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Development of reference materials for SNF NDA systems

Description: The Department of Energy has over 200 different fuel types which will be placed in a geologic repository for ultimate disposal. At the present time, DOE EM is responsible for assuring safe existing conditions, achieving interim storage, and preparing for final disposition. Each task is governed by regulations which dictate a certain degree of knowledge regarding the contents and condition of the fuel. This knowledge and other associated characteristics are referred to as data needs. It is the stance of DOE EM, that personnel and economic resources are not available to obtain the necessary data to characterize such individual fuel type for final disposal documentation purposes. In addition, it is beyond the need of DOE to do so. This report describes the effort to classify the 200+ fuel types into a subset of fuel types for the purpose of non-destructive analysis (NDA) measurement system development and demonstration testing in support of the DOE National Spent Nuclear Fuel (NSNFP) Program. The fuel types have been grouped into 37 groups based on fuel composition, fuel form, assembly size, enrichment, and other characteristics which affect NDA measurements (e.g., neutron poisons).
Date: February 29, 2000
Creator: Klann, R. T.
Partner: UNT Libraries Government Documents Department

Fast neutron (14.5 MeV) radiography: a comparative study

Description: Fast neutron (14.5 MeV) radiography is a type of non-destructive analysis tool that offers its own benefits and drawbacks. Because cross-sections vary with energy, a different range of materials can be examined with fast neutrons than can be studied with thermal neutrons, epithermal neutrons, or x-rays. This paper details these differences through a comparative study of fast neutron radiography to the other types of radiography available. The most obvious difference among the different types of radiography is in the penetrability of the sources. Fast neutrons can probe much deeper and can therefore obtain details of the internals of thick objects. Good images have been obtained through as much as 15 cm of steel, 10 cm of water, and 15 cm of borated polyethylene. In addition, some objects were identifiable through as much as 25 cm of water or 30 cm of borated polyethylene. The most notable benefit of fast neutron radiography is in the types of materials that can be tested. Fast neutron radiography can view through materials that simply cannot be viewed by X rays, thermal neutrons, or epithermal neutrons due to the high cross-sections or linear attenuation coefficients involved. Cadmium was totally transparent to the fast neutron source. Fast neutron radiography is not without drawbacks. The most pronounced drawback has been in the quality of radiograph produced. The image resolution is only about 0.8 mm for a 1.25 cm thick object, whereas, other forms of radiography have much better resolution.
Date: July 1, 1996
Creator: Klann, R.T.
Partner: UNT Libraries Government Documents Department

A system for fast neutron radiography

Description: A system has been designed and a neutron generator installed to perform fast neutron radiography. With this sytem, objects as small as a coin or as large as a waste drum can be radiographed. The neutron source is an MF Physics A-711 neutron generator which produces 3x10{sup 10} neutrons/second with an average energy of 14.5 MeV. The radiography system uses x-ray scintillation screens and film in commercially available cassettes. The cassettes have been modified to include a thin sheet of plastic to convert neutrons to protons through elastic scattering from hydrogen and other low Z materials in the plastic. For film densities from 1.8 to 3.0, exposures range from 1.9x10{sup 7} to 3.8x10{sup 8} n/cm{sup 2} depending on the type of screen and film.
Date: May 1, 1996
Creator: Klann, R.T.
Partner: UNT Libraries Government Documents Department

Determining Pu-239 content by resonance transmission analysis using a filtered reactor beam.

Description: A novel technique has been developed at Argonne National Laboratory to determine the {sup 239}Pu content in EBR-II blanket elements using resonance transmission analysis (RTA) with a filtered reactor beam. The technique uses cadmium and gadolinium filters along with a {sup 239}Pu fission chamber to isolate the 0.3 eV resonance in {sup 239}Pu. In the energy range from 0.1 to 0.5 eV, the total microscopic cross-section of {sup 239}Pu is significantly larger than the cross-sections of {sup 238}U and {sup 235}U. This large difference in cross-section allows small amounts of {sup 239}Pu to be detected in uranium samples. Tests using a direct beam from a 250 kW TRIGA reactor have been performed with stacks of depleted uranium and {sup 239}Pu foils. Preliminary measurement results are in good agreement with the predicted results up to about two weight percent of {sup 239}Pu in the sample. In addition, measured {sup 239}Pu masses were in agreement with actual sample masses with uncertainties less than 3.8 percent.
Date: November 25, 1998
Creator: Klann, R. T.
Partner: UNT Libraries Government Documents Department

Fast neutron radiography research at ANL-W

Description: Thirty-seven different elements were tested for their suitability as converter screens for direct and indirect fast neutron radiography. The use of commercial X-ray scintillator screens containing YTaO{sub 4}, LaOBr:Tm, YTaO{sub 4}:Nb, YTaO{sub 4}:Tm, CaWO{sub 4}, BaSO{sub 4}:Sr, and GdO{sub 2}S:Tb was also explored for direct fast neutron radiography. For the indirect radiographic process, only one element, holmium, was found to be better than copper. Iron was also found to work as well as copper. All other elements that were tested were inferior to copper for indirect fast neutron radiography. For direct fast neutron radiography, the results were markedly different. Copper was found to be a poor material to sue, as thirty-two of the elements performed better than the copper. Tantalum was found to be the best material to use. Several other materials that also performed remarkably well include, in order of decreasing utility, gold, lutetium, germanium, dysprosium, and thulium. Several interesting results were obtained for the commercial X-ray scintillator screens. Most notably, useful radiographs were produced with all of the various scintillation screens. However, the screens containing YTaO{sub 4}:Nb offered the greatest film densities for the shortest exposure times. Screens using GdSO{sub 4}:Tb provided the best resolution and clearest images at the sacrifice of exposure time. Also, as previous researchers found, scintillator screens offered significantly shorter exposure times than activation foils.
Date: June 1, 1996
Creator: Klann, R.T. & Natale, M.D.
Partner: UNT Libraries Government Documents Department

A system for fast neutron radiography

Description: A system has been designed and a neutron generator installed to perform fast neutron radiography. With this system, objects as small as a coin and as large as a 19 liter container have been radiographed. The neutron source is an MF Physics A-711 neutron generator which produces 3 x 10[sup 10] neutrons/second with an average energy of 14. 5 MeV. The radiography system uses x-ray scintillation screens and film in commercially available light-tight cassettes. The cassettes have been modified to include a thin sheet of plastic to produce protons from the neutron beam through elastic scattering from hydrogen and other low Z materials in the plastic. For film densities from 1.8 to 3.0, exposures range from 1.9 x 10[sup 7] n/cm[sup 2] to 3.8 x 10[sup 8] n/cm[sup 2] depending on the type of screen and film. The optimum source-to-film distance was found to be 150 cm. At this distance, the geometric unsharpness was determined to be approximately 2.2-2.3 mm and the smallest hole that could be resolved in a 1.25 cm thick sample had a diameter of 0.079 cm.
Date: April 1, 1997
Creator: Klann, R.T.
Partner: UNT Libraries Government Documents Department

Non-destructive assay of EBR-II blanket elements using resonance transmission analysis.

Description: Resonance transmission analysis utilizing a faltered reactor beam was examined as a means of determining the {sup 239}Pu content in Experimental Breeder Reactor-II depleted uranium blanket elements. The technique uses cadmium and gadolinium falters along with a {sup 239}Pu fission chamber to isolate the 0.3 eV resonance in {sup 239}Pu. In the energy range of this resonance (0.1 eV to 0.5 ev), the total microscopic cross-section of {sup 239}Pu is significantly greater than the cross-sections of {sup 238}U and {sup 235}U. This large difference allows small changes in the {sup 239}Pu content of a sample to result in large changes in the mass signal response. Tests with small stacks of depleted uranium and {sup 239}Pu foils indicate a significant change in response based on the {sup 239}Pu content of the foil stack. In addition, the tests indicate good agreement between the measured and predicted values of {sup 239}Pu up to approximately two weight percent.
Date: September 11, 1998
Creator: Klann, R.T. & Poenitz, W.P.
Partner: UNT Libraries Government Documents Department

Depletion calculations for the McClellan Nuclear Radiation Center.

Description: Depletion calculations have been performed for the McClellan reactor history from January 1990 through August 1996. A database has been generated for continuing use by operations personnel which contains the isotopic inventory for all fuel elements and fuel-followed control rods maintained at McClellan. The calculations are based on the three-dimensional diffusion theory code REBUS-3 which is available through the Radiation Safety Information Computational Center (RSICC). Burnup-dependent cross-sections were developed at zero power temperatures and full power temperatures using the WIMS code (also available through RSICC). WIMS is based on discretized transport theory to calculate the neutron flux as a function of energy and position in a one-dimensional cell. Based on the initial depletion calculations, a method was developed to allow operations personnel to perform depletion calculations and update the database with a minimal amount of effort. Depletion estimates and calculations can be performed by simply entering the core loading configuration, the position of the control rods at the start and end of cycle, the reactor power level, the duration of the reactor cycle, and the time since the last reactor cycle. The depletion and buildup of isotopes of interest (heavy metal isotopes, erbium isotopes, and fission product poisons) are calculated for all fuel elements and fuel-followed control rods in the MNRC inventory. The reactivity loss from burnup and buildup of fission product poisons and the peak xenon buildup after shutdown are also calculated. The reactivity loss from going from cold zero power to hot full power can also be calculated by using the temperature-dependent, burnup-dependent cross-sections. By calculating all of these reactivity effects, operations personnel are able to estimate the total excess reactivity necessary to run the reactor for the given cycle. This method has also been used to estimate the worth of individual control rods. Using this approach, fuel ...
Date: December 8, 1997
Creator: Klann, R. T. & Newell, D. L.
Partner: UNT Libraries Government Documents Department

Gamma-ray Sptectrometric Characterization of Overpacked CC 104/107 RH-TRU Wastes: Surrogate Tests

Description: Development of the gamma-ray spectrometric technique termed GSAK (Gamma-Ray Spectrometry with Acceptable Knowledge) for the characterization of CC104/107 remote-handled transuranic (RH-TRU) wastes continued this year. Proof-of-principle measurements have been completed on the surrogate RH-TRU waste drums configured earlier this year. The GSAK technique uses conventional gamma-ray spectrometry to quantify the detectable fission product content of overpacked RH-TRU drums. These results are then coupled with the inventory report to characterize the waste drum content. The inventory report is based on process knowledge of the waste drum loading and calculations of the isotopic distribution in the spent fuel examined to generate the drummed wastes. Three RH-TRU surrogate drums were configured with encapsulated EBR-II driver fuel rod segments arranged in the surrogate drum assemblies. Segment-specific inventory calculations initially specified the radionuclide content of the fuel segments and thus the surrogate drums. Radiochemical assays performed on representative fuel element segments identified a problem in the accuracy of some of the fission and activation product inventory values and provided a basis for adjustment of the specified surrogate drum inventories. The three waste drum surrogates, contained within their 8.9 cm (3.5 inch) thick steel overpacks, were analyzed by gamma-ray spectrometry at the TREAT facility at Argonne National Laboratory-West. Seven fission and activation product radionuclides (54Mn, 60Co, 125Sb, 134Cs, 137Cs, 144CePr, and 154Eu) were reliably detected. The gamma-ray spectral accuracy was very good. In all cases, a two-sigma error bar constructed about the measured value included the actual drum activity.
Date: April 1, 2000
Creator: Hartwell, John Kelvin; Mc Ilwain, Michael Edward & Klann, R. T.
Partner: UNT Libraries Government Documents Department

Coated gallium arsenide neutron detectors : results of characterizationmeasurements.

Description: Effective detection of special nuclear materials (SNM) is essential for reducing the threat associated with stolen or improvised nuclear devices. Passive radiation detection technologies are primarily based on gamma-ray detection and subsequent isotope identification or neutron detection (specific to neutron sources and SNM). One major effort supported by the Department of Homeland Security in the area of advanced passive detection is handheld or portable neutron detectors for search and localization tasks in emergency response and interdiction settings. A successful SNM search detector will not only be able to confirm the presence of fissionable materials but also establish the location of the source in as short of time as possible while trying to minimize false alarms due to varying background or naturally occurring radioactive materials (NORM). For instruments based on neutron detectors, this translates to detecting neutrons from spontaneous fission or alpha-n reactions and being able to determine the direction of the source (or localizing the source through subsequent measurements). Polyethylene-coated gallium arsenide detectors were studied because the detection scheme is based on measuring the signal in the gallium arsenide wafers from the electrical charge of the recoil protons produced from the scattering of neutrons from the hydrogen nucleus. The inherent reaction has a directional dependence because the neutron and hydrogen nucleus have equivalent masses. The assessment and measurement of polyethylene-coated gallium arsenide detector properties and characteristics was the first phase of a project being performed for the Department of Homeland Security and the results of these tests are reported in this report. The ultimate goal of the project was to develop a man-portable neutron detection system that has the ability to determine the direction of the source from the detector. The efficiency of GaAs detectors for different sizes of polyethylene layers and different angles between the detector and the neutron ...
Date: September 29, 2006
Creator: Klann, R. T.; Perret, G. & Sanders, J.
Partner: UNT Libraries Government Documents Department

OSMOSE experiment representativity studies.

Description: The OSMOSE program aims at improving the neutronic predictions of advanced nuclear fuels through measurements in the MINERVE facility at the CEA-Cadarache (France) on samples containing the following separated actinides: Th-232, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, Cm-244 and Cm-245. The goal of the experimental measurements is to produce a database of reactivity-worth measurements in different neutron spectra for the separated heavy nuclides. This database can then be used as a benchmark for integral reactivity-worth measurements to verify and validate reactor analysis codes and integral cross-section values for the isotopes tested. In particular, the OSMOSE experimental program will produce very accurate sample reactivity-worth measurements for a series of actinides in various spectra, from very thermalized to very fast. The objective of the analytical program is to make use of the experimental data to establish deficiencies in the basic nuclear data libraries, identify their origins, and provide guidelines for nuclear data improvements in coordination with international programs. To achieve the proposed goals, seven different neutron spectra can be created in the MINERVE facility: UO2 dissolved in water (representative of over-moderated LWR systems), UO2 matrix in water (representative of LWRs), a mixed oxide fuel matrix, two thermal spectra containing large epithermal components (representative of under-moderated reactors), a moderated fast spectrum (representative of fast reactors which have some slowing down in moderators such as lead-bismuth or sodium), and a very hard spectrum (representative of fast reactors with little moderation from reactor coolant). The different spectra are achieved by changing the experimental lattice within the MINERVE reactor. The experimental lattice is the replaceable central part of MINERVE, which establishes the spectrum at the sample location. This configuration leads to a uniform well-behaved system so that the reactor configuration is in the fundamental mode. In fact, an important ...
Date: October 10, 2007
Creator: Aliberti, G. & Klann, R.
Partner: UNT Libraries Government Documents Department

ANL pre analysis of the SHEBA/CERES experiments.

Description: The French and British nuclear programs have prepared a series of natural uranium oxide fuel samples spiked with small amounts of the individual fission products which makeup a large fraction of the total neutron absorption by fission products in spent nuclear fuel. Both programs have utilized these samples in experimental reactors and have inferred the worth of the individual fission products. These results have been used to validate the cross sections used in criticality safety calculations. These measurements constitute a major element in support of spent fuel burnup credit in these countries.
Date: May 5, 2000
Creator: Palmiotti, G.; Smith, M.; Klann, R.; Fujita, E. & Imel, G.
Partner: UNT Libraries Government Documents Department

Criticality safety issues in the disposition of BN-350 spent fuel

Description: A criticality safety analysis has been performed as part of the BN-350 spent fuel disposition project being conducted jointly by the DOE and Kazakhstan. The Kazakhstan regulations are reasonably consistent with those of the DOE. The high enrichment and severe undermoderation of this fast reactor fuel has significant criticality safety consequences. A detailed modeling approach was used that showed some configurations to be safe that otherwise would be rejected. Reasonable requirements for design and operations were needed, and with them, all operations were found to be safe.
Date: February 28, 2000
Creator: Schaefer, R. W.; Klann, R. T.; Koltyshev, S. M. & Krechetov, S.
Partner: UNT Libraries Government Documents Department

Gamma-Ray Spectrometric Characterization of Overpacked CC104/107 RH-TRU Wastes: Surrogate Tests

Description: Development of the gamma-ray spectrometric technique termed GSAK (Gamma-Ray Spectrometry with Acceptable Knowledge) for the characterization of CC104/107 remote-handled transuranic (RH-TRU) wastes continued this year. Proof-of-principle measurements have been completed on the surrogate RH-TRU waste drums configured earlier this year. The GSAK technique uses conventional gamma-ray spectrometry to quantify the detectable fission product content of overpacked RH-TRU drums. These results are then coupled with the inventory report to characterize the waste drum content. The inventory report is based on process knowledge of the waste drum loading and calculations of the isotopic distribution in the spent fuel examined to generate the drummed wastes. Three RH-TRU surrogate drums were configured with encapsulated EBR-II driver fuel rod segments arranged in the surrogate drum assemblies. Segment-specific inventory calculations initially specified the radionuclide content of the fuel segments and thus the surrogate drums. Radiochemical assays performed on representative fuel element segments identified a problem in the accuracy of some of the fission and activation product inventory values and provided a basis for adjustment of the specified surrogate drum inventories. The three waste drum surrogates, contained within their 8.9 cm (3.5 inch) thick steel overpacks, were analyzed by gamma-ray spectrometry at the TREAT facility at Argonne National Laboratory-West. Seven fission and activation product radionuclides ({sup 54}Mn, {sup 60}Co, {sup 125}Sb, {sup 134}Cs, {sup 137}Cs, {sup 144}CePr, and {sup 154}Eu) were reliably detected. The gamma-ray spectral accuracy was very good. In all cases, a two-sigma error bar constructed about the measured value included the actual drum activity.
Date: May 1, 2000
Creator: Hartwell, J. K.; Klann, R. T. & McIlwain, M. E.
Partner: UNT Libraries Government Documents Department

Modeling report of the CEA cadarache MINERVE reactor for the OSMOSE project.

Description: The OSMOSE program (Oscillation in Minerve of isotopes in ''Eupraxic'' spectra) is a collaboration between the U.S. Department of Energy (DOE) and the Commissariat a l' Energie Atomique (CEA). It aims at measuring integral absorption rates of minor actinides by the oscillation technique in the MINERVE experimental facility located at the CEA Cadarache Research Center. The OSMOSE program also includes a complete analytical program to understand and resolve potential discrepancies between calculated and measured values. The OSMOSE program began in 2001 and will continue until 2013. The Argonne National Laboratory has developed Monte Carlo and deterministic calculation models of the MINERVE facility to determine core and safety parameters such as axial and radial fission rate distributions, control rod worth, spectral indices, and the reactivity worth of oscillated samples. Oscillation samples include calibration samples with different uranium enrichments and boron concentrations and the OSMOSE samples--separated actinides including {sup 232}Th, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm and {sup 245}Cm. Seven different neutron spectra will be created in the MINERVE facility: an overmoderated UO{sub 2} matrix (representative of a fuel processing plant or flooded storage cask), a UO{sub 2} matrix in water (representative of LWRs), a mixed oxide fuel matrix (representative of cores containing MOX fuels), two epithermal spectra (representative of under-moderated reactors), a moderated fast spectrum (representative of fast reactors which have some slowing down due to moderators such as lead-bismuth or sodium), and a very hard spectrum (representative of fast reactors with little moderation from reactor coolant). The different spectra are achieved by changing the experimental lattice within the MINERVE reactor. The currently investigated core configurations are R1UO2 and R1MOX, representative of a LWR loaded with UO{sub 2} and ...
Date: February 25, 2005
Creator: Klann, R.; Perret, G.; Hudelot, J. P. & Antony, M.
Partner: UNT Libraries Government Documents Department

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

Description: An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.
Date: August 3, 2007
Creator: Zhong, Z. & Klann, R. T.
Partner: UNT Libraries Government Documents Department

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-UO2 configuration.

Description: An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R1-UO2 core configuration were completed. The reactor model was generated using the REBUS code developed at Argonne National Laboratory. The calculations are based on the specifications for fabrication, so they are considered preliminary until sampling and analysis have been completed on the fabricated samples. The estimates indicate a range of reactivity effect from -22 pcm to +25 pcm compared to the natural U sample.
Date: October 3, 2007
Creator: Klann, R. T. & Perret, G.
Partner: UNT Libraries Government Documents Department

OSMOSE program : statistical review of oscillation measurements in the MINERVE reactor R1-UO2 configuration.

Description: The OSMOSE program is a collaboration on reactor physics experiments between the United States Department of Energy and the France Commissariat Energie Atomique. At the working level, it is a collaborative effort between the Argonne National Laboratory and the CEA Cadarache Research Center. The objective of this program is to measure very accurate integral reaction rates in representative spectra for the actinides important to future nuclear system designs, and to provide the experimental data for improving the basic nuclear data files. The main outcome of the OSMOSE measurement program will be an experimental database of reactivity-worth measurements in different neutron spectra for the heavy nuclides. This database can then be used as a benchmark to verify and validate reactor analysis codes. The OSMOSE program (Oscillation in Minerve of isotopes in Eupraxic Spectra) aims at improving neutronic predictions of advanced nuclear fuels through oscillation measurements in the MINERVE facility on samples containing the following separated actinides: {sup 232}Th, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm, and {sup 245}Cm. The first part of this report provides an overview of the experimental protocol and the typical processing of a series of experimental results which is currently performed at CEA-Cadarache. In the second part of the report, improvements to this technique are presented, as well as the program that was created to process oscillation measurement results from the MINERVE facility in the future.
Date: August 28, 2007
Creator: Stoven, G.; Klann, R. & Zhong, Z.
Partner: UNT Libraries Government Documents Department

Final report of the International Nuclear Energy Research Initiative OSMOSE project (FY01-FY04).

Description: The need for better nuclear data has been stressed by various organizations throughout the world, and results of studies have been published which demonstrate that current data are inadequate for designing the projects under consideration [1] [2]. In particular, a Working Party of the OECD has been concerned with identifying these needs [3] and has produced a detailed High Priority Request List for Nuclear Data. The French Atomic Energy Commissariat (CEA) has also recognized the need for better data and launched an ambitious program aimed at measuring the integral absorption rate parameters at the CEA-Cadarache Research Center. A complete analytical program is associated with the experimental program and aims at understanding and resolving potential discrepancies between calculated and measured values. The final objective of the program is to reduce the uncertainties in predictive capabilities to a level acceptable to core designers and government regulators. Argonne National Laboratory has expertise in these areas. In the past, ANL teams have developed very accurate experimental techniques and strongly enhanced the development of several French experimental and analytical programs, and have contributed to the computational tools used at CEA-Cadarache. CEA recognized the expertise that ANL has in these areas and was interested in collaborating with ANL in the experimental design, measurements, and analysis tasks of the OSMOSE (Oscillation in Minerve of Isotopes in Eupraxic Spectra) program. The development and execution of the first phase of the OSMOSE program within the DOE I-NERI Program was a resounding success. Both parties saw improved performance in the conduct of the program because of the contribution from both parties. The collaboration included several key aspects: (1) DOE supplied specific minor actinide isotopes to CEA that were not easily obtainable in France, (2) ANL staff participated and supported the experimental program, (3) ANL and CEA personnel performed analysis for ...
Date: February 25, 2005
Creator: Klann, R. T.; Perret, G.; Hudelot, J. P.; Drin, N.; Lee, J. & Cao, Y.
Partner: UNT Libraries Government Documents Department

Annual report of the international nuclear energy research initiative OSMOSE project (FY06).

Description: The goal of the OSMOSE program is to measure the reactivity effect of minor actinides in known neutron spectra of interest to the Generation-IV reactor program and other programs and to create a database of these results for use as an international benchmark for the minor actinides. The results are then compared to calculation models to verify and validate integral absorption cross-sections for the minor actinides. The OSMOSE program includes all aspects of the experimental program--including the fabrication of fuel pellets and samples, the oscillation of the samples in the MINERVE reactor for the measurement of the reactivity effect, reactor physics modeling of the MINERVE reactor, and the data analysis and interpretation of the experimental results.
Date: August 29, 2007
Creator: Klann, R. T.; Hudelot, J. P.; Drin, N.; Zhong, Z.; Division, Nuclear Engineering & Atomique, Commissariat a l Energie
Partner: UNT Libraries Government Documents Department

Annual report of the international nuclear research initiative OSMOSE project (FY05).

Description: The goal of the OSMOSE program is to measure the reactivity effect of minor actinides in known neutron spectra of interest to the Generation-IV reactor program and other programs and to create a database of these results for use as an international benchmark for the minor actinides. The results are then compared to calculational models to verify and validate integral absorption cross-sections for the minor actinides. The OSMOSE program includes all aspects of the experimental program -- including the fabrication of fuel pellets and samples, the oscillation of the samples in the MINERVE reactor for the measurement of the reactivity effect, reactor physics modeling of the MINERVE reactor, and the data analysis and interpretation of the experimental results.
Date: October 3, 2007
Creator: Klann, R. T.; Hudelot, J. P.; Perret, G.; Drin, N.; Division, Nuclear Engineering & Atomique, Commissariat a l'Energie
Partner: UNT Libraries Government Documents Department

OSMOSE an experimental program for improving neutronic predictions of advanced nuclear fuels.

Description: This report describes the technical results of tasks and activities conducted in FY07 to support the DOE-CEA collaboration on the OSMOSE program. The activities are divided into five high-level tasks: reactor modeling and pre-experiment analysis, sample fabrication and analysis, reactor experiments, data treatment and analysis, and assessment for relevance to high priority advanced reactor programs (such as GNEP and Gen-IV).
Date: October 18, 2007
Creator: Klann, R. T.; Aliberti, G.; Zhong, Z.; Graczyk, D.; Loussi, A.; Division, Nuclear Engineering et al.
Partner: UNT Libraries Government Documents Department