeled in order to optimize the design of this kind of fuel elements and to evaluate its performances. This paper assesses the current predictability of fuel performance codes TRANSURANUS and GERMINAL V2 on the basis of post irradiation examinations of the SUPERFACT experiment for pins with low minor actinide content. Their predictions have been compared to measured data in terms of geometrical changes of fuel and cladding, fission gases behavior and actinide and fission product distributions. The results are in good agreement with the experimental results, although improvements are also pointed out for further studies, especially if larger content of minor actinide will be taken into account in the codes. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa; Oh, Soo-Youl [Korea Atomic Energy Research Institute, Yusong, Taejon (Korea)

    2000-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

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

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

    NARCIS (Netherlands)

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

    2015-01-01

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

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

    Indian Academy of Sciences (India)

    S Kailas; M Hemalatha; A Saxena

    2015-09-01

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

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

    CERN Document Server

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-09-01

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

  4. Axisymmetric Magnetic Mirror Fusion-Fission Hybrid

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Martovetsky, N. N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Molvik, A. W. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Ryutov, D. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Simonen, T. C. [Univ. of California, Berkeley, CA (United States)

    2011-05-13

    The achieved performance of the gas dynamic trap version of magnetic mirrors and today’s technology we believe are sufficient with modest further efforts for a neutron source for material testing (Q=Pfusion/Pinput~0.1). The performance needed for commercial power production requires considerable further advances to achieve the necessary high Q>>10. An early application of the mirror, requiring intermediate performance and intermediate values of Q~1 are the hybrid applications. The Axisymmetric Mirror has a number of attractive features as a driver for a fusion-fission hybrid system: geometrical simplicity, inherently steady-state operation, and the presence of the natural divertors in the form of end tanks. This level of physics performance has the virtue of low risk and only modest R&D needed and its simplicity promises economy advantages. Operation at Q~1 allows for relatively low electron temperatures, in the range of 4 keV, for the DT injection energy ~ 80 keV. A simple mirror with the plasma diameter of 1 m and mirror-to-mirror length of 35 m is discussed. Simple circular superconducting coils are based on today’s technology. The positive ion neutral beams are similar to existing units but designed for steady state. A brief qualitative discussion of three groups of physics issues is presented: axial heat loss, MHD stability in the axisymmetric geometry, microstability of sloshing ions. Burning fission reactor wastes by fissioning actinides (transuranics: Pu, Np, Am, Cm, .. or just minor actinides: Np, Am, Cm, …) in the hybrid will multiply fusion’s energy by a factor of ~10 or more and diminish the Q needed to less than 1 to overcome the cost of recirculating power for good economics. The economic value of destroying actinides by fissioning is rather low based on either the cost of long-term storage or even deep geologic disposal so most of the revenues of hybrids will come from electrical power. Hybrids that obtain revenues from

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

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

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

  6. A Study on Fission Product Model Comparison between MAAP4 and MAAP5

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Tae-young; Seo, Mi Ro [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The newly added safety goal required that the sum of the accident frequency that the release of the radioactive nuclide Cs-137 to environment exceeds the 100TBq should be less than 1.0E-6/RY. This requirement is known to be come from the provision for preventing the long term ground contamination due to the release of radioactive material. Validation of this standard was performed by many researchers recently. In the outlook of Cs-137, the mass of Cs-137 correspondent with the 100TBq is calculated as 32g. However, during the severe accident, if the containment has been failed, it is generally expected that the mass of Cs-137 released to the environment is more than 1kg for most accident sequences. The purpose of this study compare fission product model in MAAP4 and MAAP5. So the same accident will be simulated as MAAP4 and MAAP5. And will compare fission product release fraction. This will help to improvements obtained to meet the regulatory requirements of Cs-137. This paper was a comparison of MAAP4's fission product models with those of MAAP5. And this paper simulated the station blackout accident to compare MAAP4 and MAAP5 fission product release fraction. So far Level 2 PSA analysis used MAAP4. And this result failed to meet the regulatory requirements of Cs-137 up to now. Fission product release fraction calculated by MAAP5 is more conservative than that calculated by MAAP4. Therefore, using MAAP5 is more difficult to meet the requirements of Cs-137. Thus, Level 1 PSA analysis must find ways to reduce CDF and Level 2 PSA analysis must find ways to reduce CFF in order to meet regulatory requirements. Not only, it seems to be required a study on the possible safety systems to alleviate the containment failure after the core damage.

  7. Determination of uranium fission product interference factor for molybdenum quantification by instrumental neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro Junior, Ibere S.; Saiki, Mitiko; Genezini, Frederico A.; Zahn, Guilherme S., E-mail: ibere@usp.br, E-mail: mitiko@ipen.br, E-mail: fredzini@ipen.br, E-mail: gzahn@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    Instrumental Neutron Activation Analysis (INAA) is a technique that provides high precision and accuracy results for the concentration determinations of several elements in different kinds of matrices. However, if the sample contains high uranium concentration in their composition, INAA can provide inaccurate results due to uranium fission product interferences. The molybdenum is one of these elements that suffers interference by uranium fission product, because the {sup 99}Mo radioisotope formed by {sup 98}Mo neutron capture, used in INAA, is the same that formed in the uranium fission. This kind of interference can be solved by separation of uranium before irradiation or by determining the uranium interference factor to the radioisotope of interest and applying the correction. The present study aims at the following: (1) determination of the Mo interference factor (F{sub Exp}{sup Mo}) due to the uranium fission product {sup 99}Mo by irradiating standards of Mo and U with known masses of these elements (experimental interference factor); (2) determination of the theoretical F{sub Th}{sup Mo}, in this case it was necessary to determine the epithermal to thermal neutron flux ratio and use the reported nuclear parameters; (3) comparison of the results of the interference factor obtained with values reported in the literature. The interference factor for Mo analysis was obtained in a position 14b shelf 3 of the IEA-R1 nuclear research reactor. (author)

  8. Mechanistic approach for nitride fuel evolution and fission product release under irradiation

    Science.gov (United States)

    Dolgodvorov, A. P.; Ozrin, V. D.

    2017-01-01

    A model for describing uranium-plutonium mixed nitride fuel pellet burning was developed. Except fission products generating, the model includes impurities of oxygen and carbon. Nitrogen behaviour in nitride fuel was analysed and the nitrogen chemical potential in solid solution with uranium-plutonium nitride was constructed. The chemical program module was tested with the help of thermodynamic equilibrium phase distribution calculation. Results were compared with analogous data in literature, quite good agreement was achieved, especially for uranium sesquinitride, metallic species and some oxides. Calculation of a process of nitride fuel burning was also conducted. Used mechanistic approaches for fission product evolution give the opportunity to find fission gas release fractions and also volumes of intergranular secondary phases. Calculations present that the most massive secondary phases are the oxide and metallic phases. Oxide phase contain approximately 1 % wt of substance over all time of burning with slightly increasing of content. Metallic phase has considerable rising of mass and by the last stage of burning it contains about 0.6 % wt of substance. Intermetallic phase has less increasing rate than metallic phase and include from 0.1 to 0.2 % wt over all time of burning. The highest element fractions of released gaseous fission products correspond to caesium and iodide.

  9. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  10. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

    2016-11-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  11. Mass yield distributions of fission products from photo-fission of {sup 238}U induced by 11.5-17.3 MeV bremsstrahlung

    Energy Technology Data Exchange (ETDEWEB)

    Naik, H.; Goswami, A. [Bhabha Atomic Research Centre, Radiochemistry Division, Mumbai (India); Carrel, Frederick; Laine, Frederic; Sari, Adrien [SAPHIR Facility, Gif-sur-Yvette (France); Kim, G.N. [Kyungpook National University, Department of Physics, Daegu (Korea, Republic of); Normand, S. [Laboratory of Sensors and Elctronics Architectures CEA, Gif-sur-Yvette (France)

    2013-07-15

    The yields of various fission products in the 11.5, 13.4, 15.0 and 17.3 MeV bremsstrahlung-induced fission of {sup 238}U have been determined by recoil catcher and an off-line {gamma}-ray spectrometric technique using the electron linac, SAPHIR at CEA, Saclay, France. The mass yield distributions were obtained from the fission product yields using charge-distribution corrections. The peak-to-valley (P/V) ratio, average light mass (left angle A{sub L} right angle) and heavy mass (left angle A{sub H} right angle) and average number of neutrons (left angle v right angle) in the bremsstrahlung-induced fission of {sup 238}U at different excitation energies were obtained from the mass yield data. From the present and literature data in the {sup 238}U ({gamma}, f) and {sup 238}U (n, f) reactions at various energies, the following observations were obtained: i) The mass yield distributions in the {sup 238}U ({gamma}, f) reaction at various energies of the present work are double-humped, similar to those of the {sup 238}U (n, f) reaction of comparable excitation energy. ii) The yields of fission products for A = 133-134, A = 138-140, and A = 143-144 and their complementary products in the {sup 238}U ({gamma}, f) reaction are higher than other fission products due to the nuclear structure effect. iii) The yields of fission products for A = 133-134 and their complementary products are slightly higher in the {sup 238}U ({gamma}, f) than in the {sup 238}U (n, f), whereas for A = 138-140 and 143-144 and their complementary products are comparable. iv) With excitation energy, the increase of yields of symmetric products and the decrease of the peak-to-valley (P/V ratio in the {sup 238}U ({gamma}, f) reaction is similar to the {sup 238}U (n, f) reaction. v) The increase of left angle v right angle with excitation energy is also similar between the {sup 238}U ({gamma}, f) and {sup 238}U (n, f) reactions. However, it is surprising to see that the left angle A{sub L} right angle and

  12. Mass yield distributions of fission products from photo-fission of 238U induced by 11.5-17.3 MeV bremsstrahlung

    Science.gov (United States)

    Naik, H.; Carrel, Frédérick; Kim, G. N.; Laine, Frédéric; Sari, Adrien; Normand, S.; Goswami, A.

    2013-07-01

    The yields of various fission products in the 11.5, 13.4, 15.0 and 17.3 MeV bremsstrahlung-induced fission of 238U have been determined by recoil catcher and an off-line γ-ray spectrometric technique using the electron linac, SAPHIR at CEA, Saclay, France. The mass yield distributions were obtained from the fission product yields using charge-distribution corrections. The peak-to-valley ( P/ V ratio, average light mass () and heavy mass () and average number of neutrons () in the bremsstrahlung-induced fission of 238U at different excitation energies were obtained from the mass yield data. From the present and literature data in the 238U ( γ, f ) and 238U ( n, f ) reactions at various energies, the following observations were obtained: i) The mass yield distributions in the 238U ( γ, f ) reaction at various energies of the present work are double-humped, similar to those of the 238U ( n, f ) reaction of comparable excitation energy. ii) The yields of fission products for A = 133-134, A = 138-140, and A = 143-144 and their complementary products in the 238U ( γ, f) reaction are higher than other fission products due to the nuclear structure effect. iii) The yields of fission products for A = 133-134 and their complementary products are slightly higher in the 238U ( γ, f ) than in the 238U ( n, f ) , whereas for A = 138-140 and 143-144 and their complementary products are comparable. iv) With excitation energy, the increase of yields of symmetric products and the decrease of the peak-to-valley ( P/ V ratio in the 238U ( γ, f) reaction is similar to the 238U ( n, f) reaction. v) The increase of with excitation energy is also similar between the 238U ( γ, f ) and 238U ( n, f) reactions. However, it is surprising to see that the and values with excitation energy behave entirely differently from the 238U ( γ, f ) and 238U ( n, f ) reactions.

  13. Fission and Properties of Neutron-Rich Nuclei - Proceedings of the Second International Conference

    Science.gov (United States)

    Hamilton, J. H.; Phillips, W. R.; Carter, H. K.

    Long Range Alpha Emission Probabilities in the Thermal and Resonance for 239Pu * Spectroscopy of Selected Fission Fragments * Gamma-Ray Spectra of Individual Fission Products * Gamma Spectroscopy of the Transfermium Nucleus Nobelium-254 * Summary of the Yield Values from Fast Fission of 233U Measured of Studsvik * Decay Studies of Neutron-Rich Isotopes Produced in Proton-Induced Fission of Actinides * Understanding of the Fission Process From the Deformation Properties of Fissioning Nuclei * Heavy Clusterization in 252Cf: An Application of the U(3) Selection Rule * Synthesis of Superheavy Elements with Three-Dimensional Fluctuation-Dissipation Dynamical Model * Asymmetric Fission Along Nuclear Lattice Planes * Experimental Study of Correlations Between Fission Neutron Multiplicity, Mass and Kinetic Energy of Fission Fragments From Spontaneous Fission of 252Cf, 244Cm and 248Cm * Super- and Hyper-Deformed Isomeric States and Long-Lived Superheavy Elements * Systematic Investigation of the Neutron Induced Fission Reaction 235, 238U and 237Np below 6 MeV * Studying Iso-chains with RNBs * Cold Fission Studies with Large Detector Arrays * Single-Particle States in Transcurium Nuclei * Shell Effects and Fission of Superheavy Nuclei at Low Excitation Energies * Experiments on the Synthesis of Superheavy Nuclei in 48Ca-Induced Reactions * GSI Experiments on the Synthesis of Superheavy Elements - Results and Plans for the Future * Spectroscopy of Refractory Fission Products at IGISOL * Ternary Fission Induced by Polarized Neutrons * Neutron Decay of Ternary Particles in Spontaneous Fission of 252Cf * New Fission Modes * Alpha Clustering and Ternary Fission * Interplay of Fusion and Fission Dynamics * Applications of the Hartree Bogoliubov Model to Nuclei with Large Isospin Values * Structure of Neutron-Rich Pd Isotopes * Systematic Studies of Fission Saddle-Point Shapes and Their Relation to the Maxima of the Fission-Fragment Mass and Kinetic-Energy Distributions * Limits of

  14. Effect of Dissipation, Angular Momentum and Velocity Along the Fission Direction on the Production of Heavy Neck Fragments During Fission

    Science.gov (United States)

    Yanez, R.; de Souza, R. T.; Bredeweg, T. A.; Davin, B.; Kwiatkowski, K.; Viola, V. E., Jr.; Cârjan, N.

    1998-04-01

    We have coupled the Los Alamos finite range dynamical model of fission with a Coulomb trajectory model in order to understand recently observed trends in experimental measurements of neck emission during fission. In the reaction of 22 A MeV ^12C and ^232Th a significant yield of relatively heavy neck fragments (Z=9-13), which originate from peripheral collisions, was observed. Within the macroscopic model we study the effect of dissipation, angular momentum and an initial velocity along the fission direction on the formation of a third middle fragment. As compared to binary fission, ternary fission is associated with a more elongated scission configuration. Following scission, a classical Coulomb trajectory model is used to calculate the final kinetic energies of the fragments given the breakup geometry, and the radial and rotational velocities of the fission fragments as predicted by the macroscopic model. The calculated final kinetic energies are compared to the measured kinetic energies of the experimentally observed neck fragments.

  15. Innovative SANEX process for trivalent actinides separation from PUREX raffinate

    Energy Technology Data Exchange (ETDEWEB)

    Sypula, Michal

    2013-07-01

    a Zr masking agent in GANEX process where high content of Pu is expected. In this so called GANEX (Group Actinide Extraction) process all actinides are separated as a group from the lanthanides and fission products. Within the new ligands delivered by University of Twente four water-soluble and four lipophilic ones showed good An(III)/Ln(III) complexation thus they deserve further studies with possible implementation in already existing or new processes.

  16. Fission in a Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Younes, W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-26

    A three-year theory project was undertaken to study the fission process in extreme astrophysical environments, such as the crust of neutron stars. In the first part of the project, the effect of electron screening on the fission process was explored using a microscopic approach. For the first time, these calculations were carried out to the breaking point of the nucleus. In the second part of the project, the population of the fissioning nucleus was calculated within the same microscopic framework. These types of calculations are extremely computer-intensive and have seldom been applied to heavy deformed nuclei, such as fissioning actinides. The results, tools and methodologies produced in this work will be of interest to both the basic-science and nuclear-data communities.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from a LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., /sup 137/Cs, /sup 90/Sr, /sup 129/I, /sup 99/Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,..cap alpha..), (n,..gamma..), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R = 1.0 to 3.0) requirements. These studies also indicate that masses on the order of 1.0 g at densities of rho greater than or equal to 500.0 g/cm/sup 3/ are required for a practical fusion-based fission product transmutation system.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  19. Simulations of the stopping efficiencies of fission ion guides

    Science.gov (United States)

    Solders, Andreas; Al-Adili, Ali; Gorelov, Dmitry; Jansson, Kaj; Jokinen, Ari; Kolhinen, Veli; Lantz, Mattias; Mattera, Andrea; Moore, Ian; Nilsson, Niklas; Norlin, Martin; Penttilä, Heikki; Pomp, Stephan; Prokofiev, Alexander V.; Rakopoulos, Vasileios; Rinta-Antila, Sami; Simutkin, Vasily

    2017-09-01

    With the Ion Guide Isotope Separator On-Line (IGISOL) facility, located at the University of Jyväskylä, products of nuclear reactions are separated by mass. The high resolving power of the JYFLTRAP Penning trap, with full separation of individual nuclides, capacitates the study of nuclides far from the line of stability. For the production of neutron-rich medium-heavy nuclides, fissioning of actinides is a feasible reaction. This can be achieved with protons from an in-house accelerator or, alternatively, with neutrons through the addition of a newly developed Be(p,xn)-converter. The hereby-obtained fission products are used in nuclear data measurements, for example fission yields, nuclear masses, Q-values and decay spectroscopy. Prior to separation, the ionized reaction products are stopped in a helium-filled gas cell, referred to as the ion-guide. In this work we present simulations of the stopping of fission products in an ion guide developed for neutron-induced fission. The production and extraction rates are evaluated and compared against experimental values.

  20. FY2011 Annual Report for the Actinide Isomer Detection Project

    Energy Technology Data Exchange (ETDEWEB)

    Warren, Glen A.; Francy, Christopher J.; Ressler, Jennifer J.; Erikson, Luke E.; Tatishvili, Gocha; Hatarik, R.

    2011-10-01

    This project seeks to identify a new signature for actinide element detection in active interrogation. This technique works by exciting and identifying long-lived nuclear excited states (isomers) in the actinide isotopes and/or primary fission products. Observation of isomers in the fission products will provide a signature for fissile material. For the actinide isomers, the decay time and energy of the isomeric state is unique to a particular isotope, providing an unambiguous signature for SNM. This project entails isomer identification and characterization and neutron population studies. This document summarizes activities from its third year - completion of the isomer identification characterization experiments and initialization of the neutron population experiments. The population and decay of the isomeric state in 235U remain elusive, although a number of candidate gamma rays have been identified. In the course of the experiments, a number of fission fragment isomers were populated and measured [Ressler 2010]. The decays from these isomers may also provide a suitable signature for the presence of fissile material. Several measurements were conducted throughout this project. This report focuses on the results of an experiment conducted collaboratively by PNNL, LLNL and LBNL in December 2010 at LBNL. The measurement involved measuring the gamma-rays emitted from an HEU target when bombarded with 11 MeV neutrons. This report discussed the analysis and resulting conclusions from those measurements. There was one strong candidate, at 1204 keV, of an isomeric signature of 235U. The half-life of the state is estimated to be 9.3 {mu}s. The measured time dependence fits the decay time structure very well. Other possible explanations for the 1204-keV state were investigated, but they could not explain the gamma ray. Unfortunately, the relatively limited statistics of the measurement limit, and the lack of understanding of some of the systematic of the experiment, limit

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

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, G.P.

    1991-01-01

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

  2. Fission product transport and behavior during two postulated loss of flow transients in the air

    Energy Technology Data Exchange (ETDEWEB)

    Adams, J.P.; Carboneau, M.L.

    1991-01-01

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

  3. Fission product transport and behavior during two postulated loss of flow transients in the air

    Energy Technology Data Exchange (ETDEWEB)

    Adams, J.P.; Carboneau, M.L.

    1991-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-01

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

  5. A model for release of fission products from a breached fuel plate under wet storage

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, L.A.A.; Seerban, R.S.; Zeituni, C.A.; Silva, J.E.R. da; Silva, A.T. e; Castanheira, M.; Lucki, G.; Damy, M. de A.; Teodoro, C.A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: laaterre@ipen.br

    2007-07-01

    MTR fuel elements burned-up inside the core of nuclear research reactors are stored worldwide mainly under the water of storage pools. When cladding breach is present in one or more fuel plates of such elements, radioactive fission products are released into the storage pool water. This work proposes a model to describe the release mechanism considering the diffusion of nuclides of a radioactive fission product either through a postulated small cylindrical breach or directly from a large circular hole in the cladding. In each case, an analytical expression is obtained for the activity released into the water as a function of the total storage time of a breached fuel plate. Regarding sipping tests already performed at the IEA-R1 research reactor on breached MTR fuel elements, the proposed model correlates successfully the specific activity of {sup 137}Cs, measured as a function of time, with the evaluated size of the cladding breach. (author)

  6. Thermochemical prediction of chemical form distributions of fission products in LWR mixed oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1998-06-01

    Radial distribution of chemical forms of fission products (FPs) in LWR mixed oxide (MOX) fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The amounts of fission products generated in the fuel were calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. The effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that the radial distribution of chemical forms depends strongly on the LHR and the O/M ratio, and is not proportional to that of burnup. (author)

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

    CERN Document Server

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

    2011-01-01

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

  8. Uncertainties on decay heat power due to fission product data uncertainties; Incertitudes sur la puissance residuelle dues aux incertitudes sur les donnees de produits de fission

    Energy Technology Data Exchange (ETDEWEB)

    Rebah, J

    1998-08-01

    Following a reactor shutdown, after the fission process has completely faded out, a significant quantity of energy known as 'decay heat' continues to be generated in the core. The knowledge with a good precision of the decay heat released in a fuel after reactor shutdown is necessary for: residual heat removal for normal operation or emergency shutdown condition, the design of cooling systems and spent fuel handling. By the summation calculations method, the decay heat is equal to the sum of the energies released by individual fission products. Under taking into account all nuclides that contribute significantly to the total decay heat, the results from summation method are comparable with the measured ones. Without the complete covariance information of nuclear data, the published uncertainty analyses of fission products decay heat summation calculation give underestimated errors through the variance/covariance analysis in consideration of correlation between the basic nuclear data, we calculate in this work the uncertainties on the decay heat associated with the summation calculations. Contribution to the total error of decay heat comes from uncertainties in three terms: fission yields, half-lives and average beta and gamma decay energy. (author)

  9. Event-by-event study of neutron observables in spontaneous and thermal fission

    Energy Technology Data Exchange (ETDEWEB)

    Vogt, R; Randrup, J

    2011-09-14

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

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

    Science.gov (United States)

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

    2013-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Blanc, A., E-mail: blanc@ill.fr [Institut Laue-Langevin, Grenoble (France); Chebboubi, A.; Faust, H.; Jentschel, M. [Institut Laue-Langevin, Grenoble (France); Kessedjian, G. [LPSC CNRS/IN2P3, UJF Grenoble 1, INPG, Grenoble (France); Köster, U. [Institut Laue-Langevin, Grenoble (France); Materna, T.; Panebianco, S. [DSM/IRFU/SPhN, CEA Saclay (France); Sage, C. [LPSC CNRS/IN2P3, UJF Grenoble 1, INPG, Grenoble (France); Urban, W. [Institut Laue-Langevin, Grenoble (France)

    2013-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Institute of Scientific and Technical Information of China (English)

    WANG; Xiu-feng

    2013-01-01

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

  14. Enhancement of ASTEC and COCOSYS regarding fission product release during MCCI

    Energy Technology Data Exchange (ETDEWEB)

    Agethen, Kathrin [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2016-10-15

    The focus in this paper is on the enhancement of the fission product release model during molten core concrete interaction in the severe accident analysis codes ASTEC and COCOSYS. After both codes are harmonised and the model interaction as well as the input parameters are adapted, extended model approaches are implemented. These lead to an improvement of the release rates for selected semi-volatile species validated against the ACE tests under ex-vessel conditions.

  15. Fission induced by nucleons at intermediate energies

    CERN Document Server

    Meo, Sergio Lo; Massimi, Cristian; Vannini, Gianni; Ventura, Alberto

    2014-01-01

    Monte Carlo calculations of fission of actinides and pre-actinides induced by protons and neutrons in the energy range from 100 MeV to 1 GeV are carried out by means of a recent version of the Li\\`ege Intranuclear Cascade Model, INCL++, coupled with two different evaporation-fission codes, GEMINI++ and ABLA07. In order to reproduce experimental fission cross sections, model parameters are usually adjusted on available (p,f) cross sections and used to predict (n,f) cross sections for the same isotopes.

  16. Deposition of volatile fission products during commercial high-level waste vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, M.S.; Carter, J.G.

    1982-09-01

    High-level waste from spent commercial PWR fuel was vitrified during the Nuclear Waste Vitrification Project (NWVP), conducted at Pacific Northwest Laboratory in 1978 and 1979. The solidification equipment employed sintered metal filters and a conventional wet off-gas cleanup system to decontaminate the gases generated during waste processing. This report is a summary of a study completed under contract with U.S. Department of Energy (DOE) to determine the fate of volatile fission products in the filters and off-gas system. The study objectives were to determine (1) the extent of volatile fission product deposition within the sintered metal filters, (2) the distribution of fission products in the off-gas system, and (3) the factors that control the distribution. Such information is needed for the design and operation of future waste immobilization systems. The scope of the study included identification, preparation, and microprobe analyses of suitable filter samples and radiochemical analyses of off-gas system samples. The report describes the methods of sample selection and gathering, the analytical methods and results, and the interpretations of the results. Relationships to vitrification system operations are defined and operating conditions to minimize problems recommended. 37 figures, 14 tables.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Abrecht, David G.; Schwantes, Jon M.

    2015-03-03

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

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

    Directory of Open Access Journals (Sweden)

    Blanc A.

    2015-01-01

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

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

    Science.gov (United States)

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

    2016-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

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

    Science.gov (United States)

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

    2005-05-01

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

  3. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    Energy Technology Data Exchange (ETDEWEB)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-15

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl2Si2O8, BaAl2Si2O8and CsAlSi2O6 in the additional alumina-silica phase of the kernel.

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

    Energy Technology Data Exchange (ETDEWEB)

    Van der Ven, Anton [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Wang, Lumin [Univ. of Michigan, Ann Arbor, MI (United States); Taheri, Mitra [Drexel Univ., Philadelphia, PA (United States)

    2014-04-26

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

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

    Science.gov (United States)

    Abrecht, David G; Schwantes, Jon M

    2015-03-03

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

  6. ICP-MS analysis of fission product diffusion in graphite for High-Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Carter, Lukas M.

    Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-11-01

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

  8. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

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

    Energy Technology Data Exchange (ETDEWEB)

    Naik, H. [Bhabha Atomic Research Centre, Radiochemistry Division, Mumbai (India); Kyungpook National University, Department of Physics and Center for High Energy Physics, Daegu (Korea, Republic of); Kim, G.N.; Kim, K. [Kyungpook National University, Department of Physics and Center for High Energy Physics, Daegu (Korea, Republic of); Schwengner, R.; John, R.; Massarczyk, R.; Junghans, A.; Wagner, A. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Goswami, A. [Bhabha Atomic Research Centre, Radiochemistry Division, Mumbai (India)

    2015-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dickens, J.K.; Emery, J.F.; Love, T.A.; McConnell, J.W.; Northcutt, K.J.; Peelle, R.W.; Weaver, H.

    1977-10-01

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

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

    CERN Document Server

    Herndon, J M; Edgerley, Dennis A.

    2005-01-01

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

  12. Experimental investigations relevant for hydrogen and fission product issues raised by the Fukushima accident

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-15

    The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the

  13. Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

    Energy Technology Data Exchange (ETDEWEB)

    B.R. Westphal; J.J. Park; J.M. Shin; G.I. Park; K.J. Bateman; D.L. Wahlquist

    2008-07-01

    A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.

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

    Institute of Scientific and Technical Information of China (English)

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

    2013-01-01

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

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

    Science.gov (United States)

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

    2016-08-01

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

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

    Science.gov (United States)

    Mushtaq, A

    2011-04-01

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

  17. Isomer production ratios and the angular momentum distribution of fission fragments

    Science.gov (United States)

    Stetcu, I.; Talou, P.; Kawano, T.; Jandel, M.

    2013-10-01

    Latest generation fission experiments provide an excellent testing ground for theoretical models. In this contribution we compare the measurements for 235U(nth,f), obtained with the Detector for Advanced Neutron Capture Experiments (DANCE) calorimeter at Los Alamos Neutron Science Center (LANSCE), with our full-scale simulation of the primary fragment de-excitation, using the recently developed cgmf code, based on a Monte Carlo implementation of the Hauser-Feshbach theoretical model. We compute the isomer ratios as a function of the initial angular momentum of the fission fragments, for which no direct information exists. Comparison with the available experimental data allows us to determine the initial spin distribution. We also study the dependence of the isomer ratio on the knowledge of the low-lying discrete spectrum input for nuclear fission reactions, finding a high degree of sensitivity. Finally, in the same Hauser-Feshbach approach, we calculate the isomer production ratio for thermal neutron capture on stable isotopes, where the initial conditions (spin, excitation energy, etc.) are well understood. We find that with the current parameters involved in Hauser-Feshbach calculations, we obtain up to a factor of 2 deviation from the measured isomer ratios.

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

    Energy Technology Data Exchange (ETDEWEB)

    Jason M. Harp; Paul A. Demkowicz

    2014-10-01

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

  19. Two neutron correlations in photo-fission

    Science.gov (United States)

    Dale, D. S.; Kosinov, O.; Forest, T.; Burggraf, J.; Stave, S.; Warren, G.; Starovoitova, V.

    2016-09-01

    A large body of experimental work has established the strong kinematical correlation between fission fragments and fission neutrons. Here, we report on the progress of investigations of the potential for strong two neutron correlations arising from the nearly back-to-back nature of the two fission fragments that emit these neutrons in the photo-fission process. In initial measurements, a pulsed electron linear accelerator was used to generate bremsstrahlung photons that impinged upon an actinide target, and the energy and opening angle distributions of coincident neutrons were measured using a large acceptance neutron detector array. A planned comprehensive set of measurements of two neutron correlations in the photo-fission of actinides is expected to shed light on several fundamental aspects of the fission process including the multiplicity distributions associated with the light and heavy fission fragments, the nuclear temperatures of the fission fragments, and the mass distribution of the fission fragments as a function of energy released. In addition to these measurements providing important nuclear data, the unique kinematics of fission and the resulting two neutron correlations have the potential to be the basis for a new tool to detect fissionable materials. A key technical challenge of this program arises from the need to perform coincidence measurements with a low duty factor, pulsed electron accelerator. This has motivated the construction of a large acceptance neutron detector array, and the development of data analysis techniques to directly measure uncorrelated two neutron backgrounds.

  20. Effects of fissioning nuclei distributions on fragment mass distributions for high energy fission

    Directory of Open Access Journals (Sweden)

    Rossi P C R

    2012-02-01

    Full Text Available We study the effects of fissioning nuclei mass- and energy-distributions on the formation of fragments for fission induced by high energy probes. A Monte Carlo code called CRISP was used for obtaining mass distributions and spectra of the fissioning nuclei for reactions induced by 660 MeV protons on 241Am and on 239Np, by 500 MeV protons on 208Pb, and by Bremsstrahlung photons with end-point energies at 50 MeV and 3500 MeV on 238U. The results show that even at high excitation energies, asymmetric fission may still contribute significantly to the fission cross section of actinide nuclei, while it is the dominante mode in the case of lead. However, more precise data for high energy fission on actinide are necessary in order to allow definite conclusions.

  1. Development of Metallic Fuels for Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

    2015-09-01

    and 60%. In general, the performance of all of these substantially disparate metallic fuel alloys has been observed to be excellent, and their irradiation behaviors are generally consistent with historic norms for metallic fuels without minor actinide additions and having lower Pu or Zr contents. Future work is being undertaken with a view toward increasing the burnup potential of metallic fuels even more. Design innovations under investigation include: 1) lowering the fuel smear density in order to accommodate more swelling, 2) annular fuel geometry to eliminate the need for a sodium bond, 3) minor alloy additions to stabilize lanthanide fission products inside the fuel and prevent their transport to the cladding where they can participate in fuel-cladding chemical interaction (FCCI), and 4) coatings/liners on the cladding inner surface to mitigate FCCI and enable higher temperature operation. This paper will present the current state of development of metallic fuels for actinide transmutation in the US. Highlights will include recent results from metallic fuel casting experiments, experiments to identify alloy additions to immobilize lanthanide fission products, and postirradiation examinations of annular metallic fuels at low burnup.

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

    Science.gov (United States)

    Shozugawa, Katsumi; Nogawa, Norio; Matsuo, Motoyuki

    2012-04-01

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

  3. Actinide neutron induced cross section measurements using the oscillation technique in the Minerve reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, B.; Leconte, P.; Gruel, A.; Antony, M.; Di-Salvo, J.; Hudelot, J.P.; Pepino, A.; Lecluze, A. [CEA Cadarache, DEN/CAD/DER/SPRC/LEPh, 13 - Saint-Paul-lez-Durance (France)

    2009-07-01

    CEA is deeply involved research programs concerning nuclear fuel advanced studies (actinides, plutonium), waste management, the scientific and technical support of French PWR reactors and EPR reactor, and innovative systems. In this framework, specific neutron integral experiments have been carried out in the critical ZPR (zero power reactor) facilities of the CEA at Cadarache such as MINERVE, EOLE and MASURCA. This paper deals with MINERVE Pool Reactor experiments. MINERVE is mainly devoted to neutronics studies of different reactor core types. The aim is to improve the knowledge of the integral absorption cross sections of actinides (OSMOSE program), of new absorbers (OCEAN program) and also for fission Products (CBU program) in thermal, epithermal and fast neutron spectra. (authors)

  4. Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

    Energy Technology Data Exchange (ETDEWEB)

    Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

    1995-12-31

    Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

  5. Etude de la production de fragments dans la fission induite par neutrons sur l'uranium 238.

    CERN Document Server

    Casoli, P

    2003-01-01

    Mass and charge distributions of fission fragments from the fission induced on uranium 238 by neutrons from 1 to 150 MeV were studied by experimental data and theoretical calculations. Measurements in prompt gamma and x-ray spectroscopy, completed at the LANSCE laboratory of Los Alamos allowed us to determine the secondary fission fragment production yields. Photovoltaic cells were used as a fission trigger. About one hundred fragments were identified and about thirty excitation functions were extracted. Mass and charge distributions at different incident energies were obtained. The comparison with evaluated data (Wahl systematics) shows that the calculations are consistent with the measurements below 20 MeV but not predictive enough above. A potential energy surface was drawn from microscopic constrained self-consistent HFB calculations. We obtained a fragment mass distribution by solving the dynamic Schrodinger equation on this surface.

  6. Fission-like events in the 12C+169Tm system at low excitation energies

    Science.gov (United States)

    Sood, Arshiya; Singh, Pushpendra P.; Sahoo, Rudra N.; Kumar, Pawan; Yadav, Abhishek; Sharma, Vijay R.; Shuaib, Mohd.; Sharma, Manoj K.; Singh, Devendra P.; Gupta, Unnati; Kumar, R.; Aydin, S.; Singh, B. P.; Wollersheim, H. J.; Prasad, R.

    2017-07-01

    Background: Fission has been found to be a dominating mode of deexcitation in heavy-ion induced reactions at high excitation energies. The phenomenon of heavy-ion induced fission has been extensively investigated with highly fissile actinide nuclei, yet there is a dearth of comprehensive understanding of underlying dynamics, particularly in the below actinide region and at low excitation energies. Purpose: Prime objective of this work is to study different aspects of heavy-ion induced fission ensuing from the evolution of composite system formed via complete and/or incomplete fusion in the 12C+169Tm system at low incident energies, i.e., Elab≈6.4 , 6.9, and 7.4 A MeV, as well as to understand charge and mass distributions of fission fragments. Method: The recoil-catcher activation technique followed by offline γ spectroscopy was used to measure production cross sections of fission-like events. The evaporation residues were identified by their characteristic γ rays and vetted by the decay-curve analysis. Charge and mass distributions of fission-like events were studied to obtain dispersion parameters of fission fragments. Results: In the present work, 26 fission-like events (32 ≤Z ≤49 ) were identified at different excitation energies. The mass distribution of fission fragments is found to be broad and symmetric, manifesting their production via compound nuclear processes. The dispersion parameters of fission fragments obtained from the analysis of mass and isotopic yield distributions are found to be in good accord with the reported values obtained for different fissioning systems. A self-consistent approach was employed to determine the isobaric yield distribution. Conclusions: The present work suggests that fission is one of the competing modes of deexcitation of complete and/or incomplete fusion composites at low excitation energies, i.e., E*≈57 , 63, and 69 MeV, where evaporation of light nuclear particle(s) and/or γ rays are assumed to be the sole

  7. Cumulative fission yields of short-lived isotopes under natural-abundance-boron-carbide-moderated neutron spectrum

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-09

    The availability of gamma spectroscopy data on samples containing mixed fission products at short times after irradiation is limited. Due to this limitation, data interpretation methods for gamma spectra of mixed fission product samples, where the individual fission products have not been chemically isolated from interferences, are not well-developed. The limitation is particularly pronounced for fast pooled neutron spectra because of the lack of available fast reactors in the United States. Samples containing the actinide isotopes 233, 235, 238U, 237Np, and 239Pu individually were subjected to a 2$ pulse in the Washington State University 1 MW TRIGA reactor. To achieve a fission-energy neutron spectrum, the spectrum was tailored using a natural abundance boron carbide capsule to absorb neutrons in the thermal and epithermal region of the spectrum. Our tailored neutron spectrum is unique to the WSU reactor facility, consisting of a soft fission spectrum that contains some measurable flux in the resonance region. This results in a neutron spectrum at greater than 0.1 keV with an average energy of 70 keV, similar to fast reactor spectra and approaching that of 235U fission. Unique fission product gamma spectra were collected from 4 minutes to 1 week after fission using single-crystal high purity germanium detectors. Cumulative fission product yields measured in the current work generally agree with published fast pooled fission product yield values from ENDF/B-VII, though a bias was noted for 239Pu. The present work contributes to the compilation of energy-resolved fission product yield nuclear data for nuclear forensic purposes.

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

    Institute of Scientific and Technical Information of China (English)

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

    2012-01-01

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

  9. MICRO/NANO-STRUCTURAL EXAMINATION AND FISSION PRODUCT IDENTIFICATION IN NEUTRON IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    van Rooyen, I. J.; Lillo, T. M.; Wen, H. M.; Hill, C. M.; Holesinger, T. G.; Wu, Y. Q.; Aguiara, J. A.

    2016-11-01

    Advanced microscopic and microanalysis techniques were developed and applied to study irradiation effects and fission product behavior in selected low-enriched uranium oxide/uranium carbide TRISO-coated particles from fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA. Although no TRISO coating failures were detected during the irradiation, the fraction of Ag-110m retained in individual particles often varied considerably within a single compact and at the capsule level. At the capsule level Ag-110m release fractions ranged from 1.2 to 38% and within a single compact, silver release from individual particles often spanned a range that extended from 100% retention to nearly 100% release. In this paper, selected irradiated particles from Baseline, Variant 1 and Variant 3 type fueled TRISO coated particles were examined using Scanning Electron Microscopy, Atom Probe Tomography; Electron Energy Loss Spectroscopy; Precession Electron Diffraction, Transmission Electron Microscopy, Scanning Transmission Electron Microscopy (STEM), High Resolution Electron Microscopy (HRTEM) examinations and Electron Probe Micro-Analyzer. Particle selection in this study allowed for comparison of the fission product distribution with Ag retention, fuel type and irradiation level. Nano sized Ag-containing features were predominantly identified in SiC grain boundaries and/or triple points in contrast with only two sitings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). STEM and HRTEM analysis showed evidence of Ag and Pd co-existence in some cases and it was found that fission product precipitates can consist of multiple or single phases. STEM analysis also showed differences in precipitate compositions between Baseline and Variant 3 fuels. A higher density of fission product precipitate clusters were identified in the SiC layer in particles from the Variant 3 compact compared with the Variant 1 compact. Trend analysis shows

  10. Study of Rapid Radiochemical Separation of 90Rb from Fission Product

    Institute of Scientific and Technical Information of China (English)

    YUE; Yuan-zhen; MAO; Guo-shu; YANG; Lei; DING; You-qian; MA; Peng

    2015-01-01

    90Rb is a high yield and short-lived fission product,and it is a sensitive monitor at low burnup.The decay data of 90Rb are important,however the uncertainty of the decay data of 90Rb tends to be large,such as the half-life data(58±5)s(with an uncertainty of 3.2%)from ENSDF.In order to get the half-life measured more accurately,a carrier-free 90Rb sample of

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

    Energy Technology Data Exchange (ETDEWEB)

    Eibach, M., E-mail: martin.eibach@uni-mainz.d [Institut fuer Kernchemie, Johannes Gutenberg-Universitaet Mainz, Fritz-Strassmann-Weg 2, 55128 Mainz (Germany); Physikalisches Institut, Ruprecht-Karls-Universitaet Heidelberg, Philosophenweg 12, 69120 Heidelberg (Germany); Beyer, T.; Blaum, K. [Physikalisches Institut, Ruprecht-Karls-Universitaet Heidelberg, Philosophenweg 12, 69120 Heidelberg (Germany); Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, 69117 Heidelberg (Germany); Block, M. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Planckstrasse 1, 64291 Darmstadt (Germany); Eberhardt, K. [Institut fuer Kernchemie, Johannes Gutenberg-Universitaet Mainz, Fritz-Strassmann-Weg 2, 55128 Mainz (Germany); Herfurth, F. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Planckstrasse 1, 64291 Darmstadt (Germany); Geppert, C. [Institut fuer Kernchemie, Johannes Gutenberg-Universitaet Mainz, Fritz-Strassmann-Weg 2, 55128 Mainz (Germany); GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Planckstrasse 1, 64291 Darmstadt (Germany); Ketelaer, J. [Institut fuer Physik, Johannes Gutenberg-Universitaet Mainz, Staudinger Weg 7, 55128 Mainz (Germany); Ketter, J. [Physikalisches Institut, Ruprecht-Karls-Universitaet Heidelberg, Philosophenweg 12, 69120 Heidelberg (Germany); Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, 69117 Heidelberg (Germany); Kraemer, J.; Krieger, A. [Institut fuer Kernchemie, Johannes Gutenberg-Universitaet Mainz, Fritz-Strassmann-Weg 2, 55128 Mainz (Germany); Knuth, K. [Institut fuer Physik, Johannes Gutenberg-Universitaet Mainz, Staudinger Weg 7, 55128 Mainz (Germany); Nagy, Sz. [Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, 69117 Heidelberg (Germany); GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Planckstrasse 1, 64291 Darmstadt (Germany)

    2010-02-01

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

  12. Measurement of the isomeric yield ratios of fission products with JYFLTRAP

    CERN Document Server

    Gorelov, D; Hakala, J; Jokinen, A; Kolhinen, V S; Koponen, J; Lantz, M; Matteram, A; Moore, I; Penttilä, H; Pohjalainen, I; Pomp, S; Rakopoulos, V; Reponen, M; Rinta-Antilav, S; Schonnenschein, V; Simutkin, V; Solders, A; Voss, A; Äystö, J

    2014-01-01

    Several isomeric yield ratios of fission products in 25 MeV pr oton-induced fis- sion of 238 U were measured recently at the JYFLTRAP facility. The ion-g uide separator on-line method was utilized to produce radioacti ve ions. The dou- ble Penning-trap mass spectrometer was used to separate iso meric and ground states by their masses. To verify the new experimental techn ique γ -spectro- scopy method was used to obtain the same isomeric ratios.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-05-01

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

  16. Decay modes in spontaneous fission

    CERN Document Server

    Gönnenwein, F

    1999-01-01

    Spontaneous fission (SF) is considered to be the choice reaction for studying the influence of shell and pairing effects in fission in general, and in particular their impact on the mass and energy distributions of fission fragments. For the time being some 35 SF reactions have been analysed in detail for elements ranging from Pu up to Rf. Going from the lighter to the heavier actinides both, the distributions of fragment mass (or charge) and of total kinetic energy undergo dramatic changes. It is observed in experiment, however, that these distributions may be well described as a superposition of a few fission modes, each with its own characteristic mass an energy pattern. The experimental modes are traced in theory to fine structures in the potential energy surface of a fissioning nucleus, provided shell and pairing corrections to the basic liquid drop model are accounted for.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-02-03

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

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

    Science.gov (United States)

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

    1985-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M.G.; Aitken, E.A. (General Electric Co., Sunnyvale, CA (USA). Advanced Nuclear Technology Operation)

    1985-02-01

    A combination of fuel chemistry modelling and equilibrium thermodynamic calculations has been used to predict the atom ratios of Cs and Te fission products (Cs:Te) that find their way into the fuel-cladding interface region of irradiated stainless steel-clad mixed-oxide fast breeder reactor fuel pins. It has been concluded that the ratio of condensed, chemically-associated Cs and Te in the interface region, fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), the observed out-of-pile Cs:Te thresholds for FCCI (proportional4:1) and FPLME (proportional2:1) have been rationalized in terms of Cs:Te thermochemistry and phase equilibria. Also described in the paper is an updated chemical evolution model for reactive/volatile fission product behavior in irradiated oxide pins.

  20. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  1. Fission product data for thermal reactors. Part 2. Users manual for EPRI-CINDER code and data

    Energy Technology Data Exchange (ETDEWEB)

    England, T.R.; Wilson, W.B.; Stamatelatos, M.G.

    1976-12-01

    The objective of this project has been the production of a data library suitable for calculating the buildup of fission product nuclides during the operation of a thermal power reactor. This has been accomplished by reducing the fission product data from the fourth version of the national reference nuclear data base--ENDF/B into a series of linearized decay chains and calculating the effective yields and cross sections of the relevant nuclides. Two versions of the fission product library have been prepared: an 84 chain master library and a reduced 12 chain library, both of which can be used as input for the computer program CINDER. A users manual for an upgraded version of the burnup program CINDER (renamed EPRI-CINDER) is presented.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, F

    2005-06-15

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

  3. Synthesis and Evaluation of new Polyfunctional Molecules for Group Actinide Extraction; Synthese et evaluation de Nouvelles Molecules Polyfonctionnelles pour la Separation Groupee des Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Marie, C.

    2009-10-15

    The aim of this project is to design new extracting molecules for spent nuclear fuel reprocessing. In order to minimize the long-term residual radiotoxicity of the waste, the GANEX process is an option based on homogeneous recycling of actinides. All actinides (U, Np, Pu, Am, Cm), present in a highly acidic aqueous solution, would be extracted together and separated from fission products (especially from lanthanides) using liquid-liquid extraction. In this context, twenty new bi-topic ligands constituted of a nitrogen poly-aromatic unit functionalized by amide groups were synthesized. Liquid-liquid extraction tests with these ligands dissolved alone in the organic phase show that N, N, N', N'-tetra-alkyl-6, 6''(2, 2':6', 2''-terpyridine)-diamides are able to selectively extract actinides at different oxidation states (Np(V et VI), U(VI), Pu(IV), Am(III), Cm(III)) from an aqueous solution 3M HNO{sub 3}. Nevertheless, actinides(III) are poorly extracted. According to crystallographic structures of complexes with Nd(III) and U(VI) determined by X-rays diffraction, these ligands are penta-dentate. In solution (methanol), complexes stoichiometries (1:1) of Nd(III), U(VI) and Pu(IV) were determined by electro-spray ionization mass spectrometry. Stability constants, evaluated by UV-visible spectrophotometry in MeOH/H{sub 2}O solutions, confirm the selectivity of ligands toward actinides(III) with respect to lanthanides(III). Associate to nuclear magnetic resonance experiments and DFT calculations (Density Functional Theory), a better knowledge of their coordination mode was achieved. (author)

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

    Science.gov (United States)

    Sternglass, E J; Gould, J M

    1993-01-01

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

  5. Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel

    Science.gov (United States)

    Lillo, T. M.; van Rooyen, I. J.

    2016-05-01

    In this study, the fission product precipitates at silicon carbide grain boundaries from an irradiated TRISO particle were identified and correlated with the associated grain boundary characteristics. Precession electron diffraction in the transmission electron microscope provided the crystallographic information needed to identify grain boundary misorientation and boundary type (i.e., low angle, random high angle or coincident site lattice (CSL)-related). The silicon carbide layer was found to be composed mainly of twin boundaries and small fractions of random high angle and low angle grain boundaries. Most fission products were found at random, high-angle grain boundaries, with small fractions at low-angle and CSL-related grain boundaries. Palladium (Pd) was found at all types of grain boundaries while Pd-uranium and Pd-silver precipitates were only associated with CSL-related and random, high-angle grain boundaries. Precipitates containing only Ag were found only at random, high-angle grain boundaries, but not at low angle or CSL-related grain boundaries.

  6. Role of ( n,2 n) reactions in transmutation of long-lived fission products

    Science.gov (United States)

    Apse, V. A.; Kulikov, G. G.; Kulikov, E. G.

    2016-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hovgaard, J.; Korsbech, U.

    1992-12-01

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

  8. Fission Product Appearance Rate Coefficients in Design Basis Source Term Determinations - Past and Present

    Science.gov (United States)

    Perez, Pedro B.; Hamawi, John N.

    2017-09-01

    Nuclear power plant radiation protection design features are based on radionuclide source terms derived from conservative assumptions that envelope expected operating experience. Two parameters that significantly affect the radionuclide concentrations in the source term are failed fuel fraction and effective fission product appearance rate coefficients. Failed fuel fraction may be a regulatory based assumption such as in the U.S. Appearance rate coefficients are not specified in regulatory requirements, but have been referenced to experimental data that is over 50 years old. No doubt the source terms are conservative as demonstrated by operating experience that has included failed fuel, but it may be too conservative leading to over-designed shielding for normal operations as an example. Design basis source term methodologies for normal operations had not advanced until EPRI published in 2015 an updated ANSI/ANS 18.1 source term basis document. Our paper revisits the fission product appearance rate coefficients as applied in the derivation source terms following the original U.S. NRC NUREG-0017 methodology. New coefficients have been calculated based on recent EPRI results which demonstrate the conservatism in nuclear power plant shielding design.

  9. Role of (n,2n) reactions in transmutation of long-lived fission products

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-04-01

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

  11. Fission product inventory for CANFLEX-NU and CANFLEX-RU in the limiting channel of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2007-09-15

    The fission product inventory was evaluated for the 37-element fuel, CANFLEX-NU and CANFLEX-RU fuels which irradiated in the assumed limiting O6{sub m}od channel under a normal operating condition. Generally, the assessment of fission product release is divided into 3 steps which are a calculation of FGR during steady state, a calculation of a prompt release at the time of an accident, and a calculation of delayed release after the accident. Among them the first step was described in this report. The analysis was performed by using the ELESTRES code which is a computer code used in a performance evaluation of the fuel bundle during the normal irradiation. The calculation results of the fission product inventory for the 37-element fuel, CANFLEX-NU and CANFLEX-RU were compared to each other.

  12. Sensitivity analysis for CORSOR models simulating fission product release in LOFT-LP-FP-2 severe accident experiment

    Energy Technology Data Exchange (ETDEWEB)

    Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club; Pourgol-Mohammad, Mohammad [Sahand Univ. of Technology, Tabriz (Iran, Islamic Republic of). Dept. of Mechanical Engineering; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2017-03-15

    This paper deals with simulation, sensitivity and uncertainty analysis of LP-FP-2 experiment of LOFT test facility. The test facility simulates the major components and system response of a pressurized water reactor during a LOCA. MELCOR code is used for predicting the fission product release from the core fuel elements in LOFT LP-FP-2 experiment. Moreover, sensitivity and uncertainty analysis is performed for different CORSOR models simulating release of fission products in severe accident calculations for nuclear power plants. The calculated values for the fission product release are compared under different modeling options to the experimental data available from the experiment. In conclusion, the performance of 8 CORSOR modeling options is assessed for available modeling alternatives in the code structure.

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

    Directory of Open Access Journals (Sweden)

    Rudolf Muenze

    2013-01-01

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

  14. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John (Jack) K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  15. Electron microscopic evaluation and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment: A preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    I J van Rooyen; D E Janney; B D Miller; J L Riesterer; P A Demkowicz

    2012-10-01

    ABSTRACT Post-irradiation examination of coated particle fuel from the AGR-1 experiment is in progress at Idaho National Laboratory and Oak Ridge National Laboratory. In this presentation a brief summary of results from characterization of microstructures in the coating layers of selected irradiated fuel particles with burnup of 11.3% and 19.3% FIMA will be given. The main objective of the characterization were to study irradiation effects, fuel kernel porosity, layer debonding, layer degradation or corrosion, fission-product precipitation, grain sizes, and transport of fission products from the kernels across the TRISO layers. Characterization techniques such as scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, and wavelength dispersive spectroscopy were used. A new approach to microscopic quantification of fission-product precipitates is also briefly demonstrated. The characterization emphasized fission-product precipitates in the SiC-IPyC interface, SiC layer and the fuel-buffer interlayer, and provided significant new insights into mechanisms of fission-product transport. Although Pd-rich precipitates were identified at the SiC-IPyC interlayer, no significant SiC-layer thinning was observed for the particles investigated. Characterization of these precipitates highlighted the difficulty of measuring low concentration Ag in precipitates with significantly higher concentrations of contain Pd and U. Different approaches to resolving this problem are discussed. Possible microstructural differences between particles with high and low releases of Ag particles are also briefly discussed, and an initial hypothesis is provided to explain fission-product precipitate compositions and locations. No SiC phase transformations or debonding of the SiC-IPyC interlayer as a result of irradiation were observed. Lessons learned from the post-irradiation examination are described and future actions are recommended.

  16. Research on the chemistry of actinides and fission products for the development of nuclear fuel cycle; Developpement du cycle du combustible nucleaire. Recherche sur la chimie des actinides et des produits de fission

    Energy Technology Data Exchange (ETDEWEB)

    Peretroukhine, V.F. [Institute of Physical Chemistry Russian Academy of Sciences (IPC-Moscow) (Russian Federation); Lecomte, M. [CEA Valrho, Dept. Radiochimie et Procedes, 30 - Marcoule (France); David, F. [Institut de Physique Nucleaire, (IN2P3/CNRS) 91 - Orsay (France)

    2004-07-01

    The present document is prepared for CEA, IPNO and Russian Academy of Sciences administrations in compliance with the decision, accepted by the Director Committee of the three-parties-Convention on its seventh meeting in Moscow 21 May 2002. Being the general frame, determining the strategy of three-party cooperation, the Convention 1993-2003 consists of the number of contracts between organisation-participants. The present document is compiled, using annual proceedings of Director Committee and Technical Committee meetings and using published French-Russian papers and technical reports, produced in the frame of all contracts of the convention. Lists of published papers and personal summary reports of Russian participants of the contracts are presented in Appendix 1. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-01

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

  18. Fission Product Release and Survivability of UN-Kernel LWR TRISO Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M [ORNL; Ferber, Mattison K [ORNL; Lin, Hua-Tay [ORNL

    2014-01-01

    A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from range calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 m diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated with a TRISO particle as a function of fluence. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by measuring the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers as a function of fluence. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

  19. Towards high-precision mass measurements of neutron-rich fission products at TRIGA-SPEC

    Energy Technology Data Exchange (ETDEWEB)

    Nagy, Szilard [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany)

    2010-07-01

    TRIGA-TRAP, the only Penning trap mass spectrometer worldwide at a nuclear research reactor, is installed at TRIGA Mainz as part of the TRIGA-SPEC experiment. The scientific goal is to perform high-precision mass measurements on lanthanoids, actinoids and neutron-rich fission products produced by thermal neutron induced fission of a target inside the reactor. High-precision mass data are scarce in this region of the nuclear chart, and further experimental data are needed for nuclear structure studies of heavy elements, to test the predictive power of nuclear mass models, or as input to nucleosynthesis calculations of the astrophysical r-process. Ions of certain lanthanoids and most actinoids as well as carbon clusters for calibration purposes can be routinely produced by a newly developed non-resonant laser ablation ion source, allowing off-line mass measurements. Besides fundamental research, TRIGA-TRAP serves as a test bench for the development of efficient ion detection techniques, which will enable mass measurements ultimately on a single ion with a half-life of the order of one second. To this end, a unique combination of the commonly used time-of-flight technique and the non-destructive image current detection method is realized in an on-line mass spectrometer. The first mass measurement results are reported.

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

    Energy Technology Data Exchange (ETDEWEB)

    Merriman, L.D.

    1984-04-01

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

  1. IMPACT OF THE CHEMICAL FORM OF IN-CONTAINMENT SOURCE ON FISSION PRODUCT RELEASE FROM WWER-1000/V-320 TYPE NPP CONTAINMENT DURING LOCA

    Directory of Open Access Journals (Sweden)

    Adam Kecek

    2016-12-01

    Full Text Available Nuclear power plant accidents may be followed by a release of fission products into the environment. This release is dependent on several phenomena, such as chemistry, pressure, type of the accident etc. The aim of this paper is to assess the impact of the chemical form of iodine on the fission product release into the environment.

  2. A Study on Fission Product Behavior during a Severe Accident at APR1400 Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Song-Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, calculations have been carried out for a SBO sequence similar to the selected scenario, but a faster one with simple assumptions. Instead, a sensitivity study was carried out to take into account the effects of such differences on the fission product behavior. Probabilistic Safety Assessment (PSA) for Shin-Kori 3·4 nuclear power plants, which are APR1400 type reactors, were reviewed. After all, the representing scenarios were determined to be the sequences with station blackout (SBO), interfacing system LOCA (ISLOCA), and steam generator tube rupture (SGTR), which are similar to those of the U.S.NRC's State-of-the-Art Reactor Consequence Analyses (SOARCA) study. Among those sequences, SBO occupies the largest portion of the risk from severe accidents, and was selected to be analyzed at first about the fission product behavior in the containment. It includes events such as failure of the alternative AC power generator following a blackout event, successful operation of turbine-driven auxiliary feed water (AFW) pump, late recovery of offsite power before containment failure, in-vessel injection and successful actuation of cavity flooding system and spray system, and failure of hydrogen mitigation system. We use MELCOR 1.8.6 with the 35- and 2-cell compartment models of the containment. Since MELCOR does not treat organic iodide, we tried to make the results up by MELCOR-RAIM which is the MELCOR code coupled with RAIM, a stand-alone code developed for evaluation of the iodine behavior. In order to investigate the fission product behavior during a severe accident at APR1400, we have selected the representing scenarios with SBO, ISLOCA and SGTR. Among them, a SBO sequence similar to the selected scenario, but a faster one with simple assumptions, was analyzed using MELCOR v1.8.6 with 35-cell models of the containment. For the sensitivity analysis, we use the 2-cell containment model and the codes with the iodine chemistry model such as MELCOR with

  3. Some aspects of fission and quasifission processes

    Indian Academy of Sciences (India)

    B B Back

    2015-08-01

    The discovery of nuclear fission in 1938–1939 had a profound influence on the field of nuclear physics and it brought this branch of physics into the forefront as it was recognized for having the potential for its seminal influence on modern society. Although many of the basic features of actinide fission were described in a ground-breaking paper by Bohr and Wheeler only six months after the discovery, the fission process is very complex and it has been a challenge for both experimentalists and theorists to achieve a complete and satisfactory understanding of this phenomenon. Many aspects of nuclear physics are involved in fission and it continues to be a subject of intense study even three quarters of a century after its discovery. In this talk, I will review an incomplete subset of the major milestones in fission research, and briefly discuss some of the topics that I have been involved in during my career. These include studies of vibrational resonances and fission isomers that are caused by the second minimum in the fission barrier in actinide nuclei, studies of heavy-ion-induced fission in terms of the angular distributions and the mass–angle correlations of fission fragments. Some of these studies provided evidence for the importance of the quasifission process and the attendant suppression of the complete fusion process. Finally, some of the circumstances around the establishment of large-scale nuclear research in India will be discussed.

  4. Delayed fission

    Energy Technology Data Exchange (ETDEWEB)

    Hatsukawa, Yuichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-07-01

    Delayed fission is a nuclear decay process that couples {beta} decay and fission. In the delayed fission process, a parent nucleus undergoes {beta} decay and thereby populates excited states in the daughter. If these states are of energies comparable to or greater than the fission barrier of the daughter, then fission may compete with other decay modes of the excited states in the daughter. In this paper, mechanism and some experiments of the delayed fission will be discussed. (author)

  5. Actinides transmutation - a comparison of results for PWR benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  6. Utilization of Minor Actinides (Np, Am, Cm) in Nuclear Power Reactor

    Science.gov (United States)

    Gerasimov, A.; Bergelson, B.; Tikhomirov, G.

    2014-06-01

    Calculation research of the utilization process of minor actinides (transmutation with use of power released) is performed for specialized power reactor of the VVER type operating on the level of electric power of 1000 MW. Five subsequent cycles are considered for the reactor with fuel elements containing minor actinides along with enriched uranium. It was shown that one specialized reactor for the one cycle (900 days) can utilize minor actinides from several VVER-1000 reactors without any technological and structural modifications. Power released because of minor actinide fission is about 4% with respect to the total power

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-01

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

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

    Science.gov (United States)

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

    2014-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

  13. Research in actinide chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Choppin, G.R.

    1993-01-01

    This research studies the behavior of the actinide elements in aqueous solution. The high radioactivity of the transuranium actinides limits the concentrations which can be studied and, consequently, limits the experimental techniques. However, oxidation state analogs (trivalent lanthanides, tetravalent thorium, and hexavalent uranium) do not suffer from these limitations. Behavior of actinides in the environment are a major USDOE concern, whether in connection with long-term releases from a repository, releases from stored defense wastes or accidental releases in reprocessing, etc. Principal goal of our research was expand the thermodynamic data base on complexation of actinides by natural ligands (e.g., OH[sup [minus

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

    Energy Technology Data Exchange (ETDEWEB)

    Lau, Ch

    2000-09-15

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

  15. Multigroup and few-group cross sections for ENDF/B-IV fission products; the TOAFEW collapsing code and data file of 154-group fission-product cross sections. [In FORTRAN

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-02-01

    The ENDF/B-IV fission-product data file includes data describing 824 nuclides. Cross sections, given for 181 of these nuclides, were processed into 154 neutron energy groups. The production of the data file is described. The TOAFEW code, useful in collapsing the multigroup values to few-group cross sections, is presented with instructions and examples of its use. The file of multigroup cross sections is available on request. 3 figures, 11 tables.

  16. Partitioning of Minor Actinides from High Active Raffinates using Bis-Diglycol-amides (BisDGA) as new efficient Extractants

    Energy Technology Data Exchange (ETDEWEB)

    Modolo, G.; Vijgen, H. [Forschungszentrum Juelich GmbH, Institute for Energy Research, Safety Research and Reactor Technology, 52425 Juelich (Germany); Espartero, A.G. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Avda. Complutense 22, 28040-Madrid (Spain); Prados, P. [Departamento de Quimica Organica, Facultad de Ciencias, Universidad Autonoma de Madrid - UAM, carretera de Colmenar Viejo km 15.3, 28049-Madrid (Spain); Mendoza, J. de [Departamento de Quimica Organica, Facultad de Ciencias, Universidad Autonoma de Madrid - UAM, carretera de Colmenar Viejo km 15.3, 28049-Madrid (Spain); Institut Catala d' Investigacio Quimica (ICIQ) Av. Paisos Catalans 16, 43007-Tarragona (Spain)

    2008-07-01

    Two new polyamide extractants has been selected, namely UAM-069 and UAM-081, both synthesized at the University of Madrid (UAM), to develop a new separation process. These two ligands are bis-diglycol-amides, consisting of two diglycol-amides moieties grafted on an aromatic platform (UAM-069) or on an aliphatic linker (UAM-081), respectively. The extraction of actinides and fission products was studied from synthetic PUREX raffinate. Actinides(III) and lanthanides(III) are highly extracted from acidities > 1 mol/L HNO{sub 3}. The extraction of Zr, Mo and Pd could be suppressed with complexing agents such as oxalic acid and HEDTA. In the present paper the results of the batch extraction results are presented which serve for the development of a new continuous counter current process to be tested in centrifugal contactors. (authors)

  17. Thin extractive membrane for monitoring actinides in aqueous streams.

    Science.gov (United States)

    Chavan, Vivek; Paul, Sumana; Pandey, Ashok K; Kalsi, P C; Goswami, A

    2013-09-15

    Alpha spectrometry and solid state nuclear track detectors (SSNTDs) are used for monitoring ultra-trace amount of alpha emitting actinides in different aqueous streams. However, these techniques have limitations i.e. alpha spectrometry requires a preconcentration step and SSNTDs are not chemically selective. Therefore, a thin polymer inclusion membrane (PIM) supported on silanized glass was developed for preconcentraion and determination of ultra-trace concentration of actinides by α-spectrometry and SSNTDs. PIMs were formed by spin coating on hydrophobic glass slide or solvent casting to form thin and self-supported membranes, respectively. Sorption experiments indicated that uptakes of actinides in the PIM were highly dependent on acidity of solution i.e. Am(III) sorbed up to 0.1 molL(-1) HNO₃, U(VI) up to 0.5 molL(-1) HNO₃ and Pu(IV) from HNO₃ concentration as high as 4 molL(-1). A scheme was developed for selective sorption of target actinide in the PIM by adjusting acidity and oxidation state of actinide. The actinides sorbed in PIMs were quantified by alpha spectrometry and SSNTDs. For SSNTDs, neutron induced fission-fragment tracks and α-particle tracks were registered in Garware polyester and CR-39 for quantifications of natural uranium and α-emitting actinides ((241)Am/(239)Pu/(233)U), respectively. Finally, the membranes were tested to quantify Pu in 4 molL(-1) HNO3 solutions and synthetic urine samples.

  18. Superasymmetric fission at intermediate energy and production of neutron-rich nuclei with A

    NARCIS (Netherlands)

    Huhta, M; Dendooven, P; Honkanen, A; Jokinen, A; Lhersonneau, G; Oinonen, M; Penttila, H; Perajarvi, K; Rubchenya, VA; Aysto, J

    1997-01-01

    The yields of neutron-rich Ni, Cu, Zn, Ga and Ge-isotopes were measured in 25 MeV proton induced fission of U-238 using the ion guide-based isotope separator technique. The results indicate enhancement for superasymmetric mass division at intermediate excitation energy of the fissioning nucleus and

  19. Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

    Science.gov (United States)

    Chang, G S; Ambrosek, R G

    2005-01-01

    The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

  20. Neutron-energy-dependent defect production cross sections for fission and fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Doiron, D.R.

    1976-06-01

    Neutron cross sections for displacements and post-short-term cascade annealing defects are derived from nuclear kinematics calculations of primary atomic recoil energy distributions and the number of secondary defects produced per primary as a function of recoil energy. For the first time, recoil kinematics of charged- and multiple-particle emission reactions are treated rigorously using a compound-nucleus evaporation spectrum nuclear model. Secondary-defect production functions, derived from computer simulation experiments, are taken from the literature. Spectral-averaged defect production cross sections for a fusion reactor first-wall-type environment are on the order of 1.5 to 2.5 times those for a fast fission reactor core-type spectrum. The indicated range of uncertainty is primarily due to secondary-defect production model sensitivity. Nuclear model and data errors are expected to become more significant at high neutron energies, greater than approximately 20 MeV. Fusion reactor environments are found to produce some very energetic recoils and high-energy release events due to charged-particle reactions such as (n,..cap alpha..).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-11-01

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

  2. Actinide Isotopes for the Synthesis of Superheavy Nuclei

    Science.gov (United States)

    Roberto, J. B.; Alexander, C. W.; Boll, R. A.; Dean, D. J.; Ezold, J. G.; Felker, L. K.; Rykaczewski, K. P.

    2014-09-01

    Recent research resulting in the synthesis of isotopes of new elements 113-118 has demonstrated the importance of actinide targets in superheavy element research. Oak Ridge National Laboratory (ORNL) has unique facilities for the production and processing of actinide target materials, including the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC). These facilities have provided actinide target materials that have been used for the synthesis of all superheavy (SHE) elements above Copernicium (element 112). In this paper, the use of actinide targets for SHE research and discovery is described, including recent results for element 117 using 249Bk target material from ORNL. ORNL actinide capabilities are reviewed, including production and separation/purification, availabilities of actinide materials, and future opportunities including novel target materials such as 251Cf.

  3. Hidden systematics of fission channels

    Directory of Open Access Journals (Sweden)

    Schmidt Karl-Heinz

    2013-12-01

    Full Text Available It is a common procedure to describe the fission-fragment mass distributions of fissioning systems in the actinide region by a sum of at least 5 Gaussian curves, one for the symmetric component and a few additional ones, together with their complementary parts, for the asymmetric components. These components have been attributed to the influence of fragment shells, e.g. in the statistical scission-point model of Wilkins, Steinberg and Chasman. They have also been associated with valleys in the potential-energy landscape between the outer saddle and the scission configuration in the multi-channel fission model of Brosa. When the relative yields, the widths and the mean mass-asymmetry values of these components are fitted to experimental data, the mass distributions can be very well reproduced. Moreover, these fission channels are characterised by specific values of charge polarisation, total kinetic energy and prompt-neutron yields. The present contribution investigates the systematic variation of the characteristic fission-channel properties as a function of the composition and the excitation energy of the fissioning system. The mean position of the asymmetric fission channels in the heavy fragment is almost constant in atomic number. The deformation of the nascent fragments at scission, which is the main source of excitation energy of the separated fission fragments ending up in prompt-neutron emission, is found to be a unique function of Z for the light and the heavy fragment of the asymmetric fission channels. A variation of the initial excitation energy of the fissioning system above the fission saddle is only seen in the neutron yield of the heavy fragment. The charge polarisation in the two most important asymmetric fission channels is found to be constant and to appreciably exceed the macroscopic value. The variation of the relative yields and of the positions of the fission channels as a function of the composition and excitation energy

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

    Science.gov (United States)

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

    2016-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-09-01

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

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

    CERN Document Server

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

    1999-01-01

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

  7. Generalized Energy-Dependent Q Values for Fission

    Energy Technology Data Exchange (ETDEWEB)

    Vogt, R

    2010-03-31

    We extend Madland's parameterization of the energy release in fission to obtain the dependence of the fission Q value for major and minor actinides on the incident neutron energies in the range 0 {le} E{sub n} {le} 20 MeV. Our parameterization is based on the actinide evaluations recommended for the ENDF/B-VII.1 release. This paper describes the calculation of energydependent fission Q values based on the calculation of the prompt energy release in fission by Madland. This calculation was adopted for use in the LLNL ENDL database and then generalized to obtain the prompt fission energy release for all actinides. Here the calculation is further generalized to the total energy release in fission. There are several stages in a fission event, depending on the time scale. Neutrons and gammas may be emitted at any time during the fission event.While our discussion here is focussed on compound nucleus creation by an incident neutron, similar parameterizations could be obtained for incident gammas or spontaneous fission.

  8. Study and development of a method allowing the identification of actinides inside nuclear waste packages, by active neutron or photon interrogation and delayed gamma-ray spectrometry; Etude et developpement d'une technique de dosage des actinides dans les colis de dechets radioactifs par interrogation photonique ou neutronique active et spectrometrie des gamma retardes

    Energy Technology Data Exchange (ETDEWEB)

    Carrel, F

    2007-10-15

    An accurate estimation of the alpha-activity of a nuclear waste package is necessary to select the best mode of storage. The main purpose of this work is to develop a non-destructive active method, based on the fission process and allowing the identification of actinides ({sup 235}U, {sup 238}U, {sup 239}Pu). These three elements are the main alpha emitters contained inside a package. Our technique is based on the detection of delayed gammas emitted by fission products. These latter are created by irradiation with the help of a neutron or photon beam. Performances of this method have been investigated after an Active Photon or Neutron Interrogation (INA or IPA). Three main objectives were fixed in the framework of this thesis. First, we measured many yields of photofission products to compensate the lack of data in the literature. Then, we studied experimental performances of this method to identify a given actinide ({sup 239}Pu in fission, {sup 235}U in photofission) present in an irradiated mixture. Finally, we assessed the application of this technique on different mock-up packages for both types of interrogation (118 l mock-up package containing EVA in fission, 220 l mock-up package with a wall of concrete in photofission). (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Wright, A.L. [Oak Ridge National Lab., TN (United States)

    1994-06-01

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

  10. Determination of Fission Product Yields of 235U, 238U and 239Pu for Neutron Energies from 0.5 to 14.8 MeV

    Science.gov (United States)

    Gooden, Matthew; Arnold, Charles; Becker, John; Bhatia, Chitra; Bhike, Megha; Fowler, Malcolm; Howell, Calvin; Kelley, John; Stoyer, Mark; Tonchev, Anton; Tornow, Werner; Vieira, Dave; Wilhelmy, Jerry

    2014-03-01

    A joint TUNL-LANL-LLNL collaboration has been formed to study the issue of possible energy dependences for certain fission product isotopes. Work has been carried out at the TUNL 10 MV Tandem accelerator which produces nearly mono-energetic neutrons via either 2H(d,n)3He,3H(d,n)4He,or3H(p,n)3He reactions. Three dual fission ionization chambers dedicated to 235U, 238U and 239Pu thick target foils and thin monitor foils respectively, were exposed to the neutron beams. After irradiation, thick target foils were gamma counted over a period of 1-2 months and characteristic gamma rays from fission products were recorded using HPGe detectors at TUNL's low background counting area. Using the dual fission chambers, relative fission product yield were determined at a high precision of 2-3 % as well as absolute fission product yields at a lower precision of 5-6 %. Preliminary results will be presented for a number of fission product isotopes over the incident neutron energy range of 0.5 to 14.8 MeV.

  11. Excitation-energy dependence of the nuclear fission characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Baba, H.; Saito, T.; Takahashi, N. [Osaka City Univ. (Japan). Faculty of Science] [and others

    1996-03-01

    It is known that the width parameter of the fragment mass yield distribution follows a beautiful systematics with respect to the excitation energy. According to this systematics, the fission characteristics following the systematics should disappear when the excitation energy Ex goes down to 14 MeV. The present purpose is to elucidate if, where, how and why a transition takes place in the fission characteristics of the asymmetric fission of light actinide elements. Two types of experiments are performed, one is the double-energy measurement of the kinetic energies of complementary fragments in the thermal-neutron fission of {sup 235,233}U and proton-induced fission of {sup 238}U at 13.3- and 15.7-MeV excitations, and the other is the radiochemical study of proton-induced fission and photofission of {sup 238}U at various excitation energies. In conclusion, it has demonstrated that there are two distinctive fission mechanisms in the low-energy fission of light actinide elements and the transition between them takes place around 14-MeV excitation. The characteristics of proton fission and photofission in the energy range lower than the above transition point are the essentially the same as those of thermal-neutron fission and also spontaneous fission. The results of GDR fission indicates the fission in the high-energy side starts from the nuclear collective states, whereas the lower-energy fission is of non-collective nature. It is likely that thermal-neutron fission is rather of the barrier-penetrating type like spontaneous fission than the threshold fission. (S.Y.)

  12. Status of nuclear data for actinides

    Energy Technology Data Exchange (ETDEWEB)

    Guzhovskii, B.Y.; Gorelov, V.P.; Grebennikov, A.N. [Russia Federal Nuclear Centre, Arzamas (Russian Federation)] [and others

    1995-10-01

    Nuclear data required for transmutation problem include many actinide nuclei. In present paper the analysis of neutron fission, capture, (n,2n) and (n,3n) reaction cross sections at energy region from thermal point to 14 MeV was carried out for Th, Pa, U, Np, Pu, Am and Cm isotops using modern evaluated nuclear data libraries and handbooks of recommended nuclear data. Comparison of these data indicates on substantial discrepancies in different versions of files, that connect with quality and completeness of original experimental data.

  13. Simulated fission product-SiC interaction in Triso-coated LEU or MEU HTGR fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Pearson, R.L.; Lindemer, T.B.; Beahm, E.C.

    1980-11-01

    Proliferation issues relating to the use of highly enriched uranium (HEU) have led to an evaluation of the fission product-SiC interaction problems that might arise if low enriched uranium (LEU) or medium enriched uranium (MEU) were used as fissile fuel in HTGR systems. Simulated Triso-coated UO/sub 2/, UC/sub 2/, and UO/sub 2//UC/sub 2/ particles mixed with varying amounts of Mo, Ru, Rh, Pd, Ag, and Cd were prepared. These fission products were chosen because, after full burnup, their concentrations are higher in LEU and MEU fuels than in HEU fuel. After the particles were heat treated in the laboratory, their behavior was examined by use of metallography, scanning electron microscopy, and electron microprobe x-ray analysis.

  14. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 3: Fission-Product Transport and Dose PIRTs

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert Noel [ORNL

    2008-03-01

    This Fission Product Transport (FPT) Phenomena Identification and Ranking Technique (PIRT) report briefly reviews the high-temperature gas-cooled reactor (HTGR) FPT mechanisms and then documents the step-by-step PIRT process for FPT. The panel examined three FPT modes of operation: (1) Normal operation which, for the purposes of the FPT PIRT, established the fission product circuit loading and distribution for the accident phase. (2) Anticipated transients which were of less importance to the panel because a break in the pressure circuit boundary is generally necessary for the release of fission products. The transients can change the fission product distribution within the circuit, however, because temperature changes, flow perturbations, and mechanical vibrations or shocks can result in fission product movement. (3) Postulated accidents drew the majority of the panel's time because a breach in the pressure boundary is necessary to release fission products to the confinement. The accidents of interest involved a vessel or pipe break, a safety valve opening with or without sticking, or leak of some kind. Two generic scenarios were selected as postulated accidents: (1) the pressurized loss-of-forced circulation (P-LOFC) accident, and (2) the depressurized loss-of-forced circulation (D-LOFC) accidents. FPT is not an accident driver; it is the result of an accident, and the PIRT was broken down into a two-part task. First, normal operation was seen as the initial starting point for the analysis. Fission products will be released by the fuel and distributed throughout the reactor circuit in some fashion. Second, a primary circuit breach can then lead to their release. It is the magnitude of the release into and out of the confinement that is of interest. Depending on the design of a confinement or containment, the impact of a pressure boundary breach can be minimized if a modest, but not excessively large, fission product attenuation factor can be introduced into

  15. A study on the fission product release behavior during DUPIC pellet fabrication and the characteristics of cesium trapping

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Ho; Shin, Jin Myeong; Park, Geun Il; Nam, Seung Hee; Park, Jang Jin; Lee, Jung Won; Yang, Myung Seung [Korea Atomic Energy Reserach Institute, Daejon (Korea, Republic of)

    2001-11-15

    During the OREOX and sintering process in DUPIC fuel fabrication, fission products would be released from the spent fuel material. Radioactivity measurement using gamma spectrometry technique was performed to identify the radionuclide species released from the spent fuel material at high temperatures up to 1700 deg. C. It was revealed that Cs-137 is the main fission product released from the spent fuel material, and about 99% of Cs-137 is released during sintering process. Therefore, the trapping characteristics of inactive cesium using the fly ash filter in off-gas stream at high temperature was investigated in terms of reaction temperature, carrier gas and gas velocity. The amount of trapped cesium by the fly ash filter was increased with increasing reaction temperature, whereas it was decreased with increasing gas velocity.

  16. History and Actual State of Non-HEU Fission-Based Mo-99 Production with Low-Performance Research Reactors

    Directory of Open Access Journals (Sweden)

    S. Dittrich

    2013-01-01

    Full Text Available Fifty years ago, one of the worldwide first industrial production processes to produce fission-Mo-99 for medical use had been started at ZfK Rossendorf (now: HZDR, Germany. On the occasion of this anniversary, it is worth to mention that this original process (called LITEMOL now together with its target concept used at that time can still be applied. LITEMOL can be adapted very easily to various research reactors and applied at each site, which maybe still of interest for very small-scale producers. Besides this original process, two further and actually proven processes are suitable as well and recommended for small-scale LEU fission Mo-99 production also. They are known under the names KSA/KSS COMPACT and ROMOL LITE and will be described below.

  17. PREFACE: Actinides 2009

    Science.gov (United States)

    Rao, Linfeng; Tobin, James G.; Shuh, David K.

    2010-07-01

    This volume of IOP Conference Series: Materials Science and Engineering consists of 98 papers that were presented at Actinides 2009, the 8th International Conference on Actinide Science held on 12-17 July 2009 in San Francisco, California, USA. This conference was jointly organized by Lawrence Livermore National Laboratory and Lawrence Berkeley National Laboratory. The Actinides conference series started in Baden-Baden, Germany (1975) and this first conference was followed by meetings at Asilomar, CA, USA (1981), Aix-en-Provence, France (1985), Tashkent, USSR (1989), Santa Fe, NM, USA (1993), Baden-Baden, Germany (1997), Hayama, Japan (2001), and Manchester, UK (2005). The Actinides conference series provides a regular venue for the most recent research results on the chemistry, physics, and technology of the actinides and heaviest elements. Actinides 2009 provided a forum spanning a diverse range of scientific topics, including fundamental materials science, chemistry, physics, environmental science, and nuclear fuels. Of particular importance was a focus on the key roles that basic actinide chemistry and physics research play in advancing the worldwide renaissance of nuclear energy. Editors Linfeng Rao Lawrence Berkeley National Laboratory (lrao@lbl.gov) James G Tobin Lawrence Livermore National Laboratory (tobin1@llnl.gov) David K Shuh Lawrence Berkeley National Laboratory (dkshuh@lbl.gov)

  18. Fast-neutron interaction with the fission product {sup 103}Rh

    Energy Technology Data Exchange (ETDEWEB)

    Smith, A.B. [Argonne National Lab., IL (United States)]|[Arizona Univ., Tucson, AZ (United States); Guenther, P.T. [Argonne National Lab., IL (United States)

    1993-09-01

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

  19. Evaporation release behavior of volatile fission products from liquid sodium pool to the inert cover gas

    Energy Technology Data Exchange (ETDEWEB)

    Nakagiri, T.; Miyahara, S. [Oarai Engineering Center, Power Reactor and Nuclear Fuel Development Corp., Oaraimachi, Ibaraki (Japan)

    1996-12-01

    In fuel failure of sodium cooled fast breeder reactors, released volatile fission products (VFPs) such as iodine and cesium from the fuel will be dissolved into the liquid sodium coolant and transferred to the cover vaporization. In the cover gas system of the reactor, natural convection occurs due to temperature differences between the sodium pool and the gas phase. The release rates of VFPs together with sodium vaporization are considered to be controlled by the convection. In this study, three analytical models are developed and examined to calculate the transient release rates using the equilibrium partition coefficients of VFPs. The calculated release rates are compared with experimental results for sodium and sodium iodide. The release rate of sodium is closest to the calculation by the heterogeneous nucleation theory. The release rate of sodium iodide obtained from the experiment is between the release rates calculated by the model based on heat-and-mass transfer analogy and the Hill`s theory. From this study, it is confirmed that the realistic release rate of sodium is able to be calculated by the model based on the heterogeneous nucleation theory. The conservative release rate of sodium iodide is able to be calculated by the model based on the Hill`s theory using the equilibrium partition coefficient of sodium iodide. (author) 7 figs., 1 tab., 3 refs.

  20. Hot beta particles in the lung: Results from dogs exposed to fission product radionuclides

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, F.F.; Griffith, W.C.; Hobbs, C.H. [and others

    1995-12-01

    The Chernobyl nuclear reactor accident resulted in the release of uranium dioxide fuel and fission product radionuclides into the environment with the fallout of respirable, highly radioactive particles that have been termed {open_quotes}hot beta particles.{close_quotes} There is concern that these hot beta particles (containing an average of 150-20,000 Bq/particle), when inhaled and deposited in the lung, may present an extraordinary hazard for the induction of lung cancer. We reviewed data from a group of studies in dogs exposed to different quantities of beta-emitting radionuclides with varied physical half-lives to determine if those that inhaled hot beta particles were at unusual risk for lung cancer. This analysis indicates that the average dose to the lung is adequate to predict biologic effects of lung cancer for inhaled beta-emitting radionuclides in the range of 5-50 Gy to the lung and with particle activities in the range of 0.10-50 Bq/particle.

  1. Separation of the rare-earth fission product poisons from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  2. Separation of the rare-earth fission product poisons from spent nuclear fuel

    Science.gov (United States)

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

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

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, T., E-mail: yamaguti@phy.saitama-u.ac.jp [Department of Physics, Saitama University, Saitama 338-8570 (Japan); Enomoto, A.; Kouno, J.; Yamaki, S.; Matsunaga, S.; Suzaki, F.; Suzuki, T. [Department of Physics, Saitama University, Saitama 338-8570 (Japan); Abe, Y.; Nagae, D.; Okada, S.; Ozawa, A.; Saito, Y.; Sawahata, K. [Institute of Physics, University of Tsukuba, Ibaraki 305-8571 (Japan); Kitagawa, A.; Sato, S. [National Institute of Radiological Sciences, Chiba 263-8555 (Japan)

    2014-12-01

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

  4. Investigation on high temperature vapor pressure of UO 2 containing simulated fission-product elements

    Science.gov (United States)

    Yano, T.; Ohtsubo, A.; Ishii, T.

    1984-06-01

    During the hypothetical core disruptive accident (HCDA) of a fast breeder reactor (FBR), the temperature of the fuel would rise above 3000 K. The experimental data concerning the saturated fuel vapor pressure are necessary for the analysis of the HCDA. In this study, the UO 2 containing Cs, Ba, Ag, or Sn was used to simulate the irradiated fuel in the FBR. The saturated vapor pressure of pure UO 2 and UO 2 containing Cs, Ba, Ag, or Sn at 3000 to 5000 K was measured dynamically with a pulse laser and a torsion pendulum. The surface of a specimen on the pendulum was heated to eject vapor by the injection of a giant pulse ruby laser beam. The pressure of the ejected vapor was measured by both the maximum rotation angle of the pendulum and the duration of vapor ejection. The saturated vapor pressure was theoretically calculated by using the ejected vapor pressure. The surface temperature of the specimen was estimated from the irradiated energy density measured with a laser energy meter. The saturated vapor pressure of UO 2 at 3640 to 5880 K measured in this study was near the extrapolated value of Ackermann's low temperature data. The vapor pressure of UO 2 containing Cs, Ba, Ag or Sn was higher than that of UO 2. The saturated vapor pressure of UO 2 and a solid fission products system was calculated by using these experimental data.

  5. Study on collection efficiency of fission products by spray: Experimental device and modelling

    Energy Technology Data Exchange (ETDEWEB)

    Ducret, D.; Roblot, D.; Vendel, J. [Institut de Protection et de Surete Nucleaire, Gif-Sur-Yvette (France); Billarand, Y. [ECCO Pharmacie et Chimie, Neuilly (France)

    1997-08-01

    Consequences of an hypothetical overheating reactor accident in nuclear power plants can be limited by spraying cold water drops into containment building. The spray reduces the pressure and the temperature levels by condensation of steam and leads to the washout of fission products (aerosols and gaseous iodine). The present study includes a large program devoted to the evaluation of realistic washout rates. An experimental device (named CARAIDAS) was designed and built in order to determine the collection efficiency of aerosols and iodine absorption by drops with representative conditions of post-accident atmosphere. This experimental device is presented in the paper and more particularly: (1) the experimental enclosure in which representative thermodynamic conditions can be achieved, (2) the monosized drops generator, the drops diameter measurement and the drops collector, (3) the cesium iodide aerosols generator and the aerosols measurements. Modelling of steam condensation on drops aerosols collection and iodine absorption are described. First experimental and code results on drops and aerosols behaviour are compared. 8 refs., 18 figs.

  6. Chemistry of actinides; Chimie des actinides

    Energy Technology Data Exchange (ETDEWEB)

    Vitorge, P. [CEA/Saclay, Dept. d' Entreposage et de Stockage des Dechets (DESD), 91 - Gif-sur-Yvette (France)

    1999-07-01

    This article gives the basic data of the actinides chemistry, describes then qualitatively the main parts of the fuel cycle and concludes with quantitative data. The theoretical recalls give qualitative notions to explain the chemical reactivity of actinides and to understand thus the values of the thermodynamic data which allow quantitative anticipations at equilibrium. The Thermodynamic Data Base (TDB) of the NEA-OECD and the CEA in France have recently estimated some of them in using and developing methodologies whose some are presented here. Some current problems of actinides chemistry are described: analysis of the possibilities to (1)improve the reprocessing of long-lived actinides (2)anticipate their behaviour in the environment in order to compare the impact of the different options of the wastes management. The Pourbaix diagrams summarize the chemistry in solution; the author has added information on the solubility, the influence of the ionic strength and of the complexes formation in bicarbonate/carbonate (HCO{sub 3}{sup -}/CO{sub 3}{sup 2-}) media. The discussion on the choice of the equilibrium constants allows to point out the particular points, the dubiousness and the data which have to be proved. (O.M.)

  7. A generalized method for characterization of 235U and 239Pu content using short-lived fission product gamma spectroscopy

    Science.gov (United States)

    Knowles, Justin; Skutnik, Steven; Glasgow, David; Kapsimalis, Roger

    2016-10-01

    Rapid nondestructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the Oak Ridge National Laboratory High Flux Isotope Reactor Neutron Activation Analysis facility has developed a generalized nondestructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and makes use of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a complete characterization of isotopic identification, mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% recovery bias have been conducted on standards of 235U and 239Pu as low as 12 ng in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 198 ng of fissile mass with less than 7% recovery bias. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. It is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation facilities, and account for increasingly complex sample matrices.

  8. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  9. A physical description of fission product behavior fuels for advanced power reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  10. HTR-2014 Paper Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2001-10-01

    The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of fission products silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of fission products from fuel compacts and fuel particles, and retention of fission products in the compacts outside of the SiC layer. PARFUME-predicted fractional release of these fission products was determined and compared to the PIE measurements. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed silicon carbide (SiC) layers, the over-prediction is by a factor of about two, corresponding to an over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of about 100. For intact particles, whose release is much lower, the over-prediction is by an average of about an order of magnitude, which could additionally be attributed to an over-estimated diffusivity in SiC by about 30%. The release of strontium from intact particles is also over-estimated by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Furthermore, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the

  11. Production Cycle for Large Scale Fission Mo-99 Separation by the Processing of Irradiated LEU Uranium Silicide Fuel Element Targets

    Directory of Open Access Journals (Sweden)

    Abdel-Hadi Ali Sameh

    2013-01-01

    Full Text Available Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2 compound. Excellent results are achieved by the oxidizing dissolution of the fuel meat in hydrofluoric acid at room temperature. The resulting solution is directly behind added to an over stoichiometric amount of potassium hydroxide solution. Uranium and the bulk of fission products are precipitated together with the transuranium compounds. The filtrate contains the molybdenum and the soluble fission product species. It is further treated similar to the in-full scale proven process. The generated off gas stream is handled also as experienced before after passing through KOH washing solution. The generated alkaline fluoride containing waste solution is noncorrosive. Nevertheless fluoride can be selectively bonded as in soluble CaF2 by addition of a mixture of solid calcium hydroxide calcium carbonate to the sand cement mixture used for waste solidification. The generated elevated amounts of LEU remnants can be recycled and retargeted. The related technology permits the minimization of the generated fuel waste, saving environment, and improving processing economy.

  12. Managing Zirconium Chemistry and Phase Compatibility in Combined Process Separations for Minor Actinide Partitioning

    Energy Technology Data Exchange (ETDEWEB)

    Wall, Nathalie [Washington State Univ., Pullman, WA (United States); Nash, Ken [Washington State Univ., Pullman, WA (United States); Martin, Leigh [Washington State Univ., Pullman, WA (United States)

    2017-03-17

    In response to the NEUP Program Supporting Fuel Cycle R&D Separations and Waste Forms call DEFOA- 0000799, this report describes the results of an R&D project focusing on streamlining separation processes for advanced fuel cycles. An example of such a process relevant to the U.S. DOE FCR&D program would be one combining the functions of the TRUEX process for partitioning of lanthanides and minor actinides from PUREX(UREX) raffinates with that of the TALSPEAK process for separating transplutonium actinides from fission product lanthanides. A fully-developed PUREX(UREX)/TRUEX/TALSPEAK suite would generate actinides as product(s) for reuse (or transmutation) and fission products as waste. As standalone, consecutive unit-operations, TRUEX and TALSPEAK employ different extractant solutions (solvating (CMPO, octyl(phenyl)-N,Ndiisobutylcarbamoylmethylphosphine oxide) vs. cation exchanging (HDEHP, di-2(ethyl)hexylphosphoric acid) extractants), and distinct aqueous phases (2-4 M HNO3 vs. concentrated pH 3.5 carboxylic acid buffers containing actinide selective chelating agents). The separate processes may also operate with different phase transfer kinetic constraints. Experience teaches (and it has been demonstrated at the lab scale) that, with proper control, multiple process separation systems can operate successfully. However, it is also recognized that considerable economies of scale could be achieved if multiple operations could be merged into a single process based on a combined extractant solvent. The task of accountability of nuclear materials through the process(es) also becomes more robust with fewer steps, providing that the processes can be accurately modeled. Work is underway in the U.S. and Europe on developing several new options for combined processes (TRUSPEAK, ALSEP, SANEX, GANEX, ExAm are examples). There are unique challenges associated with the operation of such processes, some relating to organic phase chemistry, others arising from the

  13. Fission product release and microstructure changes of irradiated MOX fuel at high temperatures

    Science.gov (United States)

    Colle, J.-Y.; Hiernaut, J.-P.; Wiss, T.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    burnups correspond reasonably well with measurement of Walker et al. [11]. All those data are shown Fig. 2.Fragments of 2-8 mg were chosen for the experiments. Since these specimens are small compared to the drilled sample size and were taken randomly, the precise radial position could not be determined, in particular the specimens of sample type, A and B could be from close radial locations.Specimens from each drilled sample type were annealed up to complete vaporisation (˜2600 K) at a speed of about 10 K min-1 in a Knudsen effusion mass spectrometer (KEMS) described previously [13,14]. In addition to helium and to the FGs all the species present in the vapour between 83 and 300 a.m.u. were measured during the heating. Additionally, the 85Kr isotope was analysed in a cold trap by β and γ counting. The long-lived fission gas isotopes correspond to masses 131, 132, 134 and 136 for Xe and 83, 84, 85 and 86 for Kr. The absolute quantities of gas released from specimens of sample types A and B were also determined using the in-house built Q-GAMES (Quantitative gas measurement system), described in detail in [15].For each of the samples, fragments were also annealed and measured in the KEMS up to specific temperatures corresponding to different stages of the FGs or He release. These fragments were subsequently analysed by Scanning Electron Microscopy (SEM, Philips XL40) [16] in order to investigate the relationship between structural changes, burn-up, irradiation temperature and fission products release. SEM observations were also done on the samples before the KEMS experiments and the fracture surface appearance of the samples is shown in Fig. 3, revealing the presence of the high burnup structure (HBS) in the Pu-rich agglomerates.A summary of the 12 samples analysed by KEMS, SEM and Q-GAMES is given in Table 1. At 1300 K no clear change potentially related to gas release appears in the UM and PA. At 1450 K a beginning of grain boundaries opening can be observed as well as

  14. Nuclear Fission

    Science.gov (United States)

    Denschlag, J. O.

    This chapter first gives a survey on the history of the discovery of nuclear fission. It briefly presents the liquid-drop and shell models and their application to the fission process. The most important quantities accessible to experimental determination such as mass yields, nuclear charge distribution, prompt neutron emission, kinetic energy distribution, ternary fragment yields, angular distributions, and properties of fission isomers are presented as well as the instrumentation and techniques used for their measurement. The contribution concentrates on the fundamental aspects of nuclear fission. The practical aspects of nuclear fission are discussed in http://dx.doi.org/10.1007/978-1-4419-0720-2_57 of Vol. 6.

  15. Subsurface Biogeochemistry of Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Kersting, Annie B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Univ. Relations and Science Education; Zavarin, Mavrik [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Glenn T. Seaborg Inst.

    2016-06-29

    A major scientific challenge in environmental sciences is to identify the dominant processes controlling actinide transport in the environment. It is estimated that currently, over 2200 metric tons of plutonium (Pu) have been deposited in the subsurface worldwide, a number that increases yearly with additional spent nuclear fuel (Ewing et al., 2010). Plutonium has been shown to migrate on the scale of kilometers, giving way to a critical concern that the fundamental biogeochemical processes that control its behavior in the subsurface are not well understood (Kersting et al., 1999; Novikov et al., 2006; Santschi et al., 2002). Neptunium (Np) is less prevalent in the environment; however, it is predicted to be a significant long-term dose contributor in high-level nuclear waste. Our focus on Np chemistry in this Science Plan is intended to help formulate a better understanding of Pu redox transformations in the environment and clarify the differences between the two long-lived actinides. The research approach of our Science Plan combines (1) Fundamental Mechanistic Studies that identify and quantify biogeochemical processes that control actinide behavior in solution and on solids, (2) Field Integration Studies that investigate the transport characteristics of Pu and test our conceptual understanding of actinide transport, and (3) Actinide Research Capabilities that allow us to achieve the objectives of this Scientific Focus Area (SFA and provide new opportunities for advancing actinide environmental chemistry. These three Research Thrusts form the basis of our SFA Science Program (Figure 1).

  16. In-Plant Fission Product Behavior in SGTR Accident with Long-Term SBO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Han, Seok Jung; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The off-site AC power was recovered in 9 days after the accident in the NPS. Therefore safety injection by fire pump truck with fresh water or seawater is only available in the Fukushima accident. However, safety injection by fire pump truck is not always effective due to the high pressure of RPV inside or leakages of alternative water injection flow paths. In the SBO situations in pressurized water reactor plant (PWR), turbine driven auxiliary feedwater (TD-AFW) pump can inject water to the secondary side of steam generator. However, turbine inlet steam flow control valve cannot work properly when loss of vital DC power occurs. Vital DC power is designed to be maintained during 4 or 8 hours in the SBO conditions. In this paper motor-driven and turbine driven AFW pumps are all assumed to be not working at time 0 sec as a worst case assumption. Iodine pool-scrubbing can occur in the secondary side of the faulted steam generator. However, iodine pool-scrubbing in the secondary side of the faulted steam generator is assumed not to be working, due to the assumption of the loss of DC battery for turbine inlet flow control valve. Iodine pool-scrubbing is one of the long-term research issues in safety assessment of nuclear power plant severe accident. PHEBUS FPT series and THAI experiment projects are typical projects on the resolving source term issues in severe accident of nuclear power plants. However, iodine retention by pool scrubbing is still a debating issue. In such containment bypass sequences, fission products can be released out to environment directly from RCS without retention or deposition in containment structures. SGTR is one of the hazardous accident scenarios in the typical PSA, because SGTR induces a large release amount of source term to environment directly. A key operation strategy was the isolation of the broken reactor coolant system loop from the intact loop. Typical core degradation in SGTR scenarios occurs with multiple failures of the isolation

  17. Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın

    2011-07-01

    Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

  18. Chemical reactions of fission product deposits and iodine transport in primary circuit conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kalilainen, Jarmo, E-mail: jarmo.kalilainen@vtt.fi [VTT Technical Research Centre of Finland, P.O. Box 1000, BI7, FI02044 VTT Espoo (Finland); Kärkelä, Teemu; Zilliacus, Riitta; Tapper, Unto; Auvinen, Ari [VTT Technical Research Centre of Finland, P.O. Box 1000, BI7, FI02044 VTT Espoo (Finland); Jokiniemi, Jorma [VTT Technical Research Centre of Finland, P.O. Box 1000, BI7, FI02044 VTT Espoo (Finland); University of Eastern Finland, Department of Environmental Science, P.O. Box 1627, FI-70211 Kuopio (Finland)

    2014-02-15

    Highlights: • With CsI precursor, 20% of released iodine was in gaseous form in steam flow at 650 °C. • In similar conditions with Mo, gaseous iodine fraction was increased to 38–79%. • Boron trapped most of the caesium and iodine was almost completely released as gas. - Abstract: The objective of this work was to examine the chemical reactions taking place on primary circuit surfaces and their effect on fission product transport in a severe nuclear reactor accident. Especially transport of gaseous and aerosol phase iodine was studied. Caesium iodide (CsI) was used as precursor material for iodine species. Also, effects of molybdenum and boron on transport of iodine were investigated. The experimental work showed that when CsI alone was used as a precursor, as much as 20% of the released iodine was in gaseous form and the rest as aerosol particles. Aerosol particles were most likely CsI. When the amount of hydrogen in the carrier gas was increased, the fraction of gaseous iodine decreased. When Boron was added to the precursor, a glassy caesium borate surface was formed on the crucible. Boron trapped most of the caesium and also a fraction of iodine, causing almost all released iodine to be in gaseous form. When Mo was introduced in the precursor, most of the iodine was again released in gaseous form. Oxidised Mo reacted with caesium releasing iodine from CsI. The effect of Mo on iodine transport depended much on H{sub 2} concentration and was observed to be substantially greater on stainless steel surface. When stainless steel crucible was used, Mo was found in small amounts from aerosol particles, indicating that it was probably released as caesium molybdate or as molybdenum oxide.

  19. ASTEC V2 severe accident integral code: Fission product modelling and validation

    Energy Technology Data Exchange (ETDEWEB)

    Cantrel, L., E-mail: laurent.cantrel@irsn.fr; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-06-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix.

  20. NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

    Energy Technology Data Exchange (ETDEWEB)

    OH,S.Y.; CHANG,J.; MUGHABGHAB,S.

    2000-05-11

    Neutron cross section evaluations of the fission-product isotopes, {sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, {sup 141}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd, and {sup 157}Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of {sup 155}Gd and {sup 157}Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations.

  1. The Harwell Buechner magnet/PSD system as a mass spectrometer for unslowed fission products

    CERN Document Server

    Conlon, T W

    1976-01-01

    Heavy ion reaction studies (mass approximately 20) at the Harwell Tandem Accelerator utilise a Buechner magnet and position sensitive counters for spectroscopy. The feasibility of using these facilities for spectroscopic studies of fission fragments (mass approximately 80-150) is briefly described. (3 refs).

  2. Heavy ion induced damage in MgAl sub 2 O sub 4 , an inert matrix candidate for the transmutation of minor actinides

    CERN Document Server

    Wiss, T

    1999-01-01

    Magnesium aluminum spinel (MgAl sub 2 O sub 4) is a material selected as a possible matrix for transmutation of minor actinides by neutron capture or fission in nuclear reactors. To study the radiation stability of this inert matrix, especially against fission product impact, irradiations with heavy energetic ions or clusters have been performed. The high electronic energy losses of the heavy ions in this material led to the formation of visible tracks as evidenced by transmission electron microscopy for 30 MeV C sub 6 sub 0 -Buckminster fullerenes and for ions of energy close to or higher than fission energy ( sup 2 sup 0 sup 9 Bi with 120 MeV and 2.38 GeV energy). The irradiations at high energies showed a pronounced degradation of the spinel. Additionally, MgAl sub 2 O sub 4 exhibited a large swelling for irradiation at high fluences with fission products of fission energy (here I-ions of 72 MeV) and at temperatures <= 500 deg. C. These observations are discussed from the technological point of view in ...

  3. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    Science.gov (United States)

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-01

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still

  4. Neutron capture cross-section of fission products in the European activation file EAF-3. Presented at the NEA Specialists` Meeting on Fission Product Nuclear Data, JAERI, Japan, 25-27 May 1992

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-05-01

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

  5. Development of a fission product transport module predicting the behavior of radiological materials during sever accidents in a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyung Seok; Rhee, Bo Wook; Kim, Dong Ha [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-09-15

    Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ±6%. It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

  6. Decay Heat Analyses after Thermal-Neutron Fission of {sup 235}U and {sup 239}Pu by SCALE-6.1.3 with Recently Available Fission Product Yield Data

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The heat reaches about 1.5% after one hour and falls to 0.4% after a day. After a week it will be about 0.2%. The reactor, however, still requires further cooling for several years to keep the fuel rods safe. In general, the decay heat in the reactors can be calculated using a summation calculation method, which is simply the sum of the activities of the fission products produced during the fission process and after the reactor shutdown weighted by the mean decay energies. Consequently, the method is strongly dependent on the available nuclear structure data. Nowadays, the method has been implemented in various burnup and depletion programs such as ORIGEN and CINDER. In this study, the decay heat measurements after thermal-neutron fission of {sup 235}U and {sup 239}Pu have been evaluated by the ORIGEN-S with the decay data and fission product yield libraries included in the SCALE-6.1.3 software package. The new libraries were applied to the decay heat calculations, and the results were compared with those by the ORIGEN reference calculation. The decay heat measurements for very short cooling times after thermal-neutron fission of {sup 235}U and {sup 239}Pu have been evaluated by the ORIGEN-S summation calculation. The reference calculation results by the latest ORIGEN data libraries of the SCALE-6.1.3 have been validated with the measurements by ORNL and Studsvik. In addition, the generation of the new ORIGEN yield libraries has been completed based on the ENDF/B-VII.1, JEFF-3.1.1, JENDL/FPY-2011, and JENDL-4.0. The new libraries have been successfully applied to the decay heat calculations and comparative analyses have been devoted to verifying the importance of the fission product yield data when estimating the decay heat values for each isotope in a very short time. The decay data library occupies an important position in the ORIGEN summation calculation along with the fission product yield library.

  7. Recent actinide nuclear data efforts with the DANCE 4{pi} BaF{sub 2} array

    Energy Technology Data Exchange (ETDEWEB)

    Bredeweg, T.A.; Bond, E.M.; Couture, A.J.; Fitzpatrick, J.R.; Haight, R.C.; Hill, T.S.; Jandel, M.; O' Donnell, J.M.; Reifarth, R.; Rundberg, R.S.; Slemmons, A.K.; Tovesson, F.K.; Ullmann, J.L.; Vieira, D.J.; Wilhelmy, J.B.; Fowler, M.M.; Wouters, J.M. [Los Alamos National Laboratory, Los Alamos, NM (United States); Agvaanluvsan, U.; Becker, J.A.; Macri, R.A.; Parker, W.E.; Wilk, P.A.; Wu, C.Y. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Ethvignot, T.; Granier, T. [CEA Bruyeres-le-Chatel, 91 (France)

    2008-07-01

    Much of the recent work in the DANCE collaboration has focused on nuclides of interest to stockpile stewardship, attribution science and the advanced fuel cycle initiative. As an example, we have recently begun a program to produce high precision measurements of the key production and destruction reactions of important nuclear fuel elements and radiochemical diagnostic isotopes. The neutron capture targets that have been fielded under this program include several isotopes of uranium, plutonium and americium. However, neutron capture measurements on many of the actinides are complicated by the presence of {gamma}-rays arising from low energy neutron-induced fission. To overcome this difficulty we have designed and implemented a dual parallel-plate avalanche counter fission-tagging detector. This design provides a high efficiency for detecting fission fragments and is self-contained to allow loading of pre-assembled target/detector assemblies into the neutron beam line at DANCE. Neutron capture measurements have been performed on {sup 234,235,236}U. The results for {sup 236}U are consistent with the Endf/B-6 evaluation while the results for {sup 234}U are as much as 20% lower than the Endf/B-6 evaluation in the keV region. The DANCE results for {sup 234}U(n,{gamma}) have been incorporated into the Endf/B-7 evaluation. Planned measurements on {sup 238,239}Pu are also discussed.

  8. The Fission Programme at the CERN n_TOF Facility

    Science.gov (United States)

    Tsinganis, A.; Barbagallo, M.; Berthoumieux, E.; Calviani, M.; Chiaveri, E.; Colonna, N.; Diakaki, M.; Duran, I.; Guerrero, C.; Gunsing, F.; Leal-Cidoncha, E.; Leong, L.-S.; Paradela, C.; Tarrio, D.; Tassan-Got, L.; Vlastou, R.

    Since 2001, the scientific programme of the CERN n_TOF facility has focused mainly on the study of radiative neutron capture reactions, which are of great interest to nuclear astrophysics and on neutron-induced fission reactions, which are of relevance for nuclear technology, as well as essential for the development of theoretical models of fission. In particular, taking advantage of the high instantaneous neutron flux and high energy resolution of the facility, as well as of high-performance detection and acquisition systems, accurate new measurements on several long-lived major and minor actinides, from 232Th to 245Cm, have been performed so far. Data on these isotopes are needed in order to improve the safety and efficiency of conventional reactors, as well as to develop new systems for nuclear energy production and treatment of nuclear waste, such as Generation IV reactors, Accelerator Driven Systems and reactors based on innovative fuel cycles. A review of the most important results on fission cross-sections and fragment properties obtained at n_TOF for a variety of (radioactive) isotopes is presented along with the perspectives arising from the coming on line in the second half of 2014 of a new 19 m flight-path, which will allow n_TOF to expand its measurement capabilities to even more rare or short-lived isotopes, such as 230Th, 232U, 238,240Pu and 244Cm.

  9. Study of the actinide-lanthanide separation from nuclear waste by a new pyrochemical process; Etude de la separation actinides-lanthanides des dechets nucleaires par un procede pyrochimique nouveau

    Energy Technology Data Exchange (ETDEWEB)

    Lemort, F. [CEA Marcoule, Departement de Retraitement, des Dechets et du Demantelement, 30 - Bagnols-sur-Ceze (France)]|[Institut National Polytechnique, 38 - Grenoble (France)

    1997-01-01

    The theoretical extraction and separation of platinoids, actinides and lanthanides is allowed by thermodynamic using two adapted reducing agents: zinc and magnesium. Thereby, a pyrochemical method for the nuclear waste processing has been devised. The high temperature handling of the elements in fluoride forms and their processing by a reactive metallic phase required special precautions. The study of the behavior of matter in exploratory systems allowed the development of an experimental technology for the treatment and contacting of phases. The thermodynamical analysis of the experimental results shows the feasibility of the process. A model was developed to predict the distribution coefficients of zirconium, uranium and lanthanum as a function of the system composition. An estimation method was proposed in order to evaluate the distribution coefficients in diluted solution of all the actinides and lanthanides existing in the fission products between LiF CaF{sub 2} and Zn-Mg at 720 deg C. Coupled with the experimental results, the estimates results may be extrapolated to concentrated solutions allowing predictions of the separation of all actinides and lanthanides. The rapidity of element transfer is induced by a thermal effect caused by the high exothermicity of the reduction by magnesium. The kinetic coefficients have been linked with the reduction enthalpy of each element. Moreover, the kinetics seem limited by chemical reaction and not by mass transfer. (author) 66 refs.

  10. Actinide transmutation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bultman, J.H.

    1995-01-17

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP).

  11. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    Science.gov (United States)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  12. Comparison of predicted and measured fission product behavior in the Fort St. Vrain HTGR during the first three cycles of operation

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.L.; Jovanovic, V.; Burnette, R.D.

    1985-10-01

    Fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors which is consistent with plateout probe measurements.

  13. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Leng, B. [University of Wisconsin-Madison, Madison, WI 53706 (United States); Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Shanghai, 201800 (China); Rooyen, I.J. van, E-mail: Isabella.vanrooyen@inl.gov [Fuel Design and Development Department, Idaho National Laboratory, Idaho Falls, ID 83415-6188 (United States); Wu, Y.Q. [Department of Materials Science and Engineering, Boise State University, Boise, ID 83725-2090 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Szlufarska, I.; Sridharan, K. [University of Wisconsin-Madison, Madison, WI 53706 (United States)

    2016-07-15

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd{sub 2}Si{sub 2}U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously. - Highlights: • First research data in neutron irradiated TRISO coated particles showing a Ag-Pd nano-sized precipitate inside a SiC grain. • Intragranular Ag Pd

  14. Arrival time and magnitude of airborne fission products from the Fukushima, Japan, reactor incident as measured in Seattle, WA, USA

    CERN Document Server

    Leon, J Diaz; Knecht, A; Miller, M L; Robertson, R G H; Schubert, A G

    2011-01-01

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

  15. Mechanical environmental transport of actinides and ¹³⁷Cs from an arid radioactive waste disposal site.

    Science.gov (United States)

    Snow, Mathew S; Clark, Sue B; Morrison, Samuel S; Watrous, Matthew G; Olson, John E; Snyder, Darin C

    2015-10-01

    Aeolian and pluvial processes represent important mechanisms for the movement of actinides and fission products at the Earth's surface. Soil samples taken in the early 1970's near a Department of Energy radioactive waste disposal site (the Subsurface Disposal Area, SDA, located in southeastern Idaho) provide a case study for studying the mechanisms and characteristics of environmental actinide and (137)Cs transport in an arid environment. Multi-component mixing models suggest actinide contamination within 2.5 km of the SDA can be described by mixing between 2 distinct SDA end members and regional nuclear weapons fallout. The absence of chemical fractionation between (241)Am and (239+240)Pu with depth for samples beyond the northeastern corner and lack of (241)Am in-growth over time (due to (241)Pu decay) suggest mechanical transport and mixing of discrete contaminated particles under arid conditions. Occasional samples northeast of the SDA (the direction of the prevailing winds) contain anomalously high concentrations of Pu with (240)Pu/(239)Pu isotopic ratios statistically identical to those in the northeastern corner. Taken together, these data suggest flooding resulted in mechanical transport of contaminated particles into the area between the SDA and a flood containment dike in the northeastern corner, following which subsequent contamination spreading in the northeastern direction resulted from wind transport of discrete particles. Copyright © 2015 Elsevier Ltd. All rights reserved.

  16. Measurements of Fission Cross Sections for the Isotopes relevant to the Thorium Fuel Cycle

    CERN Multimedia

    2002-01-01

    The present concern about a sustainable energy supply is characterised by a considerable uncertainty: the green house effect and foreseeable limits in fossil fuel resources on the one hand, the concern about the environmental impact of nuclear fission energy and the long term fusion research on the other hand, have led to the consideration of a variety of advanced strategies for the nuclear fuel cycle and related nuclear energy systems. The present research directories concern such strategies as the extension of the life span of presently operating reactors, the increase of the fuel burn-up, the plutonium recycling, and in particular the incineration of actinides and long-Lived fission products, the accelerator driven systems (ADS), like the "Energy Amplifier" (EA) concept of C. Rubbia, and the possible use of the Thorium fuel cycle. The detailed feasibility study and safety assessment of these strategies requires the accurate knowledge of neutron nuclear reaction data. Both, higher fuel burn-up and especiall...

  17. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    Science.gov (United States)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100 days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths. SUP>235U-rich agglomerates, SUP>235U-poor areas, an intermediate phase with intermediate 235U concentrations. Short fuel rods were fabricated with these pellets. The main characteristics of these fuel rods are shown in Table 1.These rods were irradiated to high burn-ups in the IFA-609/626 of the HBWR and then one was irradiated in the IFA-702 for 100 days. Fig. 2 shows the irradiation history of this fuel. The final average burn-up of the rod was 69 GWd/tU. Due to the flux differences along the rod, however, the average burn-up of the cross section examined was 63 GWd/tU. This fuel

  18. Spectroscopy of heavy fissionable nuclei

    Indian Academy of Sciences (India)

    S K Tandel

    2015-09-01

    Structural studies of heavy nuclei are quite challenging due to increased competition from fission, particularly at high spins. Nuclei in the actinide region exhibit a variety of interesting phenomena. Recent advances in instrumentation and analysis techniques have made feasible sensitive measurements of nuclei populated with quite low cross-sections. These include isomers and rotational band structures in isotopes of Pu ( = 94) to Rf ( = 104), and octupole correlations in the Th ( = 90) region. The obtained experimental data have provided insights on various aspects like moments of inertia and nucleon alignments at high spins, quasiparticle energies and evolution of quadrupole and octupole collectivity, among others. An overview of some of these results is presented.

  19. Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor.

    Science.gov (United States)

    Košťál, Michal; Švadlenková, Marie; Koleška, Michal; Rypar, Vojtěch; Milčák, Ján

    2015-11-01

    Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins.

  20. Cement As a Waste Form for Nuclear Fission Products: The Case of (90)Sr and Its Daughters.

    Science.gov (United States)

    Dezerald, Lucile; Kohanoff, Jorge J; Correa, Alfredo A; Caro, Alfredo; Pellenq, Roland J-M; Ulm, Franz J; Saúl, Andrés

    2015-11-17

    One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DFT) to study the effects of (90)Sr insertion and decay in C-S-H (calcium-silicate-hydrate) in order to test the ability of cement to trap and hold this radioactive fission product and to investigate the consequences of its β-decay on the cement paste structure. We show that (90)Sr is stable when it substitutes the Ca(2+) ions in C-S-H, and so is its daughter nucleus (90)Y after β-decay. Interestingly, (90)Zr, daughter of (90)Y and final product in the decay sequence, is found to be unstable compared to the bulk phase of the element at zero K but stable when compared to the solvated ion in water. Therefore, cement appears as a suitable waste form for (90)Sr storage.

  1. Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

    2010-09-23

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

  2. New measurements of (n,γ) and (n,fission) cross sections and capture-to-fission ratios for ^233,235U and ^239Pu using the DANCE 4π BaF2 array

    Science.gov (United States)

    Bredeweg, T. A.; Jandel, M.; Fowler, M. M.; Bond, E. M.; Haight, R. C.; Keksis, A. L.; O'Donnell, J. M.; Reifarth, R.; Rundberg, R. S.; Ullmann, J. L.; Vieira, D. J.; Wilhelmy, J. B.; Wouters, J. M.; Becker, J. A.; Parker, W. E.; Wu, C. Y.; Baker, J. D.; McGrath, C. A.

    2009-10-01

    Accurate neutron nuclear data are important to many issues in stockpile stewardship, nuclear reactor design and re-certification, nuclear non-proliferation and nuclear forensics. Of particular interest are the production and destruction reactions for all of the major and most of the minor actinides. The competition between capture and fission in many of the actinides presents both an obstacle and an opportunity for large γ detector arrays such as DANCE. Additional instrumentation is required to deconvolve the two contributions to the total observed γ-ray spectrum. However, conducting a simultaneous measurement can simplify background treatment and other sources of systematic uncertainty. An outline of the current experimental program will be presented along with results from neutron capture measurements on ^233,235U and ^239Pu.

  3. Summary Report: Glass-Ceramic Waste Forms for Combined Fission Products

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Riley, Brian J.; Turo, Laura A.; Tang, Ming; Kossoy, Anna

    2011-09-23

    Glass-ceramic waste form development began in FY 2010 examining two combined waste stream options: (1) alkaline earth (CS) + lanthanide (Ln), and (2) + transition metal (TM) fission-product waste streams generated by the uranium extraction (UREX+) separations process. Glass-ceramics were successfully developed for both options however; Option 2 was selected over Option 1, at the conclusion of 2010, because Option 2 immobilized all three waste streams with only a minimal decrease in waste loading. During the first year, a series of three glass (Option 2) were fabricated that varied waste loading-WL (42, 45, and 50 mass%) at fixed molar ratios of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali both at 1.75. These glass-ceramics were slow cooled and characterized in terms of phase assemblage and preliminary irradiation stability. This fiscal year, further characterization was performed on the FY 2010 Option 2 glass-ceramics in terms of: static leach testing, phase analysis by transmission electron microscopy (TEM), and irradiation stability (electron and ion). Also, a new series of glass-ceramics were developed for Option 2 that varied the additives: Al{sub 2}O{sub 3} (0-6 mass%), molar ratio of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali (1.75 to 2.25) and waste loading (50, 55, and 60 mass%). Lastly, phase pure powellite and oxyapatite were synthesized for irradiation studies. Results of this fiscal year studies showed compositional flexibility, chemical stability, and radiation stability in the current glass-ceramic system. First, the phase assemblages and microstructure of all of the FY 2010 and 2011 glass-ceramics are very similar once subjected to the slow cool heat treatment. The phases identified in these glass-ceramics were oxyapatite, powellite, cerianite, and ln-borosilicate. This shows that variations in waste loading or additives can be accommodated without drastically changing the phase assemblage of the waste form, thus making the processing and performance

  4. Fission Fragment characterization with FALSTAFF at NFS

    Directory of Open Access Journals (Sweden)

    Pancin J.

    2013-03-01

    Full Text Available The Neutrons for Science (NFS facility will be one of the first installations of the SPIRAL2 facility. NFS will be composed of a time-of-flight baseline and irradiation stations and will allow studying neutron-induced reactions for energies going from some hundreds of keV up to 40 MeV. Continuous and quasi-monoenergetic energy neutron beams will be available. Taking advantage of this new installation, the development of an experimental setup for a full characterization of actinide fission fragments in this energy domain has been undertaken. To achieve this goal a new detection system called FALSTAFF (Four Arm cLover for the STudy of Actinide Fission Fragments in under development. In this paper, the characteristics of the NFS facility will be exposed and the motivations for the FALSTAFF experiment will be presented. The experimental setup will be described and the expected resolutions based on realistic GEANT4 simulations will be discussed.

  5. True ternary fission

    Science.gov (United States)

    Vijayaraghavan, K. R.; Balasubramaniam, M.; von Oertzen, W.

    2015-04-01

    The study of the ternary fission of nuclei has received new interest recently. It is of general interest for nuclear dynamics, although the process is very rare. In the present work, we discuss the possibilities of true ternary fission (fragment masses A >30 ) in 252Cf for different mass splits. These mass splits are strongly favored in a collinear geometry. Based on the three cluster model (TCM), it is shown that the true ternary fission into fragments with almost equal masses is one of the possible fission modes in 252Cf . For general decays it is shown that the formation of the lightest fragment at the center has the highest probability. Further the formation of tin isotopes and/or other closed shell fragments are favored. For the decay products the presence of closed shell nuclei among the three fragments enhances the decay probabilities.

  6. Thermochemical prediction of chemical form distributions of fission products in LWR oxide fuels irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-09-01

    Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. In addition, the effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that approximately 13 mole% of the total amount of Cs compounds exists as CsI and virtually remaining fraction as Cs{sub 2}MoO{sub 4} under the operation condition of LHR below 400 W/cm. On the other hand, when LHR is beyond 400 W/cm under the transient operation condition, its distribution did not change so much from the one under normal operation condition. (author)

  7. Fission yield measurements at IGISOL

    Directory of Open Access Journals (Sweden)

    Lantz M.

    2016-01-01

    Full Text Available The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f and Th(p,f have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  8. Exact Solution of Fractional Diffusion Model with Source Term used in Study of Concentration of Fission Product in Uranium Dioxide Particle*

    Institute of Scientific and Technical Information of China (English)

    FANG Chao; CAO Jian-Zhu; SUN Li-Feng

    2011-01-01

    The exact solution of fractional diffusion model with a location-independent source term used in the study of the concentration of fission product in spherical uranium dioxide (U02) particle is built. The adsorption effect of the fission product on the surface of the U02 particle and the delayed decay effect are also considered. The solution is given in terms of Mittag-Leffler function with finite Hankel integral transformation and Laplace transformation. At last, the reduced forms of the solution under some special physical conditions, which is used in nuclear engineering, are obtained and corresponding remarks are given to provide significant exact results to the concentration analysis of nuclear fission products in nuclear reactor.

  9. Measurements of extinct fission products in nuclear bomb debris: Determination of the yield of the Trinity nuclear test 70 y later.

    Science.gov (United States)

    Hanson, Susan K; Pollington, Anthony D; Waidmann, Christopher R; Kinman, William S; Wende, Allison M; Miller, Jeffrey L; Berger, Jennifer A; Oldham, Warren J; Selby, Hugh D

    2016-07-19

    This paper describes an approach to measuring extinct fission products that would allow for the characterization of a nuclear test at any time. The isotopic composition of molybdenum in five samples of glassy debris from the 1945 Trinity nuclear test has been measured. Nonnatural molybdenum isotopic compositions were observed, reflecting an input from the decay of the short-lived fission products (95)Zr and (97)Zr. By measuring both the perturbation of the (95)Mo/(96)Mo and (97)Mo/(96)Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the (95)Zr and (97)Zr isotopes formed in the nuclear detonation. Together with a determination of the amount of plutonium in the debris, these measurements of extinct fission products allow for new estimates of the efficiency and yield of the historic Trinity test.

  10. Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

    Directory of Open Access Journals (Sweden)

    Jaroszewicz Janusz

    2014-07-01

    Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

  11. Photofission of actinide and pre-actinide nuclei in the quasideuteron and delta energy regions

    CERN Document Server

    Berman, B L; Cole, P L; Dodge, W R; Feldman, G; Sanabria, J C; Kolb, N; Pywell, R E; Vogt, J; Nedorezov, V; Sudov, A; Kezerashvili, G Ya

    1999-01-01

    The photofission cross sections for the actinide nuclei sup 2 sup 3 sup 2 Th, sup 2 sup 3 sup 3 sup , sup 2 sup 3 sup 5 sup , sup 2 sup 3 sup 8 U, and sup 2 sup 3 sup 7 Np have been measured from 68 to 264 MeV and those for the pre-actinide nuclei sup 1 sup 9 sup 7 Au and sup N sup A sup T Pb from 122 to 222 MeV at the Saskatchewan Accelerator Laboratory, using monoenergetic tagged photons and novel parallel-plate avalanche detectors for the fission fragments. The aim of the experiment was to obtain a comprehensive and self-consistent data set and to investigate previous anomalous results in this energy region. The fission probability for transuranic nuclei is expected to be close to unity here. However, important discrepancies have been confirmed for sup 2 sup 3 sup 7 Np and sup 2 sup 3 sup 2 Th, compared with sup 2 sup 3 sup 8 U, which have serious implications for the inferred total photoabsorption strengths, and hence call into question the 'Universal Curve' for photon absorption at these energies. High-s...

  12. Prompt Fission Gamma-ray Studies at DANCE

    Science.gov (United States)

    Jandel, M.; Rusev, G.; Bond, E. M.; Bredeweg, T. A.; Chadwick, M. B.; Couture, A.; Fowler, M. M.; Haight, R. C.; Kawano, T.; Keksis, A. L.; Mosby, S. M.; O'Donnell, J. M.; Rundberg, R. S.; Stetcu, I.; Talou, P.; Ullmann, J. L.; Vieira, D. J.; Wilhelmy, J. B.; Stoyer, M. A.; Haslett, R. J.; Henderson, R. A.; Becker, J. A.; Wu, C. Y.

    Measurements of correlated data on prompt-fission γ-rays (PFG) have been carried out for various actinide isotopes in recent years using the Detector for Advanced Neutron Capture Experiments (DANCE) at Los Alamos National Laboratory (LANL). We have developed a model that conveniently parametrizes the correlated data of γ-ray multiplicity and energy. New results on two- dimensional prompt-fission γ-ray multiplicity versus energy distributions from spontaneous fission on 252Cf and neutron-induced fission on 242mAm are presented together with previously obtained results on 233,235U and 239Pu. Correlated PFG data from 252Cf are also compared to results of the detailed theoretical model developed at LANL, for different thresholds of PFG energies. Future plans to measure correlated data on fission fragments, prompt fission neutrons and γ-rays at DANCE are presented.

  13. Simultaneous measurement of (n, γ) and (n, fission) cross sections with the DANCE 4π BaF 2 array

    Science.gov (United States)

    Bredeweg, T. A.; Fowler, M. M.; Becker, J. A.; Bond, E. M.; Chadwick, M. B.; Clement, R. R. C.; Esch, E.-I.; Ethvignot, T.; Granier, T.; Jandel, M.; Macri, R. A.; O'Donnell, J. M.; Reifarth, R.; Rundberg, R. S.; Ullmann, J. L.; Vieira, D. J.; Wilhelmy, J. B.; Wouters, J. M.; Wu, C. Y.

    2007-08-01

    We have recently begun a program of high precision measurements of the key production and destruction reactions of important radiochemical diagnostic isotopes, including several isotopes of uranium, plutonium and americium. The detector for advanced neutron capture experiments (DANCE), a 4π BaF2 array located at the Los Alamos Neutron Science Center, will be used to measure the neutron capture cross sections for most of the isotopes of interest. However, neutron capture measurements on many of the actinides are complicated by the presence of prompt γ-rays arising from low energy neutron-induced fission, which competes with neutron capture to varying degrees. Previous measurements of 235U using the DANCE array have shown that we can partially resolve capture from fission events based on total γ-ray calorimetry (i.e. total γ-ray energy versus γ-ray multiplicity). The addition of a dedicated fission-tagging detector to the DANCE array has greatly improved our ability to separate these two competing processes. In addition to higher quality neutron capture data, the addition of a fission-tagging detector offers a means to determine the capture-to-fission ratio (σγ/σf) in a single measurement, which should reduce the effect of systematic uncertainties. We are currently using a dual parallel-plate avalanche counter (PPAC) with the target material electro-deposited directly on the center cathode foil. This design provides a high efficiency for detecting fission fragments and allows loading of pre-assembled target/detector assemblies into the neutron beam line at DANCE. Results from tests of the fission-tag detector, as well as preliminary results from measurements on 235U and 252Cf that utilized the fission-tag detector will be presented.

  14. Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

    Science.gov (United States)

    Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

    2016-12-01

    Efficient and safe production of molybdenum-99 ((99)Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced (99)Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient (99)Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

  15. Observation of new microsecond isomers among fission products of 345 MeV/nucleon 238U

    CERN Document Server

    Kameda, D; Ohnishi, T; Kusaka, K; Yoshida, A; Yoshida, K; Ohtake, M; Fukuda, N; Takeda, H; Tanaka, K; Inabe, N; Yanagisawa, Y; Gono, Y; Watanabe, H; Otsu, H; Baba, H; Ichihara, T; Yamaguchi, Y; Takechi, M; Nishimura, S; Ueno, H; Yoshimi, A; Sakurai, H; Motobayashi, T; Nakao, T; Mizoi, Y; Matsushita, M; Ieki, K; Kobayashi, N; Tanaka, K; Kawada, Y; Tanaka, N; Deguchi, S; Satou, Y; Kondo, Y; Nakamura, T; Yoshinaga, K; Ishii, C; Yoshii, H; Miyashita, Y; Uematsu, N; Shiraki, Y; Sumikama, T; Chiba, J; Ideguchi, E; Saito, A; Yamaguchi, T; Hachiuma, I; Suzuki, T; Moriguchi, T; Ozawa, A; Ohtsubo, T; Famiano, M A; Geissel, H; Nettleton, A S; Tarasov, O B; Bazin, D; Sherrill, B M; Manikonda, S L; Nolen, J A

    2012-01-01

    A search for isomeric gamma-decays among fission fragments from 345 MeV/nucleon 238U has been performed at the RIKEN Nishina Center RI Beam Factory. Fission fragments were selected and identified using the superconducting in-flight separator BigRIPS and were implanted in an aluminum stopper. Delayed gamma-rays were detected using three clover-type high-purity germanium detectors located at the focal plane within a time window of 20 microseconds following the implantation. We identified a total of 54 microsecond isomers with half-lives of ~0.1 - 10 microseconds, including discovery of 18 new isomers in very neutron-rich nuclei: 59Tim, 90Asm, 92Sem, 93Sem, 94Brm, 95Brm, 96Brm, 97Rbm, 108Nbm, 109Mom, 117Rum, 119Rum, 120Rhm, 122Rhm, 121Pdm, 124Pdm, 124Agm and 126Agm, and obtained a wealth of spectroscopic information such as half-lives, gamma-ray energies, gamma-ray relative intensities and gamma-gamma coincidences over a wide range of neutron-rich exotic nuclei. Proposed level schemes are presented for 59Tim, 82...

  16. Standard test method for gamma energy emission from fission products in uranium hexafluoride and uranyl nitrate solution

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This test method covers the measurement of gamma energy emitted from fission products in uranium hexafluoride (UF6) and uranyl nitrate solution. It is intended to provide a method for demonstrating compliance with UF6 specifications C 787 and C 996 and uranyl nitrate specification C 788. 1.2 The lower limit of detection is 5000 MeV Bq/kg (MeV/kg per second) of uranium and is the square root of the sum of the squares of the individual reporting limits of the nuclides to be measured. The limit of detection was determined on a pure, aged natural uranium (ANU) solution. The value is dependent upon detector efficiency and background. 1.3 The nuclides to be measured are106Ru/ 106Rh, 103Ru,137Cs, 144Ce, 144Pr, 141Ce, 95Zr, 95Nb, and 125Sb. Other gamma energy-emitting fission nuclides present in the spectrum at detectable levels should be identified and quantified as required by the data quality objectives. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its us...

  17. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Susmikanti, Mike, E-mail: mike@batan.go.id [Center for Development of Nuclear Informatics, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Dewayatna, Winter, E-mail: winter@batan.go.id [Center for Nuclear Fuel Technology, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Sulistyo, Yos, E-mail: soj@batan.go.id [Center for Nuclear Equipment and Engineering, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia)

    2014-09-30

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo{sup 99} used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 10{sup 6} cm{sup −1}) in a tube, their delta

  18. Mechanical behaviour and diffusion of gas during neutron irradiation of actinides in ceramic inert matrices

    NARCIS (Netherlands)

    Neeft, E.A.C.

    2004-01-01

    Fission of actinides from nuclear waste in inert matrices (materials without uranium) can reduce the period in time that nuclear waste is more radiotoxic than uranium ore that is the rock from which ordinary reactor fuel is made. A pioneering study is performed with the inert matrices: MgO, MgAl2O4,

  19. Mechanical behaviour and diffusion of gas during neutron irradiation of actinides in ceramic inert matrices

    NARCIS (Netherlands)

    Neeft, E.A.C.

    2004-01-01

    Fission of actinides from nuclear waste in inert matrices (materials without uranium) can reduce the period in time that nuclear waste is more radiotoxic than uranium ore that is the rock from which ordinary reactor fuel is made. A pioneering study is performed with the inert matrices: MgO, MgAl2O4,

  20. Device for Detecting Actinides, Method for Detecting Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, Fred J.; Wilkins-Stevens, Priscilla

    1998-10-29

    A heavy metal detector is provided comprising a first molecule and a second molecule, whereby the first and second molecules interact in a predetermined manner; a first region on the first molecule adapted to interact with an actinide; and a second region on the second molecule adapted to interact with the actinide, whereby the interactions of the actinide with the regions effect the predetermined manner of interaction between the molecules.

  1. Synthesis report on the relevant diffusion coefficients of fission products and helium in spent nuclear fuels; Rapport de synthese sur les coefficients de diffusion des produits de fission et de l'helium dans le combustible irradie

    Energy Technology Data Exchange (ETDEWEB)

    Lovera, P.; Ferry, C.; Poinssot, Ch. [CEA Saclay, Dept. de Physico-Chimie (DPC), 91 - Gif sur Yvette (France); Johnson, L. [Nagra, Baden (Switzerland)

    2003-07-01

    This document corresponds to the deliverable D2 of the Work Package 1 of the 'Spent Fuel Stability under repository conditions' (SFS) European project. It constitutes a synthesis report on relevant diffusion coefficients of fission products and helium in spent nuclear fuels at high and low temperatures. Coefficients corresponding to thermally activated diffusion were reviewed from literature data for O, U (self-diffusion coefficients), fission gases and other fission products. Data showed that thermal diffusion was irrelevant at temperatures expected in repository conditions. The occurrence of diffusion enhanced by alpha self-irradiation was studied through different theoretical approaches. A 'best estimate' value of the alpha self-irradiation diffusion coefficient, D (m{sup 2}.s{sup -1}), is proposed. It is extrapolated from enhanced diffusion under irradiation observed in reactor and would be proportional to the volume alpha activity in the spent nuclear fuel, A{sub v} (Bq.m{sup -3}) as: D/A{sub v} {approx_equal} 2.10{sup -41} (m{sup 5})The migration of stable Pb in Oklo's uraninites was studied in order to validate the proposed diffusion coefficient. The obtained value is one order of magnitude higher than the theoretical proposed value. As for He behaviour in spent nuclear fuel, a few data are today available in open literature. The document will be completed as soon as new experimental results are available. (authors)

  2. Fission waves can oscillate

    CERN Document Server

    Osborne, Andrew G

    2016-01-01

    Under the right conditions, self sustaining fission waves can form in fertile nuclear materials. These waves result from the transport and absorption of neutrons and the resulting production of fissile isotopes. When these fission, additional neutrons are produced and the chain reaction propagates until it is poisoned by the buildup of fission products. It is typically assumed that fission waves are soliton-like and self stabilizing. However, we show that in uranium, coupling of the neutron field to the 239U->239Np->239Pu decay chain can lead to a Hopf bifurcation. The fission reaction then ramps up and down, along with the wave velocity. The critical driver for the instability is a delay, caused by the half-life of 239U, between the time evolution of the neutron field and the production of 239Pu. This allows the 239Pu to accumulate and burn out in a self limiting oscillation that is characteristic of a Hopf bifurcation. Time dependent results are obtained using a numerical implementation of a reduced order r...

  3. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    Energy Technology Data Exchange (ETDEWEB)

    Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2004-10-18

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

  4. Selective Separation of Trivalent Actinides from Lanthanides by Aqueous Processing with Introduction of Soft Donor Atoms

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth L. Nash

    2009-09-22

    Implementation of a closed loop nuclear fuel cycle requires the utilization of Pu-containing MOX fuels with the important side effect of increased production of the transplutonium actinides, most importantly isotopes of Am and Cm. Because the presence of these isotopes significantly impacts the long-term radiotoxicity of high level waste, it is important that effective methods for their isolation and/or transmutation be developed. Furthermore, since transmutation is most efficiently done in the absence of lanthanide fission products (high yield species with large thermal neutron absorption cross sections) it is important to have efficient procedures for the mutual separation of Am and Cm from the lanthanides. The chemistries of these elements are nearly identical, differing only in the slightly stronger strength of interaction of trivalent actinides with ligand donor atoms softer than O (N, Cl-, S). Research being conducted around the world has led to the development of new reagents and processes with considerable potential for this task. However, pilot scale testing of these reagents and processes has demonstrated the susceptibility of the new classes of reagents to radiolytic and hydrolytic degradation. In this project, separations of trivalent actinides from fission product lanthanides have been investigated in studies of 1) the extraction and chemical stability properties of a class of soft-donor extractants that are adapted from water-soluble analogs, 2) the application of water soluble soft-donor complexing agents in tandem with conventional extractant molecules emphasizing fundamental studies of the TALSPEAK Process. This research was conducted principally in radiochemistry laboratories at Washington State University. Collaborators at the Radiological Processing Laboratory (RPL) at the Pacific Northwest National Laboratory (PNNL) have contributed their unique facilities and capabilities, and have supported student internships at PNNL to broaden their

  5. Selective Separation of Trivalent Actinides from Lanthanides by Aqueous Processing with Introduction of Soft Donor Atoms

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth L. Nash

    2009-09-22

    Implementation of a closed loop nuclear fuel cycle requires the utilization of Pu-containing MOX fuels with the important side effect of increased production of the transplutonium actinides, most importantly isotopes of Am and Cm. Because the presence of these isotopes significantly impacts the long-term radiotoxicity of high level waste, it is important that effective methods for their isolation and/or transmutation be developed. Furthermore, since transmutation is most efficiently done in the absence of lanthanide fission products (high yield species with large thermal neutron absorption cross sections) it is important to have efficient procedures for the mutual separation of Am and Cm from the lanthanides. The chemistries of these elements are nearly identical, differing only in the slightly stronger strength of interaction of trivalent actinides with ligand donor atoms softer than O (N, Cl-, S). Research being conducted around the world has led to the development of new reagents and processes with considerable potential for this task. However, pilot scale testing of these reagents and processes has demonstrated the susceptibility of the new classes of reagents to radiolytic and hydrolytic degradation. In this project, separations of trivalent actinides from fission product lanthanides have been investigated in studies of 1) the extraction and chemical stability properties of a class of soft-donor extractants that are adapted from water-soluble analogs, 2) the application of water soluble soft-donor complexing agents in tandem with conventional extractant molecules emphasizing fundamental studies of the TALSPEAK Process. This research was conducted principally in radiochemistry laboratories at Washington State University. Collaborators at the Radiological Processing Laboratory (RPL) at the Pacific Northwest National Laboratory (PNNL) have contributed their unique facilities and capabilities, and have supported student internships at PNNL to broaden their

  6. Fission and Properties of Neutron-Rich Nuclei

    Science.gov (United States)

    Hamilton, Joseph H.; Ramayya, A. V.; Carter, H. K.

    2008-08-01

    spontaneous fission of [symbol]Cf / A. V. Daniel ... [et al.]. Magnetic moment measurements in a radioactive beam environment / N. Benczer-Koller and G. Kumbartzki. g-Factor measurements of picosecond states: opportunities and limitations of the recoil-in-vacuum method / N. J. Stone ... [et al.]. Precision mass measurements and trap-assisted spectroscopy of fission products from Ni to Pd / A. Jokinen -- Fission II. Fission research at IRMM / F.-J. Hambsch. Fission yield measurements at the IGISOL facility with JYFLTRAP / H. Penttilä ... [et al.]. Fission of radioactive beams and dissipation in nuclear matter / A. Heinz (for the CHARMS collaboration). Fission of [symbol]U at 80 MeVlu and search for new neutron-rich isotopes / C.M. Folden, III ... [et al.]. Measurement of the average energy and multiplicity of prompt-fission neutrons and gamma rays from [symbol], [symbol], and [symbol] for incident neutron energies of 1 to 200 MeV / R. C. Haight ... [et al.]. Fission measurements with DANCE / M. Jandel ... [et al.]. Measured and calculated neutron-induced fission cross sections of [symbol]Pu / F. Tovesson and T. S. Hill. The fission barrier landscape / L. Phair and L. G. Moretto. Fast neutron-induced fission of some actinides and sub-actinides / A. B. Lautev ... [et al.] -- Fission III/Nuclear structure III. Complex structure in even-odd staggering of fission fragment yields / M. Caamāno and F. Rejmund. The surrogate method: past, present and future / S. R. Lesher ... [et al]. Effects of nuclear incompressibility on heavy-ion fusion / H. Esbensen and Ş. Mişicu. High spin states in [symbol]Pm / A. Dhal ... [et al]. Structure of [symbol]Sm, spherical vibrator versus softly deformed rotor / J. B. Gupta -- Astrophysics. Measuring the astrophysical S-factor in plasmas / A. Bonasera ... [et al.]. Is there shell quenching or shape coexistence in Cd isotopes near N = 82? / J. K. Hwang, A. V. Ramayya and J. H. Hamilton. Spectroscopy of neutron-rich palladium and cadmium isostopes

  7. Fission Cross-Section Measurements of the Odd-Odd Isotopes 232Pa, 238Np, and 236Np

    Energy Technology Data Exchange (ETDEWEB)

    Danon, Y. [Rensselaer Polytechnic Institute; Moore, M. S. [Los Alamos National Laboratory (LANL); Koehler, Paul Edward [ORNL; Littleton, P. E. [Los Alamos National Laboratory (LANL); Miller, G. G. [Los Alamos National Laboratory (LANL); Ott, M. A. [Los Alamos National Laboratory (LANL); Rowton, L. J. [Los Alamos National Laboratory (LANL); Taylor, W. A. [Los Alamos National Laboratory (LANL); Wilhelmy, J. B. [Los Alamos National Laboratory (LANL); Yates, M. A. [Los Alamos National Laboratory (LANL); Carlson, A. D. [National Institute of Standards and Technology (NIST); Hill, Nathaniel [ORNL; Harper, R. [EG& G Energy Measurements, Los Alamos, NM; Hilko, R. [EG& G Energy Measurements, Los Alamos, NM

    1996-01-01

    Transmutation of actinide waste into fission products could be enhanced by using resonance fission of odd-odd target materials; those of interest are 232Pa, 238Np, and 242Am. Fission cross-section measurements of two of these short-lived materials were performed at Los Alamos National Laboratory. Samples were produced by the (d,2n) reaction in the Los Alamos Ion Beam Facility followed by fast radiochemistry to separate the odd-odd target of interest. The fission cross section of the nanogram samples was measured in a high intensity pulsed neutron beam produced by 800-MeV proton spallation. Using this procedure, the fission cross sections of the 1.3-day 232Pa and 2.1-day 238Np were successfully measured in the energy range from 0.01 eV to 50 keV. The fission cross section of the relatively long-life isotope 236Np was also measured in the same system while the short half-life isotopes were being prepared. The results and resonance analysis are presented.

  8. Assessment of pheromone production and response in fission yeast by a halo test of induced sporulation

    DEFF Research Database (Denmark)

    Egel, R; Willer, M; Kjaerulff, S

    1994-01-01

    We describe a rapid, sensitive and semi-quantitative plate assay for monitoring pheromone activity in the fission yeast Schizosaccharomyces pombe. It is based on the observation that meiosis requires stimulation by pheromone and exploits diploid strains that will only sporulate after addition...... of exogenous pheromone. The tester strains are heterozygous for mating type, are non-switching, and are mutated in one of the early subfunctions (either mat1-Mc or mat1-Pc), so that meiosis is only induced after exposure to exogenous pheromone (M-factor or P-factor, respectively). Pheromone activity...... is assessed as an iodine-positive halo of sporulation surrounding the pheromone source, and the width of the halo is related to the amount of pheromone being produced. The assay is sufficiently sensitive to monitor the low amount of M-factor produced by an M mam1 strain, and its sensitivity towards P...

  9. Measurement of fission products β decay properties using a total absorption spectrometer

    Directory of Open Access Journals (Sweden)

    Zakari-Issoufou A.-A.

    2013-12-01

    Full Text Available In a nuclear reactor, the β decay of fission fragments is at the origin of decay heat and antineutrino flux. These quantities are not well known while they are very important for reactor safety and for our understanding of neutrino physics. One reason for the discrepancies observed in the estimation of the decay heat and antineutrinos flux coming from reactors could be linked with the Pandemonium effect. New measurements have been performed at the JYFL facility of Jyväskylä with a Total Absorption Spectrometer (TAS in order to circumvent this effect. An overview of the TAS technique and first results from the 2009 measurement campaign will be presented.

  10. Delayed neutron spectra and their uncertainties in fission product summation calculations

    Energy Technology Data Exchange (ETDEWEB)

    Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)

    1997-03-01

    Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)

  11. Hydrothermal decomposition of actinide(IV oxalates: a new aqueous route towards reactive actinide oxide nanocrystals

    Directory of Open Access Journals (Sweden)

    Walter Olaf

    2016-01-01

    Full Text Available The hydrothermal decomposition of actinide(IV oxalates (An= Th, U, Pu at temperatures between 95 and 250 °C is shown to lead to the production of highly crystalline, reactive actinide oxide nanocrystals (NCs. This aqueous process proved to be quantitative, reproducible and fast (depending on temperature. The NCs obtained were characterised by X-ray diffraction and TEM showing their size to be smaller than 15 nm. Attempts to extend this general approach towards transition metal or lanthanide oxalates failed in the 95–250 °C temperature range. The hydrothermal decomposition of actinide oxalates is therefore a clean, flexible and powerful approach towards NCs of AnO2 with possible scale-up potential.

  12. Molecular Dynamics Simulation of Thermal Transport in UO2 Containing Uranium, Oxygen, and Fission-product Defects

    Science.gov (United States)

    Liu, X.-Y.; Cooper, M. W. D.; McClellan, K. J.; Lashley, J. C.; Byler, D. D.; Bell, B. D. C.; Grimes, R. W.; Stanek, C. R.; Andersson, D. A.

    2016-10-01

    Uranium dioxide (UO2 ) is the most commonly used fuel in light-water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, thereby governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models are replaced with models that incorporate explicit thermal-conductivity-degradation mechanisms during fuel burn up. This approach is able to represent the degradation of thermal conductivity due to each individual defect type, rather than the overall burn-up measure typically used, which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham-type interatomic potential and a potential that combines the many-body embedded-atom-method potential with Morse-Buckingham pair potentials. Potential parameters for UO2 +x and ZrO2 are developed for the latter potential. Physical insights from the resonant phonon-spin-scattering mechanism due to spins on the magnetic uranium ions are introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel-performance codes

  13. Fission yield covariances for JEFF: A Bayesian Monte Carlo method

    Directory of Open Access Journals (Sweden)

    Leray Olivier

    2017-01-01

    Full Text Available The JEFF library does not contain fission yield covariances, but simply best estimates and uncertainties. This situation is not unique as all libraries are facing this deficiency, firstly due to the lack of a defined format. An alternative approach is to provide a set of random fission yields, themselves reflecting covariance information. In this work, these random files are obtained combining the information from the JEFF library (fission yields and uncertainties and the theoretical knowledge from the GEF code. Examples of this method are presented for the main actinides together with their impacts on simple burn-up and decay heat calculations.

  14. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  15. Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products.

    Science.gov (United States)

    Košťál, Michal; Švadlenková, Marie; Milčák, Ján; Rypar, Vojtěch; Koleška, Michal

    2014-07-01

    The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well.

  16. Thermal transport in UO2 with defects and fission products by molecular dynamics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cooper, Michael William Donald [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lashley, Jason Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Byler, Darrin David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-14

    The importance of the thermal transport in nuclear fuel has motivated a wide range of experimental and modelling studies. In this report, the reduction of thermal transport in UO2 due to defects and fission products has been investigated using non-equilibrium MD simulations, with two sets of empirical potentials for studying the degregation of UO2 thermal conductivity including a Buckingham type interatomic potential and a recently developed EAM type interatomic potential. Additional parameters for U5+ and Zr4+ in UO2 have been developed for the EAM potential. The thermal conductivity results from MD simulations are then corrected for the spin-phonon scattering through Callaway model formulations. To validate the modelling results, comparison was made with experimental measurements on single crystal hyper-stoichiometric UO2+x samples.

  17. First results for fluid dynamics, neutronics and fission product behavior in HTR applying the HTR code package (HCP) prototype

    Energy Technology Data Exchange (ETDEWEB)

    Allelein, H.-J., E-mail: h.j.allelein@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH Aachen University, 52064 Aachen (Germany); Kasselmann, S.; Xhonneux, A.; Tantillo, F.; Trabadela, A.; Lambertz, D. [Forschungszentrum Jülich, 52425 Jülich (Germany)

    2016-09-15

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT lead to new cross sections, which are fed back into MGT-3D. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT-3D. Comparisons will be shown against data generated by SERPENT and the legacy codes VSOP99/11, NAKURE and FRESCO-II.

  18. Nanocrystalline SiC and Ti3SiC2 Alloys for Reactor Materials: Diffusion of Fission Product Surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H.; Jiang, Weilin

    2014-11-01

    MAX phases, such as titanium silicon carbide (Ti3SiC2), have a unique combination of both metallic and ceramic properties, which make them attractive for potential nuclear applications. Ti3SiC2 has been suggested in the literature as a possible fuel cladding material. Prior to the application, it is necessary to investigate diffusivities of fission products in the ternary compound at elevated temperatures. This study attempts to obtain relevant data and make an initial assessment for Ti3SiC2. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti3SiC2, SiC, and a dual-phase nanocomposite of Ti3SiC2/SiC synthesized at PNNL. Thermal annealing and in-situ Rutherford backscattering spectrometry (RBS) were employed to study the diffusivity of the various implanted species in the materials. In-situ RBS study of Ti3SiC2 implanted with Au ions at various temperatures was also performed. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti3SiC2 occurs during ion implantation at 873 K. Cs in Ti3SiC2 is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti3SiC2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Further studies of the related materials are recommended.

  19. GGA+U study of uranium mononitride: A comparison of the U-ramping and occupation matrix schemes and incorporation energies of fission products

    Energy Technology Data Exchange (ETDEWEB)

    Claisse, Antoine, E-mail: claisse@kth.se [KTH Royal Institute of Technology, Reactor Physics, AlbaNova University Centre, 106 91 Stockholm (Sweden); Klipfel, Marco [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Lindbom, Niclas [KTH Royal Institute of Technology, Reactor Physics, AlbaNova University Centre, 106 91 Stockholm (Sweden); Freyss, Michel [CEA, DEN, DEC, Centre de Cadarache, 13108 Saint-Paul-lez-Durance (France); Olsson, Pär [KTH Royal Institute of Technology, Reactor Physics, AlbaNova University Centre, 106 91 Stockholm (Sweden)

    2016-09-15

    Uranium mononitride is studied in the DFT + U framework. Its ground state is investigated and a study of the incorporation of diverse fission products in the crystal is conducted. The U-ramping and occupation matrix control (OMC) schemes are used to eliminate metastable states. Beyond a certain amount of introduced correlation, the OMC scheme starts to find a lower total energy. The OMC scheme is chosen for the second part of this study. Furthermore, the influence of the magnetic ordering is studied using the U-ramping method, showing that antiferromagnetic order is the most stable one when the U parameter is larger than 1.75 eV. The effect on the density of states is investigated and elastic constants are provided for comparison with other methods and experiments. The incorporation energies of fission products in different defect configurations are calculated and these energies are corrected to take into account the limited size of the supercell. - Highlights: • We studied bulk uranium nitride with means of DFT+U with the U-ramping scheme and the OMC scheme. • We produces a density of states plot and calculated the elastic constants of UN. • We calculated the incorporation energy of many fission products in UN, and corrected them to take into account the elastic interactions. • The OMC scheme should be used over the U-ramping scheme. • Fission products go to larger crystal sites.

  20. HTGR fuels and core development program. Quarterly progress report for the period ending November 30, 1976. [Graphite and fuel irradiations; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1976-12-27

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-12-01

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

  3. I-NERI ANNUAL TECHNICAL PROGRESS REPORT: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels

    Energy Technology Data Exchange (ETDEWEB)

    S. Frank

    2009-09-01

    An attractive alternative to the once-through disposal of electrorefiner salt is to selectively remove the active fission products from the salt and recycle the salt back to the electrorefiner (ER). This would allow salt reuse for some number of cycles before ultimate disposal of the salt in a ceramic waste form. Reuse of ER salt would, thus, greatly reduce the volume of ceramic waste produced during the pyroprocessing of spent nuclear fuel. This final portion of the joint I-NERI research project is to demonstrate the separation of fission products from molten ER salt by two methods previously selected during phase two (FY-08) of this project. The two methods selected were salt/zeolite contacting and rare-earth fission product precipitation by oxygen bubbling. The ER salt used in these tests came from the Mark-IV electrorefiner used to anodically dissolved driver fuel from the EBR-II reactor on the INL site. The tests were performed using the Hot Fuel Dissolution Apparatus (HFDA) located in the main cell of the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels complex on the INL site. Results from these tests were evaluated during a joint meeting of KAERI and INL investigators to provide recommendations as to the future direction of fission product removal from electrorefiner salt that accumulate during spent fuel treatment. Additionally, work continued on kinetic measurements of surrogate quaternary salt systems to provide fundamental kinetics on the ion exchange system and to expand the equilibrium model system developed during the first two phases of this project. The specific objectives of the FY09 I-NERI research activities at the INL include the following: • Perform demonstration tests of the selected KAERI precipitation and INL salt/zeolite contacting processes for fission product removal using radioactive, fission product loaded ER salt • Continue kinetic studies of the quaternary Cs/Sr-LiCl-KCl system to determine the rate of ion

  4. Demonstration of a TODGA based Extraction Process for the Partitioning of Minor Actinides from a PUREX Raffinate

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, D.; Christiansen, B.; Glatz, J.P.; Malmbeck, R.; Serrano-Purroy, D. [Commiss European Communities, Joint Res Ctr, Inst Transuranium Elements, D-76125 Karlsruhe, (Germany); Modolo, G. [Forschungszentrum Julich, Inst Energy Res Safety Res and Reactor Technol, D-52425 Julich, (Germany); Sorel, Ch. [Commissariat Energie Atom Valrho CEA, DRCP SCPS, F-30207 Bagnols Sur Ceze, (France); Magnusson, D. [Chalmers, Dept Chem and Biol Engn, S-41296 Gothenburg, (Sweden)

    2009-07-01

    Efficient recovery of minor actinides (MA) from genuine PUREX raffinate has been successfully demonstrated by the TODGA + TBP extractant mixture dissolved in an industrial aliphatic solvent TPH. The process was carried out in centrifugal contactors using an optimized flow-sheet involving a total of 32 stages, divided into 4 stages for extraction, 12 stages for scrubbing and 16 stages for back-extraction. Very high feed decontamination factors were obtained (Am, Cm 40 000) and the recovery of these elements was higher than 99.99%. Of the non-lanthanide fission products only Y and a small part of Ru were co-separated into the product fraction together with the lanthanides and the MA. (authors)

  5. Development of an integrated fission product release and transport code for spatially resolved full-core calculations of V/HTRs

    Energy Technology Data Exchange (ETDEWEB)

    Xhonneux, Andre, E-mail: a.xhonneux@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology RWTH-Aachen, 52064 Aachen (Germany); Allelein, Hans-Josef [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology RWTH-Aachen, 52064 Aachen (Germany)

    2014-05-01

    The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR

  6. A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Husain, A.

    2003-02-27

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.

  7. Enhanced trigger for the NIFFTE fissionTPC in presence of high-rate alpha backgrounds

    Science.gov (United States)

    Bundgaard, Jeremy; Niffte Collaboration

    2015-10-01

    Nuclear physics and nuclear energy communities call for new, high precision measurements to improve existing fission models and design next generation reactors. The Neutron Induced Fission Fragment Tracking experiment (NIFFTE) has developed the fission Time Projection Chamber (fissionTPC) to measure neutron induced fission with unrivaled precision. The fissionTPC is annually deployed to the Weapons Neutron Research facility at Los Alamos Neutron Science Center where it operates with a neutron beam passing axially through the drift volume, irradiating heavy actinide targets to induce fission. The fissionTPC was developed at the Lawrence Livermore National Laboratory's TPC lab, where it measures spontaneous fission from radioactive sources to characterize detector response, improve performance, and evolve the design. To measure 244Cm, we've developed a fission trigger to reduce the data rate from alpha tracks while maintaining a high fission detection efficiency. In beam, alphas from 239Pu are a large background when detecting fission fragments; implementing the fission trigger will greatly reduce this background. The implementation of the cathode fission trigger in the fissionTPC will be presented along with a detailed study of its efficiency.

  8. A Time Projection Chamber for High Accuracy and Precision Fission Cross-Section Measurements

    Energy Technology Data Exchange (ETDEWEB)

    T. Hill; K. Jewell; M. Heffner; D. Carter; M. Cunningham; V. Riot; J. Ruz; S. Sangiorgio; B. Seilhan; L. Snyder; D. M. Asner; S. Stave; G. Tatishvili; L. Wood; R. G. Baker; J. L. Klay; R. Kudo; S. Barrett; J. King; M. Leonard; W. Loveland; L. Yao; C. Brune; S. Grimes; N. Kornilov; T. N. Massey; J. Bundgaard; D. L. Duke; U. Greife; U. Hager; E. Burgett; J. Deaven; V. Kleinrath; C. McGrath; B. Wendt; N. Hertel; D. Isenhower; N. Pickle; H. Qu; S. Sharma; R. T. Thornton; D. Tovwell; R. S. Towell; S.

    2014-09-01

    The fission Time Projection Chamber (fissionTPC) is a compact (15 cm diameter) two-chamber MICROMEGAS TPC designed to make precision cross-section measurements of neutron-induced fission. The actinide targets are placed on the central cathode and irradiated with a neutron beam that passes axially through the TPC inducing fission in the target. The 4p acceptance for fission fragments and complete charged particle track reconstruction are powerful features of the fissionTPC which will be used to measure fission cross-sections and examine the associated systematic errors. This paper provides a detailed description of the design requirements, the design solutions, and the initial performance of the fissionTPC.

  9. Ternary fission

    Indian Academy of Sciences (India)

    M Balasubramaniam; K R Vijayaraghavan; C Karthikraj

    2015-09-01

    We present the ternary fission of 252Cf and 236U within a three-cluster model as well as in a level density approach. The competition between collinear and equatorial geometry is studied by calculating the ternary fragmentation potential as a function of the angle between the lines joining the stationary middle fragment and the two end fragments. The obtained results for the 16O accompanying ternary fission indicate that collinear configuration is preferred to equatorial configuration. Further, for all the possible third fragments, the potential energy surface (PES) is calculated corresponding to an arrangement in which the heaviest and the lightest fragments are considered at the end in a collinear configuration. The PES reveals several possible ternary modes including true ternary modes where the three fragments are of similar size. The complete mass distributions of Si and Ca which accompanied ternary fission of 236U is studied within a level density picture. The obtained results favour several possible ternary combinations.

  10. Modernizing the Fission Basis

    Science.gov (United States)

    Tonchev, Anton; Henderson, Roger; Schunck, Nicolas; Sroyer, Mark; Vogt, Ramona

    2016-09-01

    In 1939, Niels Bohr and John Wheeler formulated a theory of neutron-induced nuclear fission based on the hypothesis of the compound nucleus. Their theory, the so-called ``Bohr hypothesis,'' is still at the heart of every theoretical fission model today and states that the decay of a compound nucleus for a given excitation energy, spin, and parity is independent of its formation. We propose the first experiment to validate to 1-2% absolute uncertainties the practical consequences of the Bohr hypothesis during induced nuclear fission. We will compare the fission product yields (FPYs) of the same 240Pu compound nucleus produced via two different reactions (i) n+239Pu and (ii) γ+240 Pu. These high-precision FPYs measurements will be extremely beneficial for our fundamental understanding of the nuclear fission process and nuclear reactions from first principles. This work was performed under the auspices of US DOE by LLNL under Contract DE-AC52-07NA27344. Funding was provided via the LDRD-ERD-069 project.

  11. Radiochemistry and the Study of Fission

    Energy Technology Data Exchange (ETDEWEB)

    Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-14

    These are slides from a lecture given at UC Berkeley. Radiochemistry has been used to study fission since it’ discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution. The following topics are covered: In the beginning: the discovery of fission; forensics using fission products: what can be learned from fission products, definitions of R-values and Q-values, fission bases, K-factors and fission chambers, limitations; the neutron energy dependence of the mass yield distribution (the two mode fission hypothesis); the influence of nuclear structure on the mass yield distribution. In summary: Radiochemistry has been used to study fission since it’s discovery. Radiochemical measurement of fission product yields have provided the highest precision data for developing fission models and for nuclear forensics. The two-mode fission hypothesis provides a description of the neutron energy dependence of the mass yield curve. However, data is still rather sparse and more work is needed near second and third chance fission. Radiochemical measurements have provided evidence for the importance of nuclear states in the compound nucleus in predicting the mass yield curve in the resonance region.

  12. Development of ionization technique for measurement of fast neutron induced fission products yields of {sup 237}Np

    Energy Technology Data Exchange (ETDEWEB)

    Goverdovski, A.A.; Khryachkov, V.A.; Ketlerov, V.V.; Mitrofanov, V.F.; Ostapenko, Yu.B.; Semenova, N.N.; Fomichev, A.N.; Rodina, L.F. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-03-01

    Twin gridded ionization chamber and corresponding software was designed for measurements of masses, kinetic energies and nuclear charges of fission fragments from fast neutron induced fission of {sup 237}Np. The ionization detector design, electronics, data acquisition and processing system and the test results are presented in this paper. (J.P.N.)

  13. Development of ionization technique for measurement of fast neutron induced fission products yields of {sup 237}Np

    Energy Technology Data Exchange (ETDEWEB)

    Goverdovski, A.A.; Khryachkov, V.A.; Ketlerov, V.V.; Mitrofanov, V.F.; Ostapenko, Yu.B.; Semenova, N.N.; Fomichev, A.N.; Rodina, L.F. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-03-01

    Twin gridded ionization chamber and corresponding software was designed for measurements of masses, kinetic energies and nuclear charges of fission fragments from fast neutron induced fission of {sup 237}Np. The ionization detector design, electronics, data acquisition and processing system and the test results are presented in this paper. (J.P.N.)

  14. Spallation and fission products in the (p+{sup 179}Hf) and (p+{sup nat}Hf) reactions

    Energy Technology Data Exchange (ETDEWEB)

    Karamian, S.A. [Joint Institute for Nuclear Research, Moscow Region, Dubna 141980 (Russian Federation); Ur, C.A. [Horia Hulubei National Institute of Physics and Nuclear Engineering, P.O. Box MG-6, Bucharest 077125 (Romania); Induced Gamma Emission Foundation, P.O. Box 34-81, Bucharest 010044 (Romania)], E-mail: ur@pd.infn.it; Adam, J. [Joint Institute for Nuclear Research, Moscow Region, Dubna 141980 (Russian Federation); Institute of Nuclear Physics, Rez, Prague CZ-25068 (Czech Republic); Kalinnikov, V.G.; Lebedev, N.A.; Vostokin, G.K. [Joint Institute for Nuclear Research, Moscow Region, Dubna 141980 (Russian Federation); Collins, C.B. [Center for Quantum Electronics, University of Texas at Dallas, Richardson, TX 75083 (United States); Popescu, I.I. [Induced Gamma Emission Foundation, P.O. Box 34-81, Bucharest 010044 (Romania)

    2009-03-01

    Production of Hf and Lu high-spin isomers has been experimentally studied in spallation reactions induced by intermediate energy protons. Targets of enriched {sup 179}Hf (91%) and {sup nat}Hf were bombarded with protons of energy in the range from 90 to 650 MeV provided by the internal beam of the Dubna Phasotron synchrocyclotron. The activation yields of the reaction products were measured by using the {gamma}-ray spectroscopy and radiochemistry methods. The production cross-sections obtained for the {sup 179m2}Hf, {sup 178m2}Hf and {sup 177m}Lu isomers are similar to the previously measured values from the spallation of Ta, Re and W targets. Therefore, the reactions involving emission of only a few nucleons, like (p,p'), (p,p'n) and (p,2pn), can transfer high enough angular momentum to the final residual nuclei with reasonable large cross-sections. A significant gain in the isomeric yields was obtained when enriched {sup 179}Hf targets were used. The mass distribution of the residual nuclei was measured over a wide range of masses and the fission-to-spallation ratio could be deduced as a function of the projectile energy. Features of the reaction mechanism are briefly discussed.

  15. Ionizing radiation accelerates Drp1-dependent mitochondrial fission, which involves delayed mitochondrial reactive oxygen species production in normal human fibroblast-like cells

    Energy Technology Data Exchange (ETDEWEB)

    Kobashigawa, Shinko, E-mail: kobashin@nagasaki-u.ac.jp [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan); Suzuki, Keiji; Yamashita, Shunichi [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan)

    2011-11-04

    Highlights: Black-Right-Pointing-Pointer We report first time that ionizing radiation induces mitochondrial dynamic changes. Black-Right-Pointing-Pointer Radiation-induced mitochondrial fission was caused by Drp1 localization. Black-Right-Pointing-Pointer We found that radiation causes delayed ROS from mitochondria. Black-Right-Pointing-Pointer Down regulation of Drp1 rescued mitochondrial dysfunction after radiation exposure. -- Abstract: Ionizing radiation is known to increase intracellular level of reactive oxygen species (ROS) through mitochondrial dysfunction. Although it has been as a basis of radiation-induced genetic instability, the mechanism involving mitochondrial dysfunction remains unclear. Here we studied the dynamics of mitochondrial structure in normal human fibroblast like cells exposed to ionizing radiation. Delayed mitochondrial O{sub 2}{sup {center_dot}-} production was peaked 3 days after irradiation, which was coupled with accelerated mitochondrial fission. We found that radiation exposure accumulated dynamin-related protein 1 (Drp1) to mitochondria. Knocking down of Drp1 expression prevented radiation induced acceleration of mitochondrial fission. Furthermore, knockdown of Drp1 significantly suppressed delayed production of mitochondrial O{sub 2}{sup {center_dot}-}. Since the loss of mitochondrial membrane potential, which was induced by radiation was prevented in cells knocking down of Drp1 expression, indicating that the excessive mitochondrial fission was involved in delayed mitochondrial dysfunction after irradiation.

  16. Predictions of Actinide Solubilities under Near-Field Conditions Expected in the WIPP