Short-lived radionuclides produced on the ORNL 86-inch cyclotron and High-Flux Isotope Reactor

International Nuclear Information System (INIS)

Lamb, E.

1985-01-01

The production of short-lived radionuclides at ORNL includes the preparation of target materials, irradiation on the 86-in. cyclotron and in the High Flux Isotope Reactor (HFIR), and chemical processing to recover and purify the product radionuclides. In some cases the target materials are highly enriched stable isotopes separated on the ORNL calutrons. High-purity 123 I has been produced on the 86-in. cyclotron by irradiating an enriched target of 123 Te in a proton beam. Research on calutron separations has led to a 123 Te product with lower concentrations of 124 Te and 126 Te and, consequently to lower concentrations of the unwanted radionuclides, 124 I and 126 I, in the 123 I product. The 86-in. cyclotron accelerates a beam of protons only but is unique in providing the highest available beam current of 1500 μA at 21 MeV. This beam current produces relatively large quantities of radionuclides such as 123 I and 67 Ga

Isotopic characterization of targets for nuclear measurements at CBNM

International Nuclear Information System (INIS)

Bievre, P. de

1985-01-01

Nuclear measurements for which ''nuclear'' targets are prepared are almost always isotope-specific i.e. they are normally related to a particular nuclide in the target. The amount of this nuclide must be accurately assessed. There are essentially two ways to determine the number of atoms of this particular nuclide. (1) By determination of the amount of element, to which the nuclide belongs, on the target via classsical means; weighing substraction of impurities, calculation of element amount using known of the chemical compound in which the element is incorporated and, finally, measurement of the isotopic composition in order to determine the fraction of the nuclide concerned in the element. An alternative way may be to perform an elemental assay on the target followed by determination of the isotopic composition. (2) Another approach is isotope dilution mass spectrometry where a change in the isotopic composition of the ''target'' is induced by adding a known number of atoms (called ''spike'') of the element with a quite different composition. Measurement of the resulting change in isotopic composition yields directly the number of atoms of the nuclide under investigation. The method is highly selective, accurate and isotope-specific. (orig.)

Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

Energy Technology Data Exchange (ETDEWEB)

Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Betzler, Ben [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Hirtz, Gregory John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Sunny, Eva [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-09-01

The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length ²⁵²Cf production targets, ⁷⁵Se production capsules, ⁶³Ni production capsules, a ¹⁸⁸W production capsule, and various materials irradiation targets. Fully loaded ²³⁸Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

Characteristics of isotope-selective chemical reactor with gas-separating device

International Nuclear Information System (INIS)

Gorshunov, N.M.; Kalitin, S.A.; Laguntsov, N.I.; Neshchimenko, Yu.P.; Sulaberidze, G.A.

1988-01-01

A study was made on characteristics of separating stage, composed of isotope-selective chemical (or photochemical) reactor and membrane separating cascade (MSC), designated for separation of isotope-enriched products from lean reagents. MSC represents the counterflow cascade for separation of two-component mixtures. Calculations show that for the process of carton isotope separation the electric power expences for MSC operation are equal to 20 kWxh/g of CO 2 final product at 13 C isotope content in it equal to 75%. Application of the membrane gas-separating cascade at rather small electric power expenses enables to perform cascading of isotope separation in the course of nonequilibrium chemical reactions

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

International Nuclear Information System (INIS)

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

1976-01-01

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

A Graphite Isotope Ratio Method: A Primer on Estimating Plutonium Production in Graphite Moderated Reactors

International Nuclear Information System (INIS)

Gesh, Christopher J.

2004-01-01

The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the Graphite Isotope Ratio Method is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel

External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)

International Nuclear Information System (INIS)

Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H.

1989-01-01

The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10 -4 . In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events

Production of Thorium-229 at the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Boll, Rose Ann; Garland, Marc A.; Mirzadeh, Saed

2008-01-01

The investigation of targeted cancer therapy using -emitters has developed considerably in recent years and clinical trials have generated promising results. In particular, the initial clinical trials for treatment of acute myeloid leukemia have demonstrated the effectiveness of the -emitter 213Bi in killing cancer cells. Pre-clinical studies have also shown the potential application of both 213Bi and its 225Ac parent radionuclide in a variety of cancer systems and targeted radiotherapy. Bismuth-213 is obtained from a radionuclide generator system from decay of the 10-d 225Ac parent, a member of the 7340-y 229Th chain. Currently, 233U is the only viable source for high purity 229Th; however, due to increasing difficulties associated with 233U safeguards, processing additional 233U is presently unfeasible. The recent decision to downblend and dispose of enriched 233U further diminished the prospects for extracting 229Th from 233U stock. Nevertheless, the anticipated growth in demand for 225Ac may soon exceed the levels of 229Th (∼40 g or ∼8 Ci; ∼80 times the current ORNL 229Th stock) present in the aged 233U stockpile. The alternative routes for the production of 229Th, 225Ra and 225Ac include both reactor and accelerator approaches. Here, we describe production of 229Th via neutron transmutation of 226Ra targets in the ORNL High Flux Isotope Reactor (HFIR).

Reactor production of 252Cf and transcurium isotopes

International Nuclear Information System (INIS)

Alexander, C.W.; Halperin, J.; Walker, R.L.; Bigelow, J.E.

1990-01-01

Berkelium, californium, einsteinium, and fermium are currently produced in the High Flux Isotope Reactor (HFIR) and recovered in the Radiochemical Engineering Development Center (REDC) at the Oak Ridge National Laboratory (ORNL). All the isotopes are used for research. In addition, 252 Cf, 253 Es, and 255 Fm have been considered or are used for industrial or medical applications. ORNL is the sole producer of these transcurium isotopes in the western world. A wide range of actinide samples were irradiated in special test assemblies at the Fast Flux Test Facility (FFTF) at Hanford, Washington. The purpose of the experiments was to evaluate the usefulness of the two-group flux model for transmutations in the special assemblies with an eventual goal of determining the feasibility of producing macro amounts of transcurium isotopes in the FFTF. Preliminary results from the production of 254g Es from 252 Cf will be discussed. 14 refs., 5 tabs

Isotopes accumulation in the thermal column of TRIGA reactor

International Nuclear Information System (INIS)

Iorgulis, C.; Diaconu, D.; Gugiu, D.; Csaba, R.

2013-01-01

The correlation of impurity observed in the virgin graphite and radionuclide content and activities measured in the irradiated graphite needs to know the irradiated history. This is a challenging process if impurity content and irradiation conditions are not accurately known. This is the case of the irradiated graphite in the thermal column of Institute for Nuclear Research Pitesti (INR)14 MW TRIGA reactor. To overcome incomplete impurity content and the unknown position in the column of the measured irradiated graphite available for characterisation and comparison, a set of preliminary simulations were performed. Following Eu 152 /Eu 154 ration they allowed the estimation of an impurity content and irradiation conditions leading to measured activities. Based on these data the radio-isotope accumulation in different positions in the thermal column was predicted. Modelling performed by INR used advanced prediction packages (e.g. WIMS, MCNP ORIGEN-S from Scale 5) to assess the isotopic content of MTR graphite types with irradiation history specific for a TRIGA research reactor. Some certain calculations points from the column were selected in order to model the burnup and isotopes productions using ORIGEN from SCALE code system. (authors)

Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

International Nuclear Information System (INIS)

Montgomery, B.H.; Thoms, K.R.; West, C.D.

1987-09-01

The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987

Operating manual for the High Flux Isotope Reactor. Description of the facility

Energy Technology Data Exchange (ETDEWEB)

None

1965-06-01

This report contains a comprehensive description of the High Flux Isotope Reactor facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procedures are presented in another report.

Target reactor development problems

International Nuclear Information System (INIS)

Lathrop, K.D.; Vigil, J.C.

1977-01-01

Target-blanket design studies are discussed for an accelerator-breeder concept employing a linear accelerator in conjunction with a modified conventional power reactor to produce both fissile fuel and power. The following problems in target and blanket system design are discussed: radiation damage, heat removal, neutronic design, and economics

Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01

An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

High Flux Isotope Reactor power upgrade status

International Nuclear Information System (INIS)

Rothrock, R.B.; Hale, R.E.; Cheverton, R.D.

1997-01-01

A return to 100-MW operation is being planned for the High Flux Isotope Reactor (HFIR). Recent improvements in fuel element manufacturing procedures and inspection equipment will be exploited to reduce hot spot and hot streak factors sufficiently to permit the power upgrade without an increase in primary coolant pressure. Fresh fuel elements already fabricated for future use are being evaluated individually for power upgrade potential based on their measured coolant channel dimensions

Pressurizer pump reliability analysis high flux isotope reactor

International Nuclear Information System (INIS)

Merryman, L.; Christie, B.

1993-01-01

During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

Determination of the theoretical feasibility for the transmutation of europium isotopes from high flux isotope reactor control cylinders

International Nuclear Information System (INIS)

Elam, K.R.; Reich, W.J.

1995-09-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) is a 100 MWth light-water research reactor designed and built in the 1960s primarily for the production of transuranic isotopes. The HFIR is equipped with two concentric cylindrical blade assemblies, known as control cylinders, that are used to control reactor power. These control cylinders, which become highly radioactive from neutron exposure, are periodically replaced as part of the normal operation of the reactor. The highly radioactive region of the control cylinders is composed of europium oxide in an aluminum matrix. The spent HFIR control cylinders have historically been emplaced in the ORNL Waste Area Grouping (WAG) 6. The control cylinders pose a potential radiological hazard due to the long lived radiotoxic europium isotopes 152 Eu, 154 Eu, and 155 Eu. In a 1991 health evaluation of WAG 6 (ERD 1991) it was shown that these cylinders were a major component of the total radioactivity in WAG 6 and posed a potential exposure hazard to the public in some of the postulated assessment scenarios. These health evaluations, though preliminary and conservative in nature, illustrate the incentive to investigate methods for permanent destruction of the europium radionuclides. When the cost of removing the control cylinders from WAG 6, performing chemical separations and irradiating the material in HFIR are factored in, the option of leaving the control cylinders in place for decay must be considered. Other options, such as construction of an engineered barrier around the disposal silos to reduce the chance of migration, should also be analyzed

Preparation of isotopically enriched mercury sulphide targets

Energy Technology Data Exchange (ETDEWEB)

Szerypo, J.; Friebel, H.U.; Frischke, D.; Grossman, R.; Maier, H.J. [Dept. fuer Physik, Univ. Muenchen (LMU) (Germany); Maier-Leibnitz-Lab. (MLL), Garching (Germany)

2007-07-01

The primary difficulty in performing nuclear reactions on mercury is to obtain a suitable target. The primary difficulty in performing nuclear reactions on mercury is to obtain a suitable target. The utilization of amalgam targets has been reported in early publications. These targets, however, were lacking homogeneity and in-beam stability. A thorough investigation of literature shows, that HgS, because of its comparatively high chemical and mechanical stability, is one of the more adequate Hg compounds for accelerator target applications. In this presentation we describe the production of HgS targets consisting of an enriched Hg isotope and S of natural isotopic abundance, starting up from HgO. Following the outline given in [3], in this special case HgS can be prepared by dissolving HgO in diluted HNO{sub 3} and subsequent precipitation of the black HgS modification with gaseous H{sub 2}S. Last step of the target production procedure is evaporation-condensation of HgS in vacuum. In the present case, HgS layers of 500 {mu}g/cm{sup 2} on a backing carbon foil of 26 {mu}g/cm{sup 2} with a protective carbon layer of about 20 {mu}g/cm{sup 2} thickness on top of the HgS layer were produced. (orig.)

Experimental spectrum of reactor antineutrinos and spectra of main fissile isotopes

Energy Technology Data Exchange (ETDEWEB)

Sinev, V. V., E-mail: vsinev@pcbai10.inr.ruhep.ru [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation)

2013-05-15

Within the period between the years 1988 and 1990, the spectrum of positrons from the inverse-beta-decay reaction on a proton was measured at the Rovno atomic power plant in the course of experiments conducted there. The measured spectrum has the vastest statistics in relation to other neutrino experiments at nuclear reactors and the lowest threshold for positron detection. An experimental reactor-antineutrino spectrum was obtained on the basis of this positron spectrum and was recommended as a reference spectrum. The spectra of individual fissile isotopes were singled out from the measured antineutrino spectrum. These spectra can be used to analyze neutrino experiments performed at nuclear reactors for various compositions of the fuel in the reactor core.

Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

International Nuclear Information System (INIS)

Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez; Universidade Federal de Pernambuco

2017-01-01

The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly "9"9Mo. Compare to multipurpose research reactors, an AHR dedicated for "9"9Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

Energy Technology Data Exchange (ETDEWEB)

Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez, E-mail: milianperez89@gmail.com, E-mail: dmilian@instec.cu, E-mail: lorenapilar1109@gmail.com, E-mail: cabol@ufpe.br [Higher Institute of Technologies and Applied Sciences (InSTEC), Havana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

2017-11-01

The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly {sup 99}Mo. Compare to multipurpose research reactors, an AHR dedicated for {sup 99}Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

Energy Technology Data Exchange (ETDEWEB)

1982-09-01

This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

International Nuclear Information System (INIS)

1982-09-01

This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

Hydrogen isotopes transport parameters in fusion reactor materials

International Nuclear Information System (INIS)

Serra, E.; Ogorodnikova, O.V.

1998-01-01

This work presents a review of hydrogen isotopes-materials interactions in various materials of interest for fusion reactors. The relevant parameters cover mainly diffusivity, solubility, trap concentration and energy difference between trap and solution sites. The list of materials includes the martensitic steels (MANET, Batman and F82H-mod.), beryllium, aluminium, beryllium oxide, aluminium oxide, copper, tungsten and molybdenum. Some experimental work on the parameters that describe the surface effects is also mentioned. (orig.)

Considerations in the design of a high power medical isotope production reactor

International Nuclear Information System (INIS)

Ball, Russell M.; Nordyke, William H.; Brown, Roy

2002-01-01

For the low enriched aqueous homogeneous reactor to be economic in the production of medical isotopes, such as Mo-99 and Sr-89, the power level should be of the order of 100 kWth. This is double the earlier designs and this paper discusses the design changes which must be considered to meet this goal. The topics considered are: 1. Heat removal from the reactor solution; 2. Recombination of radiolytic gases; 3. Adequate radiation shielding; 4. Stability of reactor power with fluctuating reactivity; 5. Adequate cooling of the reflector; 6. Independent shutdown mechanisms; 7. Required volume of the reactor; 8. Economic implementation. (author)

Availability of enriched isotopic material for accelerator targets

International Nuclear Information System (INIS)

Newman, E.

1982-01-01

The electromagnetic isotope enrichment facility at ORNL provides a broad spectrum of highly enriched stable isotopes to the worldwide scientific community. The continued timely availability of these materials is of vital importance in many areas of basic research and, in particular, as source material for the fabrication of accelerator targets. A brief description of the facility and its capabilities and limitations is presented

Reactor Fuel Isotopics and Code Validation for Nuclear Applications

Energy Technology Data Exchange (ETDEWEB)

Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-02-01

Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for ¹³³Cs and ¹⁵³Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

Laser Isotope Enrichment for Medical and Industrial Applications

Energy Technology Data Exchange (ETDEWEB)

Leonard Bond

2006-07-01

Laser Isotope Enrichment for Medical and Industrial Applications by Jeff Eerkens (University of Missouri), Jay Kunze (Idaho State University), and Leonard Bond (Idaho National Laboratory) The principal isotope enrichment business in the world is the enrichment of uranium for commercial power reactor fuels. However, there are a number of other needs for separated isotopes. Some examples are: 1) Pure isotopic targets for irradiation to produce medical radioisotopes. 2) Pure isotopes for semiconductors. 3) Low neutron capture isotopes for various uses in nuclear reactors. 4) Isotopes for industrial tracer/identification applications. Examples of interest to medicine are targets to produce radio-isotopes such as S-33, Mo-98, Mo-100, W-186, Sn-112; while for MRI diagnostics, the non-radioactive Xe-129 isotope is wanted. For super-semiconductor applications some desired industrial isotopes are Si-28, Ga-69, Ge-74, Se-80, Te-128, etc. An example of a low cross section isotope for use in reactors is Zn-68 as a corrosion inhibitor material in nuclear reactor primary systems. Neutron activation of Ar isotopes is of interest in industrial tracer and diagnostic applications (e.g. oil-logging). . In the past few years there has been a sufficient supply of isotopes in common demand, because of huge Russian stockpiles produced with old electromagnetic and centrifuge separators previously used for uranium enrichment. Production of specialized isotopes in the USA has been largely accomplished using old ”calutrons” (electromagnetic separators) at Oak Ridge National Laboratory. These methods of separating isotopes are rather energy inefficient. Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation

Laser Isotope Enrichment for Medical and Industrial Applications

International Nuclear Information System (INIS)

Leonard Bond

2006-01-01

Laser Isotope Enrichment for Medical and Industrial Applications by Jeff Eerkens (University of Missouri), Jay Kunze (Idaho State University), and Leonard Bond (Idaho National Laboratory) The principal isotope enrichment business in the world is the enrichment of uranium for commercial power reactor fuels. However, there are a number of other needs for separated isotopes. Some examples are: (1) Pure isotopic targets for irradiation to produce medical radioisotopes. (2) Pure isotopes for semiconductors. (3) Low neutron capture isotopes for various uses in nuclear reactors. (4) Isotopes for industrial tracer/identification applications. Examples of interest to medicine are targets to produce radio-isotopes such as S-33, Mo-98, Mo-100, W-186, Sn-112; while for MRI diagnostics, the non-radioactive Xe-129 isotope is wanted. For super-semiconductor applications some desired industrial isotopes are Si-28, Ga-69, Ge-74, Se-80, Te-128, etc. An example of a low cross section isotope for use in reactors is Zn-68 as a corrosion inhibitor material in nuclear reactor primary systems. Neutron activation of Ar isotopes is of interest in industrial tracer and diagnostic applications (e.g. oil-logging). . In the past few years there has been a sufficient supply of isotopes in common demand, because of huge Russian stockpiles produced with old electromagnetic and centrifuge separators previously used for uranium enrichment. Production of specialized isotopes in the USA has been largely accomplished using old ''calutrons'' (electromagnetic separators) at Oak Ridge National Laboratory. These methods of separating isotopes are rather energy inefficient. Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation

Application of expert systems to heat exchanger control at the 100-megawatt high-flux isotope reactor

International Nuclear Information System (INIS)

Clapp, N.E. Jr.; Clark, F.H.; Mullens, J.A.; Otaduy, P.J.; Wehe, D.K.

1985-01-01

The High-Flux Isotope Reactor (HFIR) is a 100-MW pressurized water reactor at the Oak Ridge National Laboratory. It is used to produce isotopes and as a source of high neutron flux for research. Three heat exchangers are used to remove heat from the reactor to the cooling towers. A fourth heat exchanger is available as a spare in case one of the operating heat exchangers malfunctions. It is desirable to maintain the reactor at full power while replacing the failed heat exchanger with the spare. The existing procedures used by the operators form the initial knowledge base for design of an expert system to perform the switchover. To verify performance of the expert system, a dynamic simulation of the system was developed in the MACLISP programming language. 2 refs., 3 figs

Heavy water isotopic rectification in the ''ORPHEE'' reactor. SACLAY studies Centre

International Nuclear Information System (INIS)

Lejeune, P.; Breant, P.

1993-01-01

ORPHEE reactor supplies neutron beams, which are got back in a heavy water reflector. The neutron beams intensity depends on the reflector quality which is determined by the isotopic content of the heavy water. The deuterium submitted to core irradiation changes in radioactive tritium which must be eliminated largely for reasons of safety. The column must keep the heavy water isotopic content of the reflector to a value higher than 99.8% by eliminating light water by fractional distillation or rectification. This column is also used for the tritium elimination of heavy water. 13 figs

Evaluation of the performance of high temperature conversion reactors for compound-specific oxygen stable isotope analysis.

Science.gov (United States)

Hitzfeld, Kristina L; Gehre, Matthias; Richnow, Hans-Hermann

2017-05-01

In this study conversion conditions for oxygen gas chromatography high temperature conversion (HTC) isotope ratio mass spectrometry (IRMS) are characterised using qualitative mass spectrometry (IonTrap). It is shown that physical and chemical properties of a given reactor design impact HTC and thus the ability to accurately measure oxygen isotope ratios. Commercially available and custom-built tube-in-tube reactors were used to elucidate (i) by-product formation (carbon dioxide, water, small organic molecules), (ii) 2nd sources of oxygen (leakage, metal oxides, ceramic material), and (iii) required reactor conditions (conditioning, reduction, stability). The suitability of the available HTC approach for compound-specific isotope analysis of oxygen in volatile organic molecules like methyl tert-butyl ether is assessed. Main problems impeding accurate analysis are non-quantitative HTC and significant carbon dioxide by-product formation. An evaluation strategy combining mass spectrometric analysis of HTC products and IRMS 18 O/ 16 O monitoring for future method development is proposed.

Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Jain, Prashant K [ORNL; Freels, James D [ORNL

2015-01-01

Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

Production yields of noble-gas isotopes from ISOLDE UC$_{x}$/graphite targets

CERN Document Server

Bergmann, U C; Catherall, R; Cederkäll, J; Diget, C A; Fraile-Prieto, L M; Franchoo, S; Fynbo, H O U; Gausemel, H; Georg, U; Giles, T; Hagebø, E; Jeppesen, H B; Jonsson, O C; Köster, U; Lettry, Jacques; Nilsson, T; Peräjärvi, K; Ravn, H L; Riisager, K; Weissman, L; Äystö, J

2003-01-01

Yields of He, Ne, Ar, Kr and Xe isotopic chains were measured from UC$_{x}$/graphite and ThC$_{x}$/graphite targets at the PSB-ISOLDE facility at CERN using isobaric selectivity achieved by the combination of a plasma-discharge ion source with a water-cooled transfer line. %The measured half-lives allowed %to calculate the decay losses of neutron-rich isotopes in the %target and ion-source system, and thus to obtain information on the in-target %productions from the measured yields. The delay times measured for a UC$_x$/graphite target allow for an extrapolation to the expected yields of very neutron-rich noble gas isotopes, in particular for the ``NuPECC reference elements'' Ar and Kr, at the next-generation radioactive ion-beam facility EURISOL. \\end{abstract} \\begin{keyword} % keywords here, in the form: keyword \\sep keyword radioactive ion beams \\sep release \\sep ion yields \\sep ISOL (Isotope Separation On-Line) \\sep uranium and thorium carbide targets. % PACS codes here, in the form: \\PACS code \\sep code...

Production of medical radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for cancer treatment and arterial restenosis therapy after PTCA

International Nuclear Information System (INIS)

Knapp, F.F. Jr.; Beets, A.L.; Mirzadeh, S.; Alexander, C.W.; Hobbs, R.L.

1998-01-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed

Development of a simplified methodology for the isotopic determination of fuel spent in Light Water Reactors

International Nuclear Information System (INIS)

Hernandez N, H.; Francois L, J.L.

2005-01-01

The present work presents a simplified methodology to quantify the isotopic content of the spent fuel of light water reactors; their application is it specific to the Laguna Verde Nucleo electric Central by means of a balance cycle of 18 months. The methodology is divided in two parts: the first one consists on the development of a model of a simplified cell, for the isotopic quantification of the irradiated fuel. With this model the burnt one is simulated 48,000 MWD/TU of the fuel in the core of the reactor, taking like base one fuel assemble type 10x10 and using a two-dimensional simulator for a fuel cell of a light water reactor (CPM-3). The second part of the methodology is based on the creation from an isotopic decay model through an algorithm in C++ (decay) to evaluate the amount, by decay of the radionuclides, after having been irradiated the fuel until the time in which the reprocessing is made. Finally the method used for the quantification of the kilograms of uranium and obtained plutonium of a normalized quantity (1000 kg) of fuel irradiated in a reactor is presented. These results will allow later on to make analysis of the final disposition of the irradiated fuel. (Author)

Importance of resonance parameters of fertile nuclei and of 239Pu isotope for fast power reactors

International Nuclear Information System (INIS)

Barre, J.Y.; Khairallah, A.

1975-01-01

The importance of resonance parameters of fertile nuclei and of 239 Pu isotope for fast power reactors will be restricted, in this presentation, to mixed oxide-uranium-plutonium fuelled sodium-cooled and uranium-oxide-sodium reflected fast reactors. The power range lies between 200 and 2000 MWe. Among the topics of this specialist meeting, the isotopes to be considered are, primarly 239 Pu then 238 U and 240 Pu. Resonance parameters are mainly used in fast power reactor calculations through the well-known concept of self shielding factors. After a short description of the determination and the use of these self-shielding factors, their sensitivities to resonance parameters are characterized from some specific examples: those sensitivities are small. Then, the main design parameters sensitive to the amplitude of self-shielding factors are considered: critical enrichment, global breeding gain. The relative importance of isotope, reaction rate and energy range are mentionned. In a third part, the Doppler effect, sensitive to the temperature variation of self-shielding factors, is considered in the same way. Finally, it is concluded that the present knowledge of resonance parameters for 238 U, 239 Pu and 240 Pu is sufficient for fast power reactors from a designer point of view [fr

Problems in producing nuclear reactor for medical isotopes and the Global Crisis of molybdenum supply

International Nuclear Information System (INIS)

Zubiarrain, A.

2011-01-01

Nuclear medicine uses drugs that incorporate a radioactive isotope radiopharmaceuticals. Every year are performed, worldwide, 35 million nuclear medicine procedures, of which 80% are done with radiopharmaceuticals containing the isotope, molybdenum-99, produced in nuclear reactors. In recent years, there have been several supply crisis of molybdenum-99, which have hampered diagnostic procedure with technitium-99m. (Author)

Canadian Neutron Source (CNS): a research reactor solution for medical isotopes and neutrons for science

International Nuclear Information System (INIS)

Chapman, D.

2009-01-01

This presentation describes a dual purpose research facility at the University of Saskatchewan for Canada for the production of medical isotopes and neutrons for scientific research. The proposed research reactor is intended to supply most of Canada's medical isotope requirements and provide a neutron source for Canada's research community. Scientific research would include materials research, biomedical research and imaging.

The Maple reactor project

International Nuclear Information System (INIS)

Malkoske, G.R.; Labrie, J.-P.

2003-01-01

MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

International Nuclear Information System (INIS)

Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok

2015-01-01

The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235 U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233 U, 242 Pu, 150 Nd, and 133 Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code

Development of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.

1997-01-01

The Reduced Enrichment Research and Test Reactor Program has continued its effort in the past 3 yr to develop use of low-enriched uranium (LEU) to produce the fission product 99 Mo. This work comprises both target and chemical processing development and demonstration. Two major target systems are now being used to produce 99 Mo with highly enriched uranium-one employing research reactor fuel technology (either uranium-aluminum alloy or uranium aluminide-aluminum dispersion) and the other using a thin deposit of UO 2 on the inside of a stainless steel (SST) tube. This paper summarizes progress in irradiation testing of targets based on LEU uranium metal foils. Several targets of this type have been irradiated in the Indonesian RSG-GAS reactor operating at 22.5 MW

Component and system simulation models for High Flux Isotope Reactor

International Nuclear Information System (INIS)

Sozer, A.

1989-08-01

Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs

Mixing rules for and effects of other hydrogen isotopes and of isotopic swamping on tritium recovery and loss to biosphere from fusion reactors

International Nuclear Information System (INIS)

Pendergrass, J.H.

1978-01-01

Efficient recovery of bred and unburnt tritium from fusion reactors, and control of its migration within reactors and of its escape into the biosphere are essential for self-sufficient fuel cycles and for public, plant personnel, and environmental protection. Tritium in fusion reactors will be mixed with unburnt deuterium and protium introduced by (n,p) reactions and diffusion into coolant loops from steam cycles. Rational design for tritium recovery and escape prevention must acknowledge this fact. Consequences of isotopic admixture are explored, mixing rules for projected fusion reactor dilute-solution conditions are developed, and a rule of thumb regarding their effects on tritium recovery methods is formulated

Isotopic nuclear reactor with on-line separation

International Nuclear Information System (INIS)

Liviu, Popa-Simil

2007-01-01

In the new reactor-waste cycle design the nuclear reactor gets features of the living beings - resembling the plants/vegetation -. The separation of waste starts inside the fuel by using the fission reaction to separate the fission products from the fuel. The fuel, which is preferred to be highly isotopic enriched, is fabricated in beads smaller than the fission product range, immersed in a gentle flowing liquid drain. If this liquid is Lead Bismuth (LBE) the fission products will be lighter, while in Sodium-Potassium (NaK) will be heavier, except for gases. This drain liquid will collect both the fission products and the collision damage, drawing them slow to give time to short lives disintegration chains to take place inside the shielded nuclear reactor area outside the reactor core in a separation unit. While the drain liquid with the fission products is outside the reactor core few choices are available: - To solidify the drain liquid freezing all elements inside and transport the metal in cryogenic conditions to a remote separation unit, or to apply a separation partitioning process online stabilizing and packing the fission products only, or a combination of these two. The radioactivity of this drain liquid is smaller than that of the actual used fuel because it represents the accumulation of a very short period (about 1 month or less) and had enough time to cool down all the short lives. The separation unit on-line with the nuclear reactor is composed of a density separation unit, followed by a phase interface concentration unit which moves out of the LBE the fission products as lighter impurities, and an electrochemical separation unit for the fission products. Further, chemical separation, stabilization processes are applied and the fission products are delivered partitioned on groups of chemical compatible products. Finally the specific waste is about 1 Kg/Gw*day, to which the stabilization products have to be added which increases this mass by 10 times

Research reactor core conversion programmes, Department of Research and Isotopes, International Atomic Energy Agency

International Nuclear Information System (INIS)

Muranaka, R.G.

1985-01-01

In order to put the problem of core conversion into perspective, statistical information on research reactors on a global scale is presented (from IAEA Research reactor Data Base). This paper describes the research reactor core conversion program of the Department of Research and Isotopes. Technical committee Meetings were held on the subject of research reactor core conversion since 1978, and results of these meetings are published in TECDOC-233, TECDOC-324, TECDOC-304. Additional publications are being prepared, several missions of experts have visited countries to discuss and help to plan core conversion programs; training courses and seminars were organised; IAEA has supported attendance of participants from developing countries to RERTR Meetings

Application of mass-predictions to isotope-abundances in breeder-reactor cores

CERN Document Server

Kirchner, G

1981-01-01

The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGEN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technical viewpoint. (15 refs).

Preparation of calcium-separated isotope targets using small samples

International Nuclear Information System (INIS)

Thomas, G.E.

1975-01-01

Targets are routinely evaporated using a few milligram quantities of separated isotopes of calcium with reducing agents. The source to target distance is 3.0 cm with the substrate, if necessary, as thin as 15 μg/cm 2 carbon or 100 μg/cm 2 of gold. A tantalum closed boat, heat shield, and special collimator system are used

Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

International Nuclear Information System (INIS)

Wulff, W.

1990-01-01

A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs

Economic targets for small PWR reactor designs

International Nuclear Information System (INIS)

Board, J.

1991-01-01

Small reactors are likely to be less economic than large reactors, but the lower financial exposure with small reactors may be attractive to utilities contemplating a restart to a nuclear programme. New nuclear plant can be economic, but success will depend more on how the plant are built, rather than what type or size is built. A target for new plant for operation early in the next century should be a generation cost of 3p to 3.5 p/kWh. This corresponds to an overnight capital cost of Pound 1000/kWh to Pound 1100/kWh. (author)

Medical Isotope Production at TRIUMF - from Imaging to Treatment

Science.gov (United States)

Hoehr, C.; Bénard, F.; Buckley, K.; Crawford, J.; Gottberg, A.; Hanemaayer, V.; Kunz, P.; Ladouceur, K.; Radchenko, V.; Ramogida, C.; Robertson, A.; Ruth, T.; Zacchia, N.; Zeisler, S.; Schaffer, P.

TRIUMF has a long history of medical isotope production. For more than 40 years, the Life Sciences Division at TRIUMF has produced isotopes for Positron Emission Tomography (PET) for the local hospitals. Recently, the division has taken on the challenge to expand the facility's isotope repertoire to isotopes for imaging to treatment. At the smallest cyclotron at TRIUMF with energy of 13 MeV, radiometals are being produced in a liquid target which is typically used for PET isotope production. This effort makes radiometals available for early stage research and preclinical trials. At beam energy of 24 MeV, we produce 99mTc from 100Mo with a cyclotron, the most common isotope for Single-Photon-Emission-Computed-Tomography (SPECT) and the most common isotope for nuclear imaging. The use of a cyclotron bypasses the common production route via a nuclear reactor as well as enriched uranium. And finally, at our 500 MeV cyclotron we have demonstrated the production of α emitters useful for targeted alpha therapy. Herein, these efforts are summarized.

Lessons learned form high-flux isotope reactor restart efforts

International Nuclear Information System (INIS)

Dahl, T.L.

1989-01-01

When the high-flux isotope reactor's (HFIR's) pressure vessel irradiation surveillance specimens were examined in December 1986, unexpected embrittlement was found. The resulting investigation disclosed widespread deficiencies in quality assurance and management practices. On March 24, 1987, the US Department of Energy (DOE) mandated a shutdown of all five Oak Ridge National Laboratory (ORNL) research reactors. Since the beginning of 1987, 18 different formal review groups have evaluated the management and operations of the HFIR. The root cause of the identified deficiencies in the HFIR program was defined as a lack of rigor in management practices and complacency built on twenty years of trouble-free operation. A number of lessons can be learned from the HFIR experience. Particular insight can be gained by comparing the HFIR organization prior to the shutdown with the organization that exists today. Key elements in such a comparison include staffing, funding, discipline, and formality in operations, maintenance, and management

ANITA (Advanced Network for Isotope and TArget laboratories) - The urgent need for a European target preparation network

Science.gov (United States)

Schumann, Dorothea; Sibbens, Goedele; Stolarz, Anna; Eberhardt, Klaus; Lommel, Bettina; Stodel, Christelle

2018-05-01

A wide number of research fields in the nuclear sector requires high-quality and well-characterized samples and targets. Currently, only a few laboratories own or have access to the equipment allowing fulfilling such demands. Coordination of activities and sharing resources is therefore mandatory to meet the increasing needs. This very urgent issue has now been addressed by six European target laboratories with an initiative called ANITA (Advanced Network for Isotope and TArget laboratories). The global aim of ANITA is to establish an overarching research infrastructure service for isotope and target production and develop a tight cooperation between the target laboratories in Europe in order to transfer the knowledge and improve the production techniques of well-characterized samples and targets. Moreover, the interaction of the target producers with the users shall be encouraged and intensified to deliver tailor-made targets best-suited to the envisaged experiments. For the realization of this ambitious goal, efforts within the European Commission and strong support by the target-using communities will be necessary. In particular, an appropriate funding instrument has to be found and applied, enabling ANITA to develop from an initiative employed by the interested parties to a real coordination platform.

Research reactor support

International Nuclear Information System (INIS)

2005-01-01

Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR [High Flux Isotope Reactor] Reactor

International Nuclear Information System (INIS)

Childs, R.L.; Rhoades, W.A.; Williams, L.R.

1988-01-01

The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs

Evolution of the hafnium isotopic composition in the RBMK reactor

International Nuclear Information System (INIS)

Jurkevicius, A.; Remeikis, V.

2002-01-01

The isotopic composition of hafnium in the radial neutron flux sensor of the RBMK-1500 reactor, the rates of the neutron absorption on Hf isotopes and the neutron spectrum in the sensor were numerically modeled. The sequence SAS2 (Shielding Analysis Sequence) program from the package SCALE 4.4A and the HELIOS code system were used for calculations. It has been obtained that the overall neutron absorption rates in hafnium for the sensors located in the 2.4 % and 2.6 % enrichment uranium-erbium nuclear fuel assemblies are by 16 % and 19 % lower than in the 2.0 % enrichment uranium nuclear fuel assemblies. The overall neutron absorption rate in hafnium decreases 2.70-2.75 times due to the sensor burnup to 5800 MW d. The sensitivity of the Hf sensors to the thermal neutron flux increases twice due to the nuclear fuel assembly burnup to 3000 MW d. The corrective factors ξ d (I) at the different integral current I of the sensors and ξ td (E) at the different burnup E of the nuclear fuel assemblies were calculated. The obtained dependence ξ d (I) calculated numerically was compared to the experimental one determined by comparing signals of the fresh sensor and the sensor with the integral current I and by processing repeated calibration results of Hf sensors in RBMK-1500 reactors. The relative relationship coefficients K T (T FA ) were found for all RBMK-1500 nuclear fuel types. (author)

Final report on production of Pu-238 in commercial power reactors: target fabrication, postirradiation examination, and plutonium and neptunium recovery

International Nuclear Information System (INIS)

Pobereskin, M.; Langendorfer, W.; Lowry, L.; Farmelo, D.; Scotti, V.; Kruger, O.

1975-01-01

Considerable interest has been generated in more extensive applications of radioisotope thermoelectric generator (RTG) systems. This raises questions concerning the availability of 238 Pu to supply an expanding demand. The development of much of this demand will depend upon a considerable reduction in cost of 238 Pu. Two neptunia--zirconia--fuel target rods, containing four sections each of different NpO 2 concentrations, were irradiated in the Connecticut Yankee Reactor for approximately one year. Following irradiation both target rods were subjected to nondestructive examination. One rod was chosen for destructive testing and analysis. Post-irradiation chemical analyses included total Pu and Np, ppM 236 Pu/ 238 Pu, and Pu isotopic abundance. The results of these analyses and of electron microprobe analysis which provided the relative Pu concentration across the pellet diameters are tabulated. It was concluded that the feasibility of all operations involved in the production of 238 Pu by irradiation of 237 NpO 2 targets in commercial nuclear power reactors was demonstrated and that the demonstration should be extended to a pilot-scale leading to installation of a full production capacity. (U.S.)

Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

International Nuclear Information System (INIS)

Bays, Samuel; Medvedev, Pavel; Pope, Michael; Ferrer, Rodolfo; Forget, Benoit; Asgari, Mehdi

2009-01-01

This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

Hybrid nuclear reactors and muon catalysis

International Nuclear Information System (INIS)

Petrov, Yu.

1983-01-01

Three methods are described of the conversion of isotope 238 U to 239 Pu by neutron capture in fast breeder reactors, in the breeding blanket of hybrid thermonuclear reactors using neutrons generated by fusion and electronuclear breeding in which the target is bombarded with 1 GeV protons. Their possible use in power production is discussed. Another prospective energy source is the use of muon catalysis in the fusion of deuterium and tritium nuclei. (J.P.)

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Preparation of homogeneous isotopic targets with rotating substrate

International Nuclear Information System (INIS)

Xu, G.J.; Zhao, Z.G.

1993-01-01

Isotopically enriched accelerator targets were prepared using the evaporation-condensation method from a resistance heating crucible. For high collection efficiency and good homogeneity the substrate was rotated at a vertical distance of 1.3 to 2.5 cm from the evaporation source. Measured collection efficiencies were 13 to 51 μg cm -2 mg -1 and homogeneity tests showed values close to the theoretically calculated ones for a point source. Targets, selfsupporting or on backings, could be fabricated with this method for elements and some compounds with evaporation temperatures up to 2300 K. (orig.)

Seismic analysis of fuel and target assemblies at a production reactor

International Nuclear Information System (INIS)

Braverman, J.I.; Wang, Y.K.

1991-01-01

This paper describes the unique modeling and analysis considerations used to assess the seismic adequacy of the fuel and target assemblies in a production reactor at Savannah River Site. This confirmatory analysis was necessary to provide assurance that the reactor can operate safely during a seismic event and be brought to a safe shutdown condition. The plant which was originally designed in the 1950's required to be assessed to more current seismic criteria. The design of the reactor internals and the magnitude of the structural responses enabled the use of a linear elastic dynamic analysis. A seismic analysis was performed using a finite element model consisting of the fuel and target assemblies, reactor tank, and a portion of the concrete structure supporting the reactor tank. The effects of submergence of the fuel and target assemblies in the water contained within the reactor tank can have a significant effect on their seismic response. Thus, the model included hydrodynamic fluid coupling effects between the assemblies and the reactor tank. Fluid coupling mass terms were based on formulations for solid bodies immersed in incompressible and frictionless fluids. The potential effects of gap conditions were also assessed in this evaluation. 5 refs., 6 figs., 1 tab

Reactor-produced radionuclides at the University of Missouri Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Ketring, A.R.; Evans-Blumer, M.S.; Ehrhardt, G.J. [University of Missouri Research Reactor, Colombia (United States). Departments of Radiology, Chemistry and Nuclear Engineering

1997-10-01

Nuclear medicine has primarily been a diagnostic science for many years, but today is facing considerable challenges from other modalities in this area. However, these competing techniques (magnetic resonance imaging, ultrasound, and computer-assisted tomography) in general are not therapeutic. Although early nuclear medicine therapy was of limited efficacy, in recent years a revolution in radiotherapy has been developing base don more sophisticated targeting methods, including radioactive intra-arterial microspheres, chemically-guided bone agents, labelled monoclonal antibodies, and isotopically-tagged polypeptide receptor-binding agents. Although primarily used for malignancies, therapeutic nuclear medicine is also applicable to the treatment of rheumatoid arthritis and possibly coronary artery re closure following angioplasty. The isotopes of choice for these applications are reactor-produced beta emitters such as Sm-153, Re-186, Re-188, Ho-166, Lu-177, and Rh-105. Although alpha emitters possess greater cell toxicity due to their high LET, the greater range of beta emitters and the typically inhomogeneous deposition of radiotherapy agents in lesions leads to greater beta `crossfire` and better overall results. The University of Missouri Research Reactor (MURR) has been in the forefront of research into means of preparing, handling and supplying these high-specific-activity isotopes in quantities appropriate not only for research, but also for patient trials in the US and around the world. Researchers at MURR in collaboration with others at the University of Missouri (MU) developed Sm-153 Quadramet{sup TM}, a drug recently approved in the US for palliation of bone tumor pain. In conjunction with researchers at the University of Missouri-Rolla, MURR also developed Y-90 TheraSphere{sup TM}, an agent for the treatment of liver cancer now approved in Canada. Considerable effort has been expended to develop techniques for irradiation, handling, and shipping isotopes

Reactor-produced radionuclides at the University of Missouri Research Reactor

International Nuclear Information System (INIS)

Ketring, A.R.; Evans-Blumer, M.S.; Ehrhardt, G.J.

1997-01-01

Nuclear medicine has primarily been a diagnostic science for many years, but today is facing considerable challenges from other modalities in this area. However, these competing techniques (magnetic resonance imaging, ultrasound, and computer-assisted tomography) in general are not therapeutic. Although early nuclear medicine therapy was of limited efficacy, in recent years a revolution in radiotherapy has been developing base don more sophisticated targeting methods, including radioactive intra-arterial microspheres, chemically-guided bone agents, labelled monoclonal antibodies, and isotopically-tagged polypeptide receptor-binding agents. Although primarily used for malignancies, therapeutic nuclear medicine is also applicable to the treatment of rheumatoid arthritis and possibly coronary artery re closure following angioplasty. The isotopes of choice for these applications are reactor-produced beta emitters such as Sm-153, Re-186, Re-188, Ho-166, Lu-177, and Rh-105. Although alpha emitters possess greater cell toxicity due to their high LET, the greater range of beta emitters and the typically inhomogeneous deposition of radiotherapy agents in lesions leads to greater beta 'crossfire' and better overall results. The University of Missouri Research Reactor (MURR) has been in the forefront of research into means of preparing, handling and supplying these high-specific-activity isotopes in quantities appropriate not only for research, but also for patient trials in the US and around the world. Researchers at MURR in collaboration with others at the University of Missouri (MU) developed Sm-153 Quadramet TM , a drug recently approved in the US for palliation of bone tumor pain. In conjunction with researchers at the University of Missouri-Rolla, MURR also developed Y-90 TheraSphere TM , an agent for the treatment of liver cancer now approved in Canada. Considerable effort has been expended to develop techniques for irradiation, handling, and shipping isotopes

Determination of spallation residues in thin target: toward an hybrid reactor lead target simulation

International Nuclear Information System (INIS)

Audouin, L.; Tassan-Got, L.; Bernas, M.; Rejmund, F.; Stephan, C.; Taieb, J.; Boudard, A.; Fernandez, B.; Legrain, R.; Leray, S.; Volant, C.; Wlazlo, W.; Benlliure, J.; Casajeros, E.; Pereira, J.; Czajkowski, S.

2001-01-01

The production of spallation primary residual nuclei in thin target has been studied by measurement of isotopic yields distributions for several systems. Issues relevant for the design of accelerator-driven systems are presented. Monte-Carlo code abilities to reproduce data are studied in details; it is shown that calculations do not reproduce data in a satisfactory way. Future work orientations leading to an improvement of thin targets calculations and ultimately to a thick target simulation are discussed. (author)

Determination of spallation residues in thin target: toward an hybrid reactor lead target simulation

Energy Technology Data Exchange (ETDEWEB)

Audouin, L.; Tassan-Got, L.; Bernas, M.; Rejmund, F.; Stephan, C.; Taieb, J. [Paris-11 Univ., 91- Orsay (France). Inst. de Physique Nucleaire; Enqvist, T.; Armbruster, P.; Ricciardi, M.V.; Schmidt, K.H. [GSI, Planckstrasse 1, Darmstadt (Germany); Boudard, A.; Fernandez, B.; Legrain, R.; Leray, S.; Volant, C.; Wlazlo, W. [CEA Saclay, Dept. d' Astrophysique, de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee, 91 - Gif sur Yvette (France); Benlliure, J.; Casajeros, E.; Pereira, J. [University of Santiago de Compostella (Spain); Czajkowski, S. [Centre d' Etudes Nucleaires de Bordeaux Gradignan, CENBG, CNRS-IN2P3, 33 - Gradignan (France)

2001-07-01

The production of spallation primary residual nuclei in thin target has been studied by measurement of isotopic yields distributions for several systems. Issues relevant for the design of accelerator-driven systems are presented. Monte-Carlo code abilities to reproduce data are studied in details; it is shown that calculations do not reproduce data in a satisfactory way. Future work orientations leading to an improvement of thin targets calculations and ultimately to a thick target simulation are discussed. (author)

Management of safety and risk at the HFIR [High-Flux Isotope Reactor

International Nuclear Information System (INIS)

Glovier, H.A.

1990-01-01

This paper discusses the management of safety and risk at the High-Flux Isotope Reactor (HFIR), a category A research reactor at Oak Ridge National Laboratory (ORNL). The HFIR went critical in 1966 and operated at its designed 100 MW for 20 yr until it was shut down on November 14, 1986. It operated at a >90% availability and without significant event during this period. The result was a complacent management program lacking rigor. This complacency came to an end with the Chernobyl accident, which led to the appointment of an internal committee to assess the safety of ORNL reactor operations. This committee found that HFIR pressure vessel material specimens removed several years earlier had not been analyzed. This issue led to a general review of management practices that were found lacking in quality assurance, safety documentation, training process, and emergency planning, among others. Management accountability was lacking, as shown by design basis and safety analyses that were not up to data and by the fact that reactor operators whose requalification examinations had not been graded were allowed to continue operating the reactor over a long period of time. Between shutdown in 1986 and restart in April 1989, significant management changes and initiatives were made in the area of risk and safety management of ORNL reactors. These are presented briefly in this paper

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N. [National Research Centre Kurchatov Institute (Russian Federation); Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu., E-mail: yuri.titarenko@itep.ru [Institute for Theoretical and Experimental Physics (Russian Federation)

2015-12-15

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

Behavior of antimony isotopes in the primary coolant of WWER-1000-type nuclear reactors in NPP Kozloduy during operation and shutdown

International Nuclear Information System (INIS)

Dobrevski, Ivan D.; Zaharieva, Neli N.; Minkova, Katia F.; Gerchev, Nikolay B.

2009-01-01

This paper focuses on the behavior of the antimony isotopes 122 Sb and 124 Sb in the coolant of the WWER reactors in the nuclear power plant Kozloduy (Bulgaria) during operation and shutdown. It is concluded that the chemical properties of their actual precursor, the isotope 121 Sb, determine the behavior of 122 Sb and 124 Sb during operation, load fluctuations, and shutdown as well as during the reactor coolant purification process. It is supposed that differences between the reactor bulk and the core fuel cladding surface chemistry as well as the presence of sub-cooled nucleate boiling at the fuel cladding may create conditions under which a local oxidizing environment may come into existence. (orig.)

Low-enriched uranium high-density target project. Compendium report

Energy Technology Data Exchange (ETDEWEB)

Vandegrift, George; Brown, M. Alex; Jerden, James L.; Gelis, Artem V.; Stepinski, Dominique C.; Wiedmeyer, Stanley; Youker, Amanda; Hebden, Andrew; Solbrekken, G; Allen, C; Robertson., D; El-Gizawy, Sherif; Govindarajan, Srisharan; Hoyer, Annemarie; Makarewicz, Philip; Harris, Jacob; Graybill, Brian; Gunn, Andy; Berlin, James; Bryan, Chris; Sherman, Steven; Hobbs, Randy; Griffin, F. P.; Chandler, David; Hurt, C. J.; Williams, Paul; Creasy, John; Tjader, Barak; McFall, Danielle; Longmire, Hollie

2016-09-01

At present, most ⁹⁹Mo is produced in research, test, or isotope production reactors by irradiation of highly enriched uranium targets. To achieve the denser form of uranium needed for switching from high to low enriched uranium (LEU), targets in the form of a metal foil (~125-150 µm thick) are being developed. The LEU High Density Target Project successfully demonstrated several iterations of an LEU-fission-based Mo-99 technology that has the potential to provide the world’s supply of Mo-99, should major producers choose to utilize the technology. Over 50 annular high density targets have been successfully tested, and the assembly and disassembly of targets have been improved and optimized. Two target front-end processes (acidic and electrochemical) have been scaled up and demonstrated to allow for the high-density target technology to mate up to the existing producer technology for target processing. In the event that a new target processing line is started, the chemical processing of the targets is greatly simplified. Extensive modeling and safety analysis has been conducted, and the target has been qualified to be inserted into the High Flux Isotope Reactor, which is considered above and beyond the requirements for the typical use of this target due to high fluence and irradiation duration.

Data book of the isotopic composition of spent fuel in light water reactors

International Nuclear Information System (INIS)

Naito, Yoshitaka; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

1994-03-01

In the framework of the activity of the working group on Evaluation of Nuclide Generation and Depletion in the Japanese Nuclear Data Committee, we summarized the assay data of the isotopic composition of LWR spent fuels in order to verify the accuracy of the burnup calculation codes. The report contains the data collected from the 13 light water reactors (LWRs) including the 9 LWRs (5 PWRs and 4 BWRs) in Europe and USA, the 4 LWRs (2 PWRs and 2 BWRs) in Japan. The collected data were sorted into the irradiation history of the fuel samples, the composition of the fuel assemblies, the sampling position and the isotopic composition of the fuel samples. (author)

Use of LEU in the aqueous homogeneous medical isotope production reactor

Energy Technology Data Exchange (ETDEWEB)

Ball, R.M. [Babock & Wilcox, Lynchburg, VA (United States)

1997-08-01

The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.

Use of LEU in the aqueous homogeneous medical isotope production reactor

International Nuclear Information System (INIS)

Ball, R.M.

1997-01-01

The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution

On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

International Nuclear Information System (INIS)

Freels, J.D.

1993-01-01

This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed

Fabrication of control rods for the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Sease, J.D.

1998-01-01

The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A

Fabrication of control rods for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Sease, J.D.

1998-03-01

The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

The development of fast tantalum foil targets for short-lived isotopes

CERN Document Server

Bennett, J R J; Drumm, P V; Ravn, H L

2003-01-01

The development of fast tantalum foil targets for short-lived isotopes was discussed. It was found that the effusion was faster but the diffusion out of the foils was a limiting factor. The performance of the targets at ISOLDE with beams of **1**1Li, **1**2Be and **1**4Be was also analyzed. (Edited abstract) 13 Refs.

Status of High Flux Isotope Reactor (HFIR) post-restart safety analysis and documentation upgrades

International Nuclear Information System (INIS)

Cook, D.H.; Radcliff, T.D.; Rothrock, R.B.; Schreiber, R.E.

1990-01-01

The High Flux Isotope Reactor (HFIR), an experimental reactor located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, was shut down in November, 1986 after the discovery of unexpected neutron embrittlement of the reactor vessel. The reactor was restarted in April, 1989, following an extensive review by DOE and ORNL of the HFIR design, safety, operation, maintenance and management, and the implementation of several upgrades to HFIR safety-related hardware, analyses, documents and procedures. This included establishing new operating conditions to provide added margin against pressure vessel failure, as well as the addition, or upgrading, of specific safety-related hardware. This paper summarizes the status of some of the follow-on (post-restart) activities which are currently in progress, and which will result in a comprehensive set of safety analyses and documentation for the HFIR, comparable with current practice in commercial nuclear power plants. 8 refs

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Eddy-current inspection of high flux isotope reactor nuclear control rods

International Nuclear Information System (INIS)

Smith, J.H.; Chitwood, L.D.

1981-07-01

Inner control rods for the High Flux Isotope Reactor were nondestructively inspected for defects by eddy-current techniques. During these examinations aluminum cladding thickness and oxide thickness on the cladding were also measured. Special application techniques were required because of the high-radiation levels (approx. 10 5 R/h at 30 cm) present and the relatively large temperature gradients that occurred on the surface of the control rods. The techniques used to perform the eddy-current inspections and the methods used to reduce the associated data are described

The influence Of On Power Target Loading At RSG-GAS Reactor

International Nuclear Information System (INIS)

Kusno; Pardi

2001-01-01

One of utilization purposes of the RSG-GAS reactor is to produce radioisotope. The target insertion or withdrawal on the irradiation position D-6 and E-7 in the core can be done manually, during reactor operation or shut down condition. The problem arises when the loading is under operation mode, because of flow and reactivity change the reactivity. The behavior of operation parameter changes due to target handling will be investigated in this paper. The behavior will be applied as a guidance of the operating group in handling the irradiation target

Balancing the risks: the NRU reactor and the isotope crisis in Canada

International Nuclear Information System (INIS)

Morrison, B.; Meneley, D.

2008-01-01

The extended shutdown of the NRU reactor at Chalk River at the end of 2007 caused a critical shortage of medical radioisotopes in Canada and the world, led to a unique meeting of Canada's Parliament to pass emergency legislation, and cost the President of the Canadian Nuclear Safety Commission her job. This paper, based on the public record, reviews these events from the perspective of the balance of risk between the safety of the NRU reactor and the impact of a shortage of isotopes. This leads to important questions about the mandate, independence and flexibility of the nuclear regulator, relations between the regulator, the government, and the licensee, and the government's overall management of risks. We argue that the government approaches individual risks in isolation and needs a mechanism to deal with multiple risks. (author)

Production of Cs and Fr isotopes from a high-density UC targets with different grain dimensions

Energy Technology Data Exchange (ETDEWEB)

Panteleev, V.N.; Barzakh, A.E.; Fedorov, D.V.; Ivanov, V.S.; Mezilev, K.A.; Molkanov, P.L.; Moroz, F.V.; Orlov, S.Yu.; Volkov, Yu.M. [Petersburg Nuclear Physics Institute RAS, Gatchina (Russian Federation); Alyakrinskiy, O.; Barbui, M.; Stroe, L.; Tecchio, L.B.; Tonezzer, M. [Laboratori Nationali di Legnaro, Legnaro (Padova) (Italy); Lhersonneau, G. [GANIL, Caen Cedex 5 (France)

2009-12-15

A UC target material of 11.3{+-}0.5 g/cm{sup 3} uranium density with the grain size of 20 and 5{mu}m manufactured in a form of pills by the method of powder metallurgy has been tested on-line within the temperature range of 1800-2100 C. The mass of uranium exposed to the beam was 4-7g. The yields and release rates of Cs and Fr isotopes produced by fission and spallation reactions of {sup 238}U by 1GeV protons have been measured. The yields of Cs and Fr isotopes obtained from the tested target materials have been compared, including yields of very short-lived Fr isotopes with half-lives down to 1ms. Temperature-resistant materials (porous graphite and tantalum foil) have been used for the internal-container construction, which holds the UC target pills inside a tungsten external container heated by the resistant heating. The fastest release and the highest efficiency for short-lived isotopes have been obtained for the targets with the internal container manufactured from the tantalum foil. Results of on-line tests of a big mass target (730g of 5{mu}m grain UC target material) have been discussed. (orig.)

Production of Cs and Fr isotopes from a high-density UC targets with different grain dimensions

International Nuclear Information System (INIS)

Panteleev, V.N.; Barzakh, A.E.; Fedorov, D.V.; Ivanov, V.S.; Mezilev, K.A.; Molkanov, P.L.; Moroz, F.V.; Orlov, S.Yu.; Volkov, Yu.M.; Alyakrinskiy, O.; Barbui, M.; Stroe, L.; Tecchio, L.B.; Tonezzer, M.; Lhersonneau, G.

2009-01-01

A UC target material of 11.3±0.5 g/cm 3 uranium density with the grain size of 20 and 5μm manufactured in a form of pills by the method of powder metallurgy has been tested on-line within the temperature range of 1800-2100 C. The mass of uranium exposed to the beam was 4-7g. The yields and release rates of Cs and Fr isotopes produced by fission and spallation reactions of 238 U by 1GeV protons have been measured. The yields of Cs and Fr isotopes obtained from the tested target materials have been compared, including yields of very short-lived Fr isotopes with half-lives down to 1ms. Temperature-resistant materials (porous graphite and tantalum foil) have been used for the internal-container construction, which holds the UC target pills inside a tungsten external container heated by the resistant heating. The fastest release and the highest efficiency for short-lived isotopes have been obtained for the targets with the internal container manufactured from the tantalum foil. Results of on-line tests of a big mass target (730g of 5μm grain UC target material) have been discussed. (orig.)

High Flux Isotope Reactor system RELAP5 input model

International Nuclear Information System (INIS)

Morris, D.G.; Wendel, M.W.

1993-01-01

A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model

Status and future plan of KUR-ISOL for new isotope search

Energy Technology Data Exchange (ETDEWEB)

Taniguchi, Akihiro [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

1997-07-01

He gas-jet type ISOL (KUR-ISOL: Kyoto University Reactor-Isotope Separator On-Line) was set up in Kyoto University Reactor in 1979. The thema of researches using KUR-ISOL are investigation of new isotope elements, study of nuclear structure of neutron-enrich nucleide in the neighborhood of 150 of mass number, development of unstable nuclide production unit and research of physical properties using unstable nuclide as probe. By KUR-ISOL, four kinds of new isotopes such as {sup 156}Pm, {sup 155}Nd, {sup 154}Pr and {sup 152}Ce and {beta}-decay of {sup 150}La had been identified. {beta}-decay of {sup 150}La as a sample of them was explained in this report. Today, the experiment of {sup 153}Pr, {sup 152}Pr and {sup 149,150}Ce are proceeding. For future plans, new beam line and new target used transuranic elements will be developed. (S.Y.)

Simultaneous nuclear data target accuracy study for innovative fast reactors

International Nuclear Information System (INIS)

Aliberti, G.; Palmiotti, G.; Salvatores, M.

2007-01-01

The present paper summarizes the major outcomes of a study conducted within a Nuclear Energy Agency Working Party on Evaluation Cooperation (NEA WPEC) initiative aiming to investigate data needs for future innovative nuclear systems, to quantify them and to propose a strategy to meet them. Within the NEA WPEC Subgroup 26 an uncertainty assessment has been carried out using covariance data recently processed by joint efforts of several US and European Labs. In general, the uncertainty analysis shows that for the wide selection of fast reactor concepts considered, the present integral parameters uncertainties resulting from the assumed uncertainties on nuclear data are probably acceptable in the early phases of design feasibility studies. However, in the successive phase of preliminary conceptual designs and in later design phases of selected reactor and fuel cycle concepts, there will be the need for improved data and methods, in order to reduce margins, both for economic and safety reasons. It is then important to define as soon as possible priority issues, i.e. which are the nuclear data (isotope, reaction type, energy range) that need improvement, in order to quantify target accuracies and to select a strategy to meet the requirements needed (e.g. by some selected new differential measurements and by the use of integral experiments). In this context one should account for the wide range of high accuracy integral experiments already performed and available in national or, better, international data basis, in order to indicate new integral experiments that will be needed to account for new requirements due to innovative design features, and to provide the necessary full integral data base to be used for validation of the design simulation tools.

The development of a small inherently safe homogeneous reactor for the production of medical isotopes

Energy Technology Data Exchange (ETDEWEB)

Carlin, G.E.; Bonin, H.W., E-mail: george.carlin@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

2013-07-01

The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. New interest has been found in the use of liquid fueled nuclear reactors to produce these isotopes due to the ease of fuel processing and ability to efficiently use LEU as the fuel source. A version of this reactor is being developed at the Royal Military College of Canada to act as a successor to the SLOWPOKE-2 platform. The thermal hydraulic and transient characteristics of a 20 kWt version are being studied to verify inherent safety abilities. (author)

Production Situation and Technology Prospect of Medical Isotopes

Directory of Open Access Journals (Sweden)

GAO Feng;LIN Li;LIU Yu-hao;MA Xing-jun

2016-10-01

Full Text Available The isotope production technology was overviewed, including traditional and newest technology. The current situation of medical isotope production was introduced. The problems faced by isotope supply and demand were analyzed. The future development trend of medical isotopes and technology prospect were put forward. As the most populous country, nuclear medicine develops rapidly, however, domestic isotope mainly relies on imports. The highly productive and relatively safe MIPR is expected to be an effective way to breakthrough the bottleneck of the development of nuclear medicine. Traditional isotope production technologies with reactor can be improved. It's urgent to research and promote new isotope production technologies with reactor. Those technologies which do not depend on reactor will have a bright market prospects.

Rare isotope accelerator—conceptual design of target areas

Science.gov (United States)

Bollen, Georg; Baek, Inseok; Blideanu, Valentin; Lawton, Don; Mantica, Paul F.; Morrissey, David J.; Ronningen, Reginald M.; Sherrill, Bradley S.; Zeller, Albert; Beene, James R.; Burgess, Tom; Carter, Kenneth; Carrol, Adam; Conner, David; Gabriel, Tony; Mansur, Louis; Remec, Igor; Rennich, Mark; Stracener, Dan; Wendel, Mark; Ahle, Larry; Boles, Jason; Reyes, Susana; Stein, Werner; Heilbronn, Lawrence

2006-06-01

The planned rare isotope accelerator facility RIA in the US would become the most powerful radioactive beam facility in the world. RIA's driver accelerator will be a device capable of providing beams from protons to uranium at energies of at least 400 MeV per nucleon, with beam power up to 400 kW. Radioactive beam production relies on both the in-flight separation of fast beam fragments and on the ISOL technique. In both cases the high beam power poses major challenges for target technology and handling and on the design of the beam production areas. This paper will give a brief overview of RIA and discuss aspects of ongoing conceptual design work for the RIA target areas.

Rare isotope accelerator - conceptual design of target areas

International Nuclear Information System (INIS)

Bollen, Georg; Baek, Inseok; Blideanu, Valentin; Lawton, Don; Mantica, Paul F.; Morrissey, David J.; Ronningen, Reginald M.; Sherrill, Bradley S.; Zeller, Albert; Beene, James R; Burgess, Tom; Carter, Kenneth; Carrol, Adam; Conner, David; Gabriel, Tony A; Mansur, Louis K; Remec, Igor; Rennich, Mark J; Stracener, Daniel W; Wendel, Mark W; Ahle, Larry; Boles, Jason; Reyes, Susana; Stein, Werner; Heilbronn, Lawrence

2006-01-01

The planned rare isotope accelerator facility RIA in the US would become the most powerful radioactive beam facility in the world. RIA's driver accelerator will be a device capable of providing beams from protons to uranium at energies of at least 400MeV per nucleon, with beam power up to 400 kW. Radioactive beam production relies on both the in-flight separation of fast beam fragments and on the ISOL technique. In both cases the high beam power poses major challenges for target technology and handling and on the design of the beam production areas. This paper will give a brief overview of RIA and discuss aspects of ongoing conceptual design work for the RIA target areas

Development of a simplified methodology for the isotopic determination of fuel spent in Light Water Reactors; Desarrollo de una metodologia simplificada para la determinacion isotopica del combustible gastado en reactores de agua ligera

Energy Technology Data Exchange (ETDEWEB)

Hernandez N, H.; Francois L, J.L. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: hermilo@lairn.fi-b.unam.mx

2005-07-01

The present work presents a simplified methodology to quantify the isotopic content of the spent fuel of light water reactors; their application is it specific to the Laguna Verde Nucleo electric Central by means of a balance cycle of 18 months. The methodology is divided in two parts: the first one consists on the development of a model of a simplified cell, for the isotopic quantification of the irradiated fuel. With this model the burnt one is simulated 48,000 MWD/TU of the fuel in the core of the reactor, taking like base one fuel assemble type 10x10 and using a two-dimensional simulator for a fuel cell of a light water reactor (CPM-3). The second part of the methodology is based on the creation from an isotopic decay model through an algorithm in C++ (decay) to evaluate the amount, by decay of the radionuclides, after having been irradiated the fuel until the time in which the reprocessing is made. Finally the method used for the quantification of the kilograms of uranium and obtained plutonium of a normalized quantity (1000 kg) of fuel irradiated in a reactor is presented. These results will allow later on to make analysis of the final disposition of the irradiated fuel. (Author)

Evaluation of HFIR [High Flux Isotope Reactor] pressure-vessel integrity considering radiation embrittlement

International Nuclear Information System (INIS)

Cheverton, R.D.; Merkle, J.G.; Nanstad, R.K.

1988-04-01

The High Flux Isotope Reactor (HFIR) pressure vessel has been in service for 20 years, and during this time, radiation damage was monitored with a vessel-material surveillance program. In mid-November 1986, data from this program indicated that the radiation-induced reduction in fracture toughness was greater than expected. As a result, a reevaluation of vessel integrity was undertaken. Updated methods of fracture-mechanics analysis were applied, and an accelerated irradiations program was conducted using the Oak Ridge Research Reactor. Results of these efforts indicate that (1) the vessel life can be extended 10 years if the reactor power level is reduced 15% and if the vessel is subjected to a hydrostatic proof test each year; (2) during the 10-year life extension, significant radiation damage will be limited to a rather small area around the beam tubes; and (3) the greater-than-expected damage rate is the result of the very low neutron flux in the HFIR vessel relative to that in samples of material irradiated in materials-testing reactors (a factor of ∼10 4 less), that is, a rate effect

Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

International Nuclear Information System (INIS)

Nagy, B.; Rigali, M.J.; Davis, D.W.; Parnell, J.

1991-01-01

Some of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the re-solidified, graphitic bituminous organics at Oklo thus enhanced radionuclide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lanthanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pre-treated nuclear waste warrants further investigation. (author)

Fast reactor development strategy targets study in China

International Nuclear Information System (INIS)

Xu Mi

2008-01-01

China is a big developing Country who needs a huge energy resources and a rapid growing rate. Considering energy resources limited and environment issues it is sure that the nuclear energy will be becoming one of the main energy resources. The Government has decided to develop the nuclear power capacity to 40 GW in 2020. It is envisaged that it will reach to 240 GW in 2050. It is stimulate us to consider conscientiously the development of the fast breeder reactor's and related closed nuclear fuel cycle by the limitation of Uranium resources and uncertainties of international Uranium market. Followings are the proposed strategic targets of fast reactor development in China. (1) To realize the operation of commercial fast breeder reactors with an unit size of 800-900 MWe and one site-multi reactors in 2030. (2) To develop the nuclear power capacity to 240 GW in 2050. (3) To replace step by step the fossil fuel utilization in large scale by nuclear energy beyond 2050. (authors)

Rare Isotope Accelerator - Conceptual Design of Target Areas

Energy Technology Data Exchange (ETDEWEB)

Bollen, Georg [Michigan State University, East Lansing; Baek, Inseok [Michigan State University, East Lansing; Blideanu, Valentin [CEA, Saclay, France; Lawton, Don [Michigan State University, East Lansing; Mantica, Paul F. [Michigan State University, East Lansing; Morrissey, David J. [Michigan State University, East Lansing; Ronningen, Reginald M. [Michigan State University, East Lansing; Sherrill, Bradley S. [Michigan State University, East Lansing; Zeller, Albert [Michigan State University, East Lansing; Beene, James R [ORNL; Burgess, Tom [Oak Ridge National Laboratory (ORNL); Carter, Kenneth [Oak Ridge National Laboratory (ORNL); Carrol, Adam [Oak Ridge National Laboratory (ORNL); Conner, David [ORNL; Gabriel, Tony A [ORNL; Mansur, Louis K [ORNL; Remec, Igor [ORNL; Rennich, Mark J [ORNL; Stracener, Daniel W [ORNL; Wendel, Mark W [ORNL; Ahle, Larry [Lawrence Livermore National Laboratory (LLNL); Boles, Jason [Lawrence Livermore National Laboratory (LLNL); Reyes, Susana [Lawrence Livermore National Laboratory (LLNL); Stein, Werner [Lawrence Livermore National Laboratory (LLNL); Heilbronn, Lawrence [Lawrence Berkeley National Laboratory (LBNL)

2006-01-01

The planned rare isotope accelerator facility RIA in the US would become the most powerful radioactive beam facility in the world. RIA s driver accelerator will be a device capable of providing beams from protons to uranium at energies of at least 400MeV per nucleon, with beam power up to 400 kW. Radioactive beam production relies on both the in-flight separation of fast beam fragments and on the ISOL technique. In both cases the high beam power poses major challenges for target technology and handling and on the design of the beam production areas. This paper will give a brief overview of RIA and discuss aspects of ongoing conceptual design work for the RIA target areas.

Rare isotope accelerator-conceptual design of target areas

Energy Technology Data Exchange (ETDEWEB)

Bollen, Georg [National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824 (United States)]. E-mail: bollen@nscl.msu.edu; Baek, Inseok; Blideanu, Valentin; Lawton, Don; Mantica, Paul F.; Morrissey, David J.; Ronningen, Reginald M.; Sherrill, Bradley S.; Zeller, Albert [National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824 (United States); Beene, James R.; Burgess, Tom; Carter, Kenneth; Carrol, Adam; Conner, David; Gabriel, Tony; Mansur, Louis; Remec, Igor; Rennich, Mark; Stracener, Dan; Wendel, Mark [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States); Ahle, Larry; Boles, Jason; Reyes, Susana; Stein, Werner [Lawrence Livermore Laboratory, Livermore, CA 94550 (United States); Heilbronn, Lawrence [Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States)

2006-06-23

The planned rare isotope accelerator facility RIA in the US would become the most powerful radioactive beam facility in the world. RIA's driver accelerator will be a device capable of providing beams from protons to uranium at energies of at least 400 MeV per nucleon, with beam power up to 400 kW. Radioactive beam production relies on both the in-flight separation of fast beam fragments and on the ISOL technique. In both cases the high beam power poses major challenges for target technology and handling and on the design of the beam production areas. This paper will give a brief overview of RIA and discuss aspects of ongoing conceptual design work for the RIA target areas.

Discrimination of source reactor type by multivariate statistical analysis of uranium and plutonium isotopic concentrations in unknown irradiated nuclear fuel material.

Science.gov (United States)

Robel, Martin; Kristo, Michael J

2008-11-01

The problem of identifying the provenance of unknown nuclear material in the environment by multivariate statistical analysis of its uranium and/or plutonium isotopic composition is considered. Such material can be introduced into the environment as a result of nuclear accidents, inadvertent processing losses, illegal dumping of waste, or deliberate trafficking in nuclear materials. Various combinations of reactor type and fuel composition were analyzed using Principal Components Analysis (PCA) and Partial Least Squares Discriminant Analysis (PLSDA) of the concentrations of nine U and Pu isotopes in fuel as a function of burnup. Real-world variation in the concentrations of (234)U and (236)U in the fresh (unirradiated) fuel was incorporated. The U and Pu were also analyzed separately, with results that suggest that, even after reprocessing or environmental fractionation, Pu isotopes can be used to determine both the source reactor type and the initial fuel composition with good discrimination.

Seismic, high wind, tornado, and probabilistic risk assessment of the high flux isotope reactor

International Nuclear Information System (INIS)

Harris, S.P.; Hashimoto, P.S.; Dizon, J.O.; Hashimoto, P.S.

1989-01-01

Natural phenomena analyses were performed on the High Flux Isotope Reactor (HFIR). Deterministic and probabilistic evaluations were made to determine the risks resulting from earthquakes, high winds, and tornadoes. Analytic methods in conjunction with field evaluations and an earthquake experience data base evaluation methods were used to provide more realistic results in a shorter amount of time. Plant modifications completed in preparation for HFIR restart and potential future enhancements are discussed

Incineration of actinide targets in a pressurized water reactor spin project

International Nuclear Information System (INIS)

Puill, A.; Bergeron, J.

1993-01-01

The ability of Pressurized Water Reactors (PWR) with uranium fuel to limit the inventory growth of minor actinides (237 neptunium, and americium) produced by the French nuclear powerplants is studied. Targets containing an actinide oxide mixed to an inert matrix are loaded in some reactors. After being irradiated along with the fuel, the target is specially reprocessed. The remaining actinide and the plutonium which is produced, added to fresh actinide, are recycled in new targets. The radiotoxicity balance, with and without incineration, is examined considering that only the losses coming from the target reprocessing treated as waste. A scenario arbitrarily based on 18 years of operation results in a reduction of the radiotoxicity of the waste by a factor between 10 and 20, depending on the actinide considered. 6 refs., 6 figs., 6 tabs

Thermal analysis of LEU modified Cintichem target irradiated in TRIGA reactor

International Nuclear Information System (INIS)

Catana, A; Toma, C.

2009-01-01

Actions conceived during last years at international level for conversion of Molybdenum fabrication process from HEU to LEU targets utilization created opportunities for INR to get access to information and participating to international discussions under IAEA auspices. Concrete steps for developing fission Molybdenum technology were facilitated. Institute of Nuclear Research bringing together a number of conditions like suitable irradiation possibilities, direct communication between reactor and hot cell facility, handling capacity of high radioactive sources, and simultaneously the existence of an expanding internal market, decided to undertake the necessary steps in order to produce fission molybdenum. Over the course of last years of efforts in this direction we developed the steps for fission Molybdenum technology development based on modified Cintichem process in accordance with the Argonne National Laboratory proved methodology. Progress made by INR to heat transfer computations of annular target using is presented. An advanced thermal-hydraulic analysis was performed to estimate the heat removal capability for an enriched uranium (LEU) foil annular target irradiated in TRIGA reactor core. As a result, the present analysis provides an upper limit estimate of the LEU-foil and external target surface temperatures during irradiation in TRIGA 14 MW reactor. (authors)

Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

International Nuclear Information System (INIS)

Bultman, J.

1992-04-01

Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

Implementation of isotope correlation technique for safeguards

International Nuclear Information System (INIS)

Persiani, P.J.; Bucher, R.G.

1989-01-01

The isotopic correlation technique (ICT) is based on the fundamental physics principle that the isotopic compositions of nuclear material in the fuel cycle systems contain information regarding the design and history of nuclear material flow from fuel fabrication, reactor operation, and through input to the reprocessing plant. Isotopic Correlation in conjunction with the gravimetric (or Pu/U) method for mass determination can be developed to provide an independent in-field verification of the reprocessing input accountancy at the dissolver and/or accountancy stage of the reprocessing plant. The Argonne National Laboratory program in isotope correlation techniques is based on three-dimensional reactor physics calculations of characteristic geometries/composition in each reactor class. 10 refs., 1 fig., 3 tabs

Seismic, high wind, tornado, and probabilistic risk assessments of the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Harris, S.P.; Stover, R.L.; Hashimoto, P.S.; Dizon, J.O.

1989-01-01

Natural phenomena analyses were performed on the High Flux Isotope Reactor (HFIR) Deterministic and probabilistic evaluations were made to determine the risks resulting from earthquakes, high winds, and tornadoes. Analytic methods in conjunction with field evaluations and an earthquake experience data base evaluation methods were used to provide more realistic results in a shorter amount of time. Plant modifications completed in preparation for HFIR restart and potential future enhancements are discussed. 5 figs

Emergency diesel generator reliability analysis high flux isotope reactor

International Nuclear Information System (INIS)

Merryman, L.; Christie, B.

1993-01-01

A program to apply some of the techniques of reliability engineering to the High Flux Isotope Reactor (HFIR) was started on August 8, 1992. Part of the program was to track the conditional probabilities of the emergency diesel generators responding to a valid demand. This was done to determine if the performance of the emergency diesel generators (which are more than 25 years old) has deteriorated. The conditional probabilities of the diesel generators were computed and trended for the period from May 1990 to December 1992. The calculations indicate that the performance of the emergency diesel generators has not deteriorated in recent years, i.e., the conditional probabilities of the emergency diesel generators have been fairly stable over the last few years. This information will be one factor than may be considered in the decision to replace the emergency diesel generators

Project requirements for reconstruction of the RA reactor ventilation system, Task 2.8. Measurement of radioactive iodine and other isotopes contents in the gas system of the RA reactor, Annex of the task

International Nuclear Information System (INIS)

Vujisic, Lj. et al

1981-01-01

This report is a supplement to the task 2.8. When planning and constructing the ventilation system, it was found that it is necessary to perform additional experiments during RA reactor operation at 2 MW power level for a longer period. In addition to the helium system, the potential source of radioactive pollutants is the space below the upper water shielding of the reactor. All the experimental and fuel channels are ending in this space. During repair and fuel exchange radioactivity can be released in this space. For that reason this space is important when planing and designing the filtration system for incidental conditions or planned dehermetisation of the reactor. The third point where radioactive isotope identification was done, was the entrance into the chimney during steady state operation and planned dehermetisation of the reactor. The following samples were measured: gas system during reactor operation at 2 MW power; entrance into the chimney during last 48 hours of reactor operation at 2 MW power; sample on the platform under the upper water shield with the opened fuel channel after the reactor shutdown; and simultaneously with the latter, measurement at the entrance to the chimney. This annex contains the list of identified radioactive isotopes, volatile and gaseous as well as concentration of volatile 131 I on the adsorbents [sr

Identification of Thioredoxin Target Disulfides Using Isotope-Coded Affinity Tags

DEFF Research Database (Denmark)

Hägglund, Per; Bunkenborg, Jakob; Maeda, Kenji

2014-01-01

Thioredoxins (Trx) are small redox proteins that reduce disulfide bonds in various target proteins and maintain cellular thiol redox control. Here, a thiol-specific labeling and affinity enrichment approach for identification and relative quantification of Trx target disulfides in complex protein...... reduction is determined by LC-MS/MS-based quantification of tryptic peptides labeled with "light" (12C) and "heavy" (13C) ICAT reagents. The methodology can be adapted to monitor the effect of different reductants or oxidants on the redox status of thiol/disulfide proteomes in biological systems....... extracts is described. The procedure utilizes the isotope-coded affinity tag (ICAT) reagents containing a thiol reactive iodoacetamide group and a biotin affinity tag to target peptides containing reduced cysteine residues. The identification of substrates for Trx and the extent of target disulfide...

Extraction of gadolinium from high flux isotope reactor control plates

International Nuclear Information System (INIS)

Kohring, M.W.

1987-04-01

Gadolinium-153 is an important radioisotope used in the diagnosis of various bone disorders. Recent medical and technical developments in the detection and cure of osteoporosis, a bone disease affecting an estimated 50 million people, have greatly increased the demand for this isotope. The Oak Ridge National Laboratory (ORNL) has produced 153 Gd since 1980 primarily through the irradiation of a natural europium-oxide powder followed by the chemical separation of the gadolinium fraction from the europium material. Due to the higher demand for 153 Gd, an alternative production method to supplement this process has been investigated. This process involves the extraction of gadolinium from the europium-bearing region of highly radioactive, spent control plates used at the High Flux Isotope Reactor (HFIR) with a subsequent re-irradiation of the extracted material for the production of the 153 Gd. Based on the results of experimental and calculational analyses, up to 25 grams of valuable gadolinium (≥60% enriched in 152 Gd) resides in the europium-bearing region of the HFIR control components of which 70% is recoverable. At a specific activity yield of 40 curies of 153 Gd for each gram of gadolinium re-irradiated, 700 one-curie sources can be produced from each control plate assayed

Uses of stable isotopes

International Nuclear Information System (INIS)

Axente, Damian

1998-01-01

The most important fields of stable isotope use with examples are presented. These are: 1. Isotope dilution analysis: trace analysis, measurements of volumes and masses; 2. Stable isotopes as tracers: transport phenomena, environmental studies, agricultural research, authentication of products and objects, archaeometry, studies of reaction mechanisms, structure and function determination of complex biological entities, studies of metabolism, breath test for diagnostic; 3. Isotope equilibrium effects: measurement of equilibrium effects, investigation of equilibrium conditions, mechanism of drug action, study of natural processes, water cycle, temperature measurements; 4. Stable isotope for advanced nuclear reactors: uranium nitride with 15 N as nuclear fuel, 157 Gd for reactor control. In spite of some difficulties of stable isotope use, particularly related to the analytical techniques, which are slow and expensive, the number of papers reporting on this subject is steadily growing as well as the number of scientific meetings organized by International Isotope Section and IAEA, Gordon Conferences, and regional meeting in Germany, France, etc. Stable isotope application development on large scale is determined by improving their production technologies as well as those of labeled compound and the analytical techniques. (author)

Method of eliminating gaseous hydrogen isotopes

International Nuclear Information System (INIS)

Nagakura, Masaaki; Imaizumi, Hideki; Suemori, Nobuo; Aizawa, Takashi; Naito, Taisei.

1983-01-01

Purpose: To prevent external diffusion of gaseous hydrogen isotopes such as tritium or the like upon occurrence of tritium leakage accident in a thermonuclear reactor by recovering to eliminate the isotopes rapidly and with safety. Method: Gases at the region of a reactor container where hydrogen isotopes might leak are sucked by a recycing pump, dehumidified in a dehumidifier and then recycled from a preheater through a catalytic oxidation reactor to a water absorption tower. In this structure, the dehumidifier is disposed at the upstream of the catalytic oxidation reactor to reduce the water content of the gases to be processed, whereby the eliminating efficiency for the gases to be processed can be maintained well even when the oxidation reactor is operated at a low temperature condition near the ambient temperature. This method is based on the fact that the oxidating reactivity of the catalyst can be improved significantly by eliminating the water content in the gases to be processed. (Yoshino, Y.)

The method of life extension for the High Flux Isotope Reactor vessel

International Nuclear Information System (INIS)

Chang, Shib-Jung.

1995-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A hydrostatic pressure test (hydrotest) is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the reactor life from the safe hydrostatic pressure. The life extension calculation provides the following information on the remaining life of the reactor as a function of the nil ductility temperature increase: the probability of vessel fracture due to hydrotest vs vessel life at several hydrotest pressures; the hydrotest time interval vs the uncertainty of the nil ductility temperature increase rate; and the hydrotest pressure vs the uncertainty of the nil ductility temperature increase rate. It is understood that the use of a complete range of uncertainties of the nil ductility temperature increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85 degree F

Particle transfer spectroscopy using radioactive targets

CERN Document Server

Naumann, R A

1976-01-01

The practicality of general use of radioactive targets to study nuclei off the stability line by transfer spectroscopy is examined. Some advantages of this spectroscopic technique are illustrated with recent results from (p, t) and (t, p) stable target studies of negative parity core-coupled states systematically occurring in 4 adjacent odd silver isotopes. Preliminary results from the study of the /sup 205/Pb (t, p)/sup 207/Pb reaction using reactor produced 3*10/sup 7/ year lead 205 are given. (3 refs).

Temperature and void reactivity coefficient calculations for the high flux isotope reactor safety analysis report

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Williams, L.R.

1994-07-01

This report provides documentation of a series of calculations performed in 1991 in order to provide input for the High Flux Isotope Reactor Safety Analysis Report. In particular, temperature and void reactivity coefficients were calculated for beginning-of-life, end-of-life, and xenon equilibrium (29 h) conditions. Much of the data used to prepare the computer models for these calculations was derived from the original HFIR nuclear design study

Nuclear Data Target Accuracy Requirements For MA Burners

International Nuclear Information System (INIS)

Palmiotti, G.; Salvatores, M.

2011-01-01

A nuclear data target accuracy assessment has been carried out for two types of transmuters: a critical sodium fast reactor(SFR) and an accelerator driven system (ADMAB). Results are provided for a 7 group energy structure. Considerations about fuel cycle parameters uncertainties illustrate their dependence from the isotope final densities at end of cycle.

Neutron capture reactions on Lu isotopes at DANCE

CERN Document Server

Roig, O

2010-01-01

The DANCE (Detector for Advanced Neutron Capture Experiments) array located at the Los Alamos national laboratory has been used to obtain the neutron capture cross sections for 175Lu and 176Lu with neutron energies from thermal up to 100 keV. Both isotopes are of current interest for the nucleosynthesis s-process in astrophysics and for applications as in reactor physics or in nuclear medicine. Three targets were used to perform these measurements. One was natLu foil and the other two were isotope-enriched targets of 175Lu and 176Lu. The cross sections are obtained for now through a precise neutron flux determination and a normalization at the thermal neutron cross section value. A comparison with the recent experimental data and the evaluated data of ENDF/B-VII.0 will be presented. In addition, resonances parameters and spin assignments for some resonances will be featured.

Minor actinides incineration by loading moderated targets in fast reactor

International Nuclear Information System (INIS)

Wu Hongchun; Sato, Daisuke; Takeda, Toshikazu

2000-01-01

The effect of hydrogen concentration and loaded mass of minor actinides (MAs) in the target on the core performance and MAs transmutation rate was analyzed in this paper. An optimum core was proposed which has 96 MAs target assemblies of which MAs fuel pins per assembly is 38 with the composition ratio U/MA/Zr/H of 1/4/10/50. This optimized core offers good core performance and can transmute MAs very effectively, the transmutation rate was about 67% (939 kg) and the incinerate (transmute by fission) rate was about 35% (489 kg) through 3 years of reactor operation. It is about 2-3 times larger than current transmutation method that MAs are loaded homogeneously in the PWR and fast reactor core. (author)

High Flux Isotope Reactor (HFIR)

Data.gov (United States)

Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

Accumulation of the radionuclides in a target irradiated in the reactor of tajoura nuclear research center

International Nuclear Information System (INIS)

Abdunnobi, A.R.; Arebi, B.

1998-01-01

One of the main stages of radionuclides production in reactor is the distinguishing of a regime on target irradiation in order to acquire the sufficient activity and the purity of radioisotope required. The authors have derived formula for calculating radionuclidic accumulation on a target irradiated in the reactor operating 10 hours per day, 4 days a week during 4 weeks. The results of I-131 and other radionuclide accumulation are illustrated by a tellurium target irradiation in the reactor operating continuously or with interruptions

Neutron scattering at the high-flux isotope reactor

International Nuclear Information System (INIS)

Cable, J.W. Chakoumakos, B.C.; Dai, P.

1995-01-01

The title facilities offer the brightest source of neutrons in the national user program. Neutron scattering experiments probe the structure and dynamics of materials in unique and complementary ways as compared to x-ray scattering methods and provide fundamental data on materials of interest to solid state physicists, chemists, biologists, polymer scientists, colloid scientists, mineralogists, and metallurgists. Instrumentation at the High- Flux Isotope Reactor includes triple-axis spectrometers for inelastic scattering experiments, a single-crystal four diffractometer for crystal structural studies, a high-resolution powder diffractometer for nuclear and magnetic structure studies, a wide-angle diffractometer for dynamic powder studies and measurements of diffuse scattering in crystals, a small-angle neutron scattering (SANS) instrument used primarily to study structure-function relationships in polymers and biological macromolecules, a neutron reflectometer for studies of surface and thin-film structures, and residual stress instrumentation for determining macro- and micro-stresses in structural metals and ceramics. Research highlights of these areas will illustrate the current state of neutron science to study the physical properties of materials

Production of americium isotopes in France

International Nuclear Information System (INIS)

Koehly, G.; Bourges, J.; Madic, C.; Nguyen, T.H.; Lecomte, M.

1984-12-01

The program of productions of americium 241 and 243 isotopes is based respectively on the retreatment of aged plutonium alloys or plutonium dioxide and on the treatment of plutonium targets irradiated either in CELESTIN reactors for Pu-Al alloys or OSIRIS reactor for plutonium 242 dioxide. All the operations, including americium final purifications, are carried out in hot cells equipped with remote manipulators. The chemical processes are based on the use of extraction chromatography with hydrophobic SiO 2 impregnated with extracting agents. Plutonium targets and aged plutonium alloys are dissolved in nitric acid using conventional techniques while plutonium dioxide dissolutions are performed routine at 300 grams scale with electrogenerated silver II in 4M HNO 3 at room temperature. The separation between plutonium and americium is performed by extraction of Pu(IV) either on TBP/SiO 2 or TOAHNO 3 /SiO 2 column. Americium recovery from waste streams rid of plutonium is realized by chromatographic extraction of Am(III) using mainly TBP and episodically DHDECMP as extractant. The final purification of both americium isotopes uses the selective extraction of Am(VI) on HDDiBMP/SiO 2 column at 60 grams scale. Using the overall process a total amount of 1000 grams of americium 241 and 100 grams of americium 243 has been produced nowadays and the AmO 2 final product indicates a purity better than 98.5%

Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

International Nuclear Information System (INIS)

Ashoub, N.; Saleh, H.

1995-01-01

The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

Determination of Light Water Reactor Fuel Burnup with the Isotope Ratio Method

International Nuclear Information System (INIS)

Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

2007-01-01

For the current project to demonstrate that isotope ratio measurements can be extended to zirconium alloys used in LWR fuel assemblies we report new analyses on irradiated samples obtained from a reactor. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers new measurements done on irradiated and unirradiated zirconium alloys, Unirradiated zircaloy samples serve as reference samples and indicate starting values or natural values for the Ti isotope ratio measured. New measurements of irradiated samples include results for 3 samples provided by AREVA. New results indicate: 1. Titanium isotope ratios were measured again in unirradiated samples to obtain reference or starting values at the same time irradiated samples were analyzed. In particular, 49Ti/48Ti ratios were indistinguishably close to values determined several months earlier and to expected natural values. 2. 49Ti/48Ti ratios were measured in 3 irradiated samples thus far, and demonstrate marked departures from natural or initial ratios, well beyond analytical uncertainty, and the ratios vary with reported fluence values. The irradiated samples appear to have significant surface contamination or radiation damage which required more time for SIMS analyses. 3. Other activated impurity elements still limit the sample size for SIMS analysis of irradiated samples. The sub-samples chosen for SIMS analysis, although smaller than optimal, were still analyzed successfully without violating the conditions of the applicable Radiological Work Permit

Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Cook, David Howard

2009-01-01

A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

Energy Technology Data Exchange (ETDEWEB)

Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

2009-03-01

This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

Inelastic scattering of Ni and Zn isotopes off a proton target

Energy Technology Data Exchange (ETDEWEB)

Cortes Sua, Martha Liliana

2016-07-18

Inelastic proton scattering of {sup 70,72,74}Ni and {sup 76,78,80}Zn was performed at the RIBF facility of the RIKEN Nishina Center, Japan, as part of the first SEASTAR campaign. Radioactive isotopes were produced by the in-flight fission of a beam of {sup 238}U ions incident on a 3 mm thick Beryllium target. After production, neutron-rich radioactive isotopes were selected and identified on an event-by-event basis using the BigRIPS separator. Selected isotopes of interest were focused onto the liquid hydrogen target of the MINOS device and γ-rays from inelastic (p,p{sup '}) reactions were detected with the DALI2 array, consisting of 186 NaI crystals. Outgoing beam-like particles were identified using the ZeroDegree spectrometer. γ-rays produced in the reaction were Doppler corrected and the first 2{sup +} and 4{sup +} states in all the isotopes were identified. Detailed data analysis was performed including the implementation of algorithms that discriminate events where more than one particle was present. Using detailed Geant4 simulations, exclusive cross-sections for inelastic proton scattering were obtained. Deformation lengths were deduced from the experimental cross-sections using the coupled-channel calculation code ECIS-97. The deformations lengths of {sup 72,74}Ni and {sup 76,80}Zn were found to be fairly constant at a value of 0.8(2) fm, suggesting similar vibrational amplitudes, while the isomeric presence in the secondary beams of {sup 70}Ni and {sup 78}Zn allowed only lower limits for those two isotopes. By combining the deformation lengths with the known B(E2;0{sup +}{sub gs}→2{sup +}{sub 1}) values, the neutron-to-proton matrix element ratios, M{sub n}/M{sub p}, were obtained. A clear indication of the closed proton shell in the {sup 72,74}Ni could be observed, as M{sub n}/M{sub p}>N/Z, indicating an increased contribution of the neutrons to the vibrational amplitude. For the case of {sup 76,80}Zn, M{sub n}/M{sub p}

FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

International Nuclear Information System (INIS)

Auterinen, I.; Salmenhaara, S.E.J. . Author

2004-01-01

The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

Design and fabrication of foam-insulated cryogenic target for wet-wall laser fusion reactor

International Nuclear Information System (INIS)

Norimatsu, T.; Takeda, T.; Nagai, K.; Mima, K.; Yamanaka, T.

2003-01-01

A foam insulated cryogenic target was proposed for use in a future laser fusion reactor with a wet wall. This scheme can protect the solid DT layer from melting due to surface heating by adsorption of metal vapor without significant reduction in the target gain. Design spaces for the injection velocity and the acceptable vapor pressure in the reactor are discussed. Basic technology to fabricate such structure was demonstrated by emulsion process. Concept of a cryogenic fast-ignition target with a gold guiding cone was proposed together with direct injection filling of liquid DT. (author)

Chemical investigations of isotope separation on line target units for carbon and nitrogen beams

CERN Document Server

Franberg, H; Gäggeler, H W; Köster, U

2006-01-01

Radioactive ion beams (RIBs) are of significant interest in a number of applications. Isotope separation on line (ISOL) facilities provide RIB with high beam intensities and good beam quality. An atom that is produced within the ISOL target will first diffuse out from the target material. During the effusion towards the transfer line and into the ion source the many contacts with the surrounding surfaces may cause unacceptable delays in the transport and, hence, losses of the shorter-lived isotopes. We performed systematic chemical investigations of adsorption in a temperature and concentration regime relevant for ISOL targets and ion source units, with regard to CO/sub x/ and NOmaterials are potential construction materials for the above-mentioned areas. Off-line and on-line tests have been performed using a gas thermochromatography setup with radioactive tracers. The experiments were performed at the production of tracers for atmospheric chemistry (PROTRAC) facility at the Paul Schener Institute in Villigen...

Isotopic composition of fission gases in LWR fuel

International Nuclear Information System (INIS)

Jonsson, T.

2000-01-01

Many fuel rods from power reactors and test reactors have been punctured during past years for determination of fission gas release. In many cases the released gas was also analysed by mass spectrometry. The isotopic composition shows systematic variations between different rods, which are much larger than the uncertainties in the analysis. This paper discusses some possibilities and problems with use of the isotopic composition to decide from which part of the fuel the gas was released. In high burnup fuel from thermal reactors loaded with uranium fuel a significant part of the fissions occur in plutonium isotopes. The ratio Xe/Kr generated in the fuel is strongly dependent on the fissioning species. In addition, the isotopic composition of Kr and Xe shows a well detectable difference between fissions in different fissile nuclides. (author)

Job/task analysis for I ampersand C [Instrumentation and Controls] instrument technicians at the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Duke, L.L.

1989-09-01

To comply with Department of Energy Order 5480.XX (Draft), a job/task analysis was initiated by the Maintenance Management Department at Oak Ridge National Laboratory (ORNL). The analysis was applicable to instrument technicians working at the ORNL High Flux Isotope Reactor (HFIR). This document presents the procedures and results of that analysis. 2 refs., 2 figs

Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

Energy Technology Data Exchange (ETDEWEB)

Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

2016-09-15

A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.

Production of Radioisotopes in Pakistan Research Reactor: Past, Present and Future

International Nuclear Information System (INIS)

Mushtaq, A.

2013-01-01

Radioisotope production to service different sectors of economic significance constitutes an important ongoing activity of many national nuclear programs. Radioisotopes, formed by nuclear reactions on targets in a reactor or cyclotron, require further processing in almost all cases to obtain them in a form suitable for use. The availability of short-lived radionuclides from radionuclide generators provides an inexpensive and convenient alternative to in-house radioisotope production facilities such as cyclotrons and reactors. The reactor offers large volume for irradiation, simultaneous irradiation of several samples, economy of production and possibility to produce a wide variety of radioisotopes. The accelerator-produced isotopes relatively constitute a smaller percentage of total use. (author)

Contributions of each isotope in structural material on radiation damage in a hybrid reactor

International Nuclear Information System (INIS)

Günay, Mehtap

2016-01-01

In this study, the fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. In this study, salt-heavy metal mixtures consisting of 93–85% Li_2_0Sn_8_0 + 5% SFG-PuO_2 and 2-10% UO_2, 93–85% Li_2_0Sn_8_0 + 5% SFG-PuO_2 and 2-10% NpO_2, and 93–85% Li_2_0Sn_8_0 + 5% SFG-PuO_2 and 2-10% UCO were used as fluids. In this study, the effect on the radiation damage of spent fuel-grade (SFG)-PuO_2, UO_2, NpO_2 and UCO contents was investigated in the structural material of a designed fusion–fission hybrid reactor system. In the designed hybrid reactor system were investigated the effect on the radiation damage of the selected fluid according to each isotopes of structural material in the structural material for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library

Total quality management for addressing suspect parts at the Oak Ridge High Flux Isotope Reactor

International Nuclear Information System (INIS)

Hendrix, K.A.; Tulay, M.P.

1993-01-01

Martin Marietta Energy System (MMES) Research Reactors Division (RRD), operator of the High Flux Isotope Reactor (HFIR) recently embarked on an aggressive Program to address the issue of suspect Parts and to enhance their procurement process. Through the application of TQM process improvement, RRD has already achieved improved efficiency in specifying, procuring, and accepting replacement items for its largest research reactor. These process improvements have significantly decreased the risk of installing suspect parts in the HFIR safety systems. To date, a systematic plan has been implemented, which includes the following elements: Process assessment and procedure review; Procedural enhancements; On-site training and technology transfer; Enhanced receiving inspections; Performance supplier evaluations and source verifications integrated processes for utilizing commercial grade products in nuclear safety-related applications. This paper will describe the above elements, how a partnership between MMES and Gilbert/Commonwealth facilitated the execution of the plan, and how process enhancements were applied. We will also present measures for improved efficiency and productivity, that MMES intends to continually address with Quality Action Teams

Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Hurt, Christopher J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hobbs, Randy W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-08-01

There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (²³⁸Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (²³⁷Np) dioxide (NpO₂)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs

Electron linac for medical isotope production with improved energy efficiency and isotope recovery

Science.gov (United States)

Noonan, John; Walters, Dean; Virgo, Matt; Lewellen, John

2015-09-08

A method and isotope linac system are provided for producing radio-isotopes and for recovering isotopes. The isotope linac is an energy recovery linac (ERL) with an electron beam being transmitted through an isotope-producing target. The electron beam energy is recollected and re-injected into an accelerating structure. The ERL provides improved efficiency with reduced power requirements and provides improved thermal management of an isotope target and an electron-to-x-ray converter.

Breeding description for fast reactors and symbiotic reactor systems

International Nuclear Information System (INIS)

Hanan, N.A.

1979-01-01

A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

Development of key technology for the medical isotope production

International Nuclear Information System (INIS)

Oh, Soo Youl; Kim, I. S.; Kim, W. W.; Rhee, C. K.; Park, K. B.; Park, S. J.; Shin, H. S.; Shin, Y. J.

2005-06-01

The objective of this project is to experimentally verify and enhance Mo-99 and Sr-89 recovery/purification processes as the key technologies for the medical isotope production from a solution fuel reactor. A joint experiment was planned between KAERI and Kurchatov Institute (KI), Russia. The kinds of experiments planed are, a series of Mo-99 recovery/purification experiments from the ARGUS reactor which uses High Enriched Uranium (HEU) fuel, a series of the same experiments but from the Low Enriched Uranium (LEU) solution target, a demonstration of the mechanism of Sr-89 delivery from the air medium in the reactor vessel. Meanwhile, the survey and legalistic interpretation of relevant patents shows a possibility of infringement of TCI Inc.'s patents in case of exporting medical isotopes produced at the MIP to Japan and the US so far as the MIP adopts the concept of the Russian ARGUS and recovery/purification process. Eliminating, not minor changing, step(s) or condition(s) of patent processes would help to avoid the patent infringement. Because of a difficulty in the KAERI-KI full-time co-experiments at KI labs, a different idea between two parties about the depth of background information to be provided to KAERI, and other reasons, the experiment plan was not executed

Safety aspects of tritium in ICF reactors with internally-breeding targets

International Nuclear Information System (INIS)

Ragheb, M.; Miley, G.H.; University of Illinois, Urbana, IL)

1985-01-01

The LOTRIT inertial confinement reactor concept employs a deuterium burning target with a DT spark trigger core. This eliminates the need for tritium breeding in a blanket, and leads to a minimization of the tritium inventory and of the possibility of metal fire hazards if lead is used instead of lithium for first wall protection. The active fuel inventory in the fuel cycle and blanket per MJ of energy produced is only 5 percent of the DT case. The most significant reduction in the total tritium inventory is in the target manufacture and storage areas, and is about 1.8% of the DT case per unit of fusion energy produced. If the goal is to reduce the risk from tritium releases from fusion reactors to below that of fission reactors, it is estimated that the tritium releases must be maintained at 0.13-5.0 Ci/day. Attaining these values will be costly, technologically difficult and will constrain the design options in DTbased systems, but may be within the realm of systems using the LOTRIT concept

Eurisol-DS Multi-MW target A proposal for improving overall performance in relation to the isotope yield

CERN Document Server

Samec, K; Kadi, Yacine; Rocca, Roberto; Kharoua, Cyril

2008-01-01

The Eurisol Design Study has been initiated by the European Commission to demonstrate the feasibility of a facility for producing large yields of exotic isotopes. At the core of the projected facility, the neutron source produces spallation neutrons from a proton beam impacting dense liquid metal. The neutrons emitted from the source are used to fission Uranium targets which, in turn, produce high yields of isotopes. This technical report summarises efforts to improve the overall performance of the planned facility, by optimising the neutron source and the disposition of the fission targets.

EURISOL-DS Multi-MW Target. A proposal for improving overall performance in relation to the isotope yield

CERN Document Server

Karel Samec, Mats Lindroos, Yacine Kadi,Roberto Rocca, Cyril KharouaAB Dept. ATB

The Eurisol Design Study has been initiated by the European Commission to demonstratethe feasibility of a facility for producing large yields of exotic isotopes.At the core of the projected facility, the neutron source produces spallation neutrons from aproton beam impacting dense liquid metal. The neutrons emitted from the source are usedto fission Uranium targets which, in turn, produce high yields of isotopes.This technical report summarises efforts to improve the overall performance of the plannedfacility, by optimising the neutron source and the disposition of the fission targets.

The future of producing separated stable isotopes at Oak Ridge National Laboratory for accelerator applications

International Nuclear Information System (INIS)

Collins, E.D.

1994-01-01

Separated stable isotopes, produced in the calutrons at Oak Ridge National Laboratory, are essential target materials for production of numerous radioisotopes in accelerators and reactors. Recently, separated stable isotope production has been curtailed because government appropriations were discontinued and salts revenues decreased. The calutrons were placed in standby and the operating staff reduced to enable support by sales from existing inventories. Appeals were made to industry and government to preserve this national capability. Methods for providing volume-based price reductions were created to attract support from commercial isotope users. In 1994, the Department of Energy's Isotope Production and Distribution Program was restructured and a strategy produced to seek appropriated funding for the future production of rare, nonprofitable isotopes for research uses. This strategy, together with new demands for medical isotopes, will enable future operation of the calutrons. Moreover, production may be enhanced by complementing calutron capabilities with the Plasma Separation Process

Reactivity margins in heavy water moderated production reactors

International Nuclear Information System (INIS)

Benton, F.D.

1981-11-01

The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

A level-playing field for medical isotope production - How to phase-out reliance on HEU

International Nuclear Information System (INIS)

Kuperman, A.J.

1999-01-01

Two decades ago, civilian commerce in highly enriched uranium (HEU) for use as targets in the production of medical isotopes was considered a relatively minor security concern for three reasons. First, the number of producers was small. Second, the amount of HEU involved was small. Third, the amount of HEU was dwarfed by the quantities of HEU in civilian commerce as fuel for nuclear research and test reactors. Now, however, all three variables have changed. First, as the use of medical isotopes has expanded rapidly, production programs are proliferating. Second, as the result of such new producers and the expansion of existing production facilities, the amounts of HEU involved are growing. Third, as the RERTR program has facilitated the phase-out of HEU as fuel in most research and test reactors, the quantities of HEU for isotope production have come to represent a significant percentage of global commerce in this weapons-usable material. Medical isotope producers in several states are cooperating with the RERTR program to convert to low-enriched uranium (LEU) targets within the next few years, and one already relies on LEU for isotope production. However, the three biggest isotope producers - in Canada and the European Union - continue to rely on HEU, creating a double-standard that endangers the goal of the RERTR program. Each of these three producers has expressed economic concerns about being put at a competitive disadvantage if it alone converts. This paper proposes forging a firmer international consensus that all present and future isotope producers should convert to LEU, and calls for codifying such a commitment in a statement of intent to be prepared by producers over the next year. With such a level playing field, no producer would need fear being put at a competitive disadvantage by conversion, or being stigmatized by pressure groups for continued reliance on HEU. The phase-out of all HEU commerce for isotope production could be achieved within about

Nitrous oxide production pathways in a partial nitritation-anammox reactor: Isotopic evidence for nitrous oxide production associated anaerobic ammonium oxidation?

Science.gov (United States)

Wunderlin, P.; Harris, E. J.; Joss, A.; Emmenegger, L.; Kipf, M.; Mohn, J.; Siegrist, H.

2014-12-01

Nitrous oxide (N2O) is a strong greenhouse gas and a major sink for stratospheric ozone. In biological wastewater treatment N2O can be produced via several pathways. This study investigates the dynamics of N2O emissions from a nitritation-anammox reactor, and links its interpretation to the nitrogen and oxygen isotopic signature of the emitted N2O. A 400-litre single-stage nitritation-anammox reactor was operated and continuously fed with digester liquid. The isotopic composition of N2O emissions was monitored online with quantum cascade laser absorption spectroscopy (QCLAS; Aerodyne Research, Inc.; Waechter et al., 2008). Dissolved ammonium and nitrate were monitored online (ISEmax, Endress + Hauser), while nitrite was measured with test strips (Nitrite-test 0-24mgN/l, Merck). Table 1. Summary of experiments conducted to understand N2O emissions Experimental conditions O2[mgO2/L] NO2-[mgN/L] NH4+[mgN/L] N2O/NH4+[%] Normal operation production pathway, which is hypothesized to be mediated by anammox activity (Figure 1). A less likely explanation is that the SP of N2O was increased by partial N2O reduction by heterotrophic denitrification. Various experiments were conducted to further investigate N2O formation pathways in the reactor. Our data reveal that N2O emissions increased when reactor operation was not ideal, for example when dissolved oxygen was too high (Table 1). SP measurements confirmed that these N2O peaks were due to enhanced nitrifier denitrification, generally related to nitrite build-up in the reactor (Figure 1; Table 1). Overall, process control via online N2O monitoring was confirmed to be an ideal method to detect imbalances in reactor operation and regulate aeration, to ensure optimal reactor conditions and minimise N2O emissions. ReferencesWaechter H. et al. (2008) Optics Express, 16: 9239-9244. Wunderlin, P et al. (2013) Environmental Science & Technology 47: 1339-1348.

Problems in producing nuclear reactor for medical isotopes and the Global Crisis of molybdenum supply; Problemas en la produccion en reactores nucleares de isotopos con fines medicos y la crisis mundial de suministro de molibdeno ({sup 9}9Mo)

Energy Technology Data Exchange (ETDEWEB)

Zubiarrain, A.

2011-07-01

Nuclear medicine uses drugs that incorporate a radioactive isotope radiopharmaceuticals. Every year are performed, worldwide, 35 million nuclear medicine procedures, of which 80% are done with radiopharmaceuticals containing the isotope, molybdenum-99, produced in nuclear reactors. In recent years, there have been several supply crisis of molybdenum-99, which have hampered diagnostic procedure with technitium-99m. (Author)

Medical isotope production: A new research initiative for the Annular Core Research Reactor

International Nuclear Information System (INIS)

Coats, R.L.; Parma, E.J.

1993-01-01

An investigation has been performed to evaluate the capabilities of the Annular Core Research Reactor and its supporting Hot Cell Facility for the production of 99 Mo and its separation from the fission product stream. Various target irradiation locations for a variety of core configurations were investigated, including the central cavity, fuel and reflector locations, and special target configurations outside the active fuel region. Monte Carlo techniques, in particular MCNP using ENDF B-V cross sections, were employed for the evaluation. The results indicate that the reactor, as currently configured, and with its supporting Hot Cell Facility, would be capable in meeting the current US demand if called upon. Modest modifications, such as increasing the capacity of the external heat exchangers, would permit significantly higher continuous power operation and even greater 99 Mo production ensuring adequate capacity for future years

Dissolution flowsheet for high flux isotope reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Foster, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U₃O₈) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H₂. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H₂) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H₂ and other permanent gases in the dissolution offgas allowing the development of H₂ generation rate profiles. The H₂ generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

Stable isotopic labeling-based quantitative targeted glycomics (i-QTaG).

Science.gov (United States)

Kim, Kyoung-Jin; Kim, Yoon-Woo; Kim, Yun-Gon; Park, Hae-Min; Jin, Jang Mi; Hwan Kim, Young; Yang, Yung-Hun; Kyu Lee, Jun; Chung, Junho; Lee, Sun-Gu; Saghatelian, Alan

2015-01-01

Mass spectrometry (MS) analysis combined with stable isotopic labeling is a promising method for the relative quantification of aberrant glycosylation in diseases and disorders. We developed a stable isotopic labeling-based quantitative targeted glycomics (i-QTaG) technique for the comparative and quantitative analysis of total N-glycans using matrix-assisted laser desorption/ionization time-of-flight mass spectrometry (MALDI-TOF MS). We established the analytical procedure with the chemical derivatizations (i.e., sialic acid neutralization and stable isotopic labeling) of N-glycans using a model glycoprotein (bovine fetuin). Moreover, the i-QTaG using MALDI-TOF MS was evaluated with various molar ratios (1:1, 1:2, 1:5) of (13) C6 /(12) C6 -2-aminobenzoic acid-labeled glycans from normal human serum. Finally, this method was applied to direct comparison of the total N-glycan profiles between normal human sera (n = 8) and prostate cancer patient sera (n = 17). The intensities of the N-glycan peaks from i-QTaG method showed a good linearity (R(2) > 0.99) with the amount of the bovine fetuin glycoproteins. The ratios of relative intensity between the isotopically 2-AA labeled N-glycans were close to the theoretical molar ratios (1:1, 1:2, 1:5). We also demonstrated that the up-regulation of the Lewis antigen (~82%) in sera from prostate cancer patients. In this proof-of-concept study, we demonstrated that the i-QTaG method, which enables to achieve a reliable comparative quantitation of total N-glycans via MALDI-TOF MS analysis, has the potential to diagnose and monitor alterations in glycosylation associated with disease states or biotherapeutics. © 2015 American Institute of Chemical Engineers.

Reactor potential for magnetized target fusion

International Nuclear Information System (INIS)

Dahlin, J.E.

2001-06-01

Magnetized Target Fusion (MTF) is a possible pathway to thermonuclear fusion different from both magnetic fusion and inertial confinement fusion. An imploding cylindrical metal liner compresses a preheated and magnetized plasma configuration until thermonuclear conditions are achieved. In this report the Magnetized Target Fusion concept is evaluated and a zero-dimensional computer model of the plasma, liner and circuit as a connected system is designed. The results of running this code are that thermonuclear conditions are achieved indeed, but only during a very short time. At peak compression the pressure from the compressed plasma and magnetic field is so large reversing the liner implosion into an explosion. The time period of liner motion reversal is termed the dwell time and is crucial to the performance of the fusion system. Parameters as liner thickness and plasma density are certainly of significant importance to the dwell time, but it seems like a reactor based on the MTF principle hardly can become economic if not innovative solutions are introduced. In the report two such solutions are presented as well

Reactor potential for magnetized target fusion

Energy Technology Data Exchange (ETDEWEB)

Dahlin, J.E

2001-06-01

Magnetized Target Fusion (MTF) is a possible pathway to thermonuclear fusion different from both magnetic fusion and inertial confinement fusion. An imploding cylindrical metal liner compresses a preheated and magnetized plasma configuration until thermonuclear conditions are achieved. In this report the Magnetized Target Fusion concept is evaluated and a zero-dimensional computer model of the plasma, liner and circuit as a connected system is designed. The results of running this code are that thermonuclear conditions are achieved indeed, but only during a very short time. At peak compression the pressure from the compressed plasma and magnetic field is so large reversing the liner implosion into an explosion. The time period of liner motion reversal is termed the dwell time and is crucial to the performance of the fusion system. Parameters as liner thickness and plasma density are certainly of significant importance to the dwell time, but it seems like a reactor based on the MTF principle hardly can become economic if not innovative solutions are introduced. In the report two such solutions are presented as well.

Isotopic marking and tracers

International Nuclear Information System (INIS)

Morel, F.

1997-01-01

The use of radioactive isotopes as tracers in biology has been developed thanks to the economic generation of the required isotopes in accelerators and nuclear reactors, and to the multiple applications of tracers in the life domain; the most usual isotopes employed in biology are carbon, hydrogen, phosphorus and sulfur isotopes, because these elements are present in most of organic molecules. Most of the life science knowledge appears to be dependent to the extensive use of nuclear tools and radioactive tracers; the example of the utilization of radioactive phosphorus marked ATP to study the multiple reactions with proteins, nucleic acids, etc., is given

Neutronic analysis for the fission Mo99 production by irradiation of leu targets in TRIGA 14 MW reactor

International Nuclear Information System (INIS)

Dulugeac, S. D.; Mladin, M.; Budriman, A. G.

2013-01-01

Molybdenum production can be a solution for the future in the utilization of the Romanian TRIGA, taking into account the international market supply needs. Generally, two different techniques are available for Mo 99 production for use in medical Tc 99 generation.The first one is based on neutron irradiation of molybdenum targets of natural isotopic composition or enriched in Mo 98 . In a second process, Mo 99 is obtained as a result of the neutron induced fission of U 235 according to U 235 (n,f) Mo 99 . The objectives of the paper are related to Mo 99 production as a result of fission. Neutron physics parameters are determined and presented, such as: thermal flux axial distribution for the critical reactor at 10 MW inside the irradiation location; reactivity introduced by three Uranium foil containers; neutron fluxes and fission rates in the Uranium foils; released and deposited power in the Uranium foils; Mo 99 activity in the Uranium foils. (authors)

In-reactor oxidation of zircaloy-4 under low water vapor pressures

Science.gov (United States)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

In-reactor oxidation of zircaloy-4 under low water vapor pressures

International Nuclear Information System (INIS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

International Nuclear Information System (INIS)

Anon.

1981-01-01

The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

International Nuclear Information System (INIS)

1994-03-01

On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci 192 Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape

PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Masood, Z.

2016-01-01

The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

Dynamic response of the high flux isotope reactor structure caused by nearby heavy load drop

International Nuclear Information System (INIS)

Chang, Shih-Jung.

1995-01-01

A heavy load of 50,000 lb is assumed to drop from 10 ft above the bottom of the High Flux Isotope Reactor (HFIR) pool at the loading station. The consequences of the dynamic impact to the bottom slab of the pool and to the nearby HFIR reactor vessel are analyzed by applying the ABAQUS computer code The results show that both the BM vessel structure and its supporting legs are subjected to elastic disturbances only and, therefore, will not be damaged. The bottom slab of the pool, however, will be damaged to about half of the slab thickness. The velocity response spectrum at the concrete floor next to the HFIR vessel as a result of the vibration caused by the impact is obtained. It is concluded, that the damage caused by heavy load drop at the loading station is controlled by the slab damage and the nearby HFIR vessel and the supporting legs will not be damaged

Fuel reprocessing data validation using the isotope correlation technique

International Nuclear Information System (INIS)

Persiani, P.J.; Bucher, R.G.; Pond, R.B.; Cornella, R.J.

1990-01-01

The Isotope Correlation Technique (ICT), in conjunction with the gravimetric (Pu/U ratio) method for mass determination, provides an independent verification of the input accountancy at the dissolver or accountancy stage of the reprocessing plant. The Isotope Correlation Technique has been applied to many classes of domestic and international reactor systems (light-water, heavy-water, and graphite reactors) operating in a variety of modes (power, research, and production reactors), and for a variety of reprocessing fuel cycle management strategies. Analysis of reprocessing operations data based on isotopic correlations derived for assemblies in a PWR environment and fuel management scheme, yielded differences between the measurement-derived and ICT-derived plutonium mass determinations of (- 0.02 ± 0.23)% for the measured U-235 and (+ 0.50 ± 0.31)% for the measured Pu-239, for a core campaign. The ICT analyses has been implemented for the plutonium isotopics in a depleted uranium assembly in a heavy-water, enriched uranium system and for the uranium isotopes in the fuel assemblies in light-water, highly-enriched systems

Apparatus for isotopic alteration of mercury vapor

International Nuclear Information System (INIS)

Grossman, M.W.; George, W.A.; Marcucci, R.V.

1988-01-01

This patent describes an apparatus for enriching the isotopic content of mercury. It comprises: a low pressure electric discharge lamp, the lamp comprising an envelope transparent to ultraviolet radiation and containing a fill comprising mercury and an inert gas; a filter concentrically arranged around the low pressure electric discharge lamp, the filter being transparent to ultraviolet radiation and containing mercury including 196 Hg isotope; means for controlling mercury pressure in the filter; and a reactor arranged around the filter such that radiation passes from the low pressure electric discharge lamp through the filter and into Said reactor, the reactor being transparent to ultraviolet light

Behavior of actinides in the Integral Fast Reactor fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

1994-06-01

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

Actinide behavior in the Integral Fast Reactor. Final project report

Energy Technology Data Exchange (ETDEWEB)

Courtney, J.C.

1994-11-01

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

Actinide behavior in the Integral Fast Reactor. Final project report

International Nuclear Information System (INIS)

Courtney, J.C.

1994-11-01

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

Behavior of actinides in the Integral Fast Reactor fuel cycle

International Nuclear Information System (INIS)

Courtney, J.C.; Lineberry, M.J.

1994-01-01

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors' confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

Radioisotope production in fusion reactors

International Nuclear Information System (INIS)

Engholm, B.A.; Cheng, E.T.; Schultz, K.R.

1986-01-01

Radioisotope production in fusion reactors is being investigated as part of the Fusion Applications and Market Evaluation (FAME) study. /sup 60/Co is the most promising such product identified to date, since the /sup 60/Co demand for medical and food sterilization is strong and the potential output from a fusion reactor is high. Some of the other radioisotopes considered are /sup 99/Tc, /sup 131/l, several Eu isotopes, and /sup 210/Po. Among the stable isotopes of interest are /sup 197/Au, /sup 103/Rh and Os. In all cases, heat or electricity can be co-produced from the fusion reactor, with overall attractive economics

High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

Energy Technology Data Exchange (ETDEWEB)

Pudelek, R. E.; Gilbert, W. C.

2002-02-26

This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste

Geometric optimization of spallation targets for the MYRRHA reactor using MCNPX simulations

Energy Technology Data Exchange (ETDEWEB)

Rebello Junior, Andre Luiz P.; Martinez, Aquilino S.; Golcalves, Alessandro C., E-mail: junior.rebello@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Nuclear

2013-07-01

The present work aims to evaluate the behavior of neutron multiplicity in a spallation target using MCNPX simulations, focusing on its application in the MYRRHA reactor. It was studied the two types of spallation target proposed for the MYRRHA project, windowless and windows target, in order to compare them and nd saturation boundaries. Some saturation boundaries were found and the windowless target proved to be as viable as the windows one. Each one produced nearly the same number of neutrons per incident proton. Using the concept of neutron cost, it was also observed that the optimum conditions on neutron production occur at about 1GeV, for both target designs. (author)

Geometric optimization of spallation targets for the MYRRHA reactor using MCNPX simulations

International Nuclear Information System (INIS)

Rebello Junior, Andre Luiz P.; Martinez, Aquilino S.; Golcalves, Alessandro C.

2013-01-01

The present work aims to evaluate the behavior of neutron multiplicity in a spallation target using MCNPX simulations, focusing on its application in the MYRRHA reactor. It was studied the two types of spallation target proposed for the MYRRHA project, windowless and windows target, in order to compare them and nd saturation boundaries. Some saturation boundaries were found and the windowless target proved to be as viable as the windows one. Each one produced nearly the same number of neutrons per incident proton. Using the concept of neutron cost, it was also observed that the optimum conditions on neutron production occur at about 1GeV, for both target designs. (author)

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

Energy Technology Data Exchange (ETDEWEB)

Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

2012-03-01

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

International Nuclear Information System (INIS)

Renfro, David G.; Cook, David Howard; Freels, James D.; Griffin, Frederick P.; Ilas, Germina; Sease, John D.; Chandler, David

2012-01-01

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Development of low enrichment technologies for high density fuels and for isotope production targets

International Nuclear Information System (INIS)

Taboada, Horacio; Gonzalez, Alfredo G.

2005-01-01

Since more than twenty years ago, CNEA has carried out RERTR activities. Main goals are to convert the RA 6 reactor core from HEU to LEU, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, and to optimize techniques to recover U from silicide scrap samples. The future plans include: 1) Completion the RA 6 reactor conversion to LEU; 2) Qualification by irradiation of the promising solutions found for the high density fuels; 3) Irradiation of mini plates and full scale fuel assemblies at the RA 3 reactor and at higher flux and temperature reactors; 4) Optimization of LEU target and radiochemical techniques for radioisotope production. (author) [es

New method for the hydrogen isotope exchange reaction in a hydrophobic catalyst bed

International Nuclear Information System (INIS)

Asakura, Y.; Kikuchi, M.; Yusa, H.

1982-01-01

To improve the isotope exchange reaction efficiency between water and hydrogen, a new reactor in which water mists and hydrogen gas react cocurrently was studied. To apply this to the enrichment of tritium in heavy water, a dual temperature isotope exchange reactor which is composed of cocurrent low temperature reactors and the usual countercurrent high temperature reactor was proposed and analyzed using a McCabe-Thiele diagram. By utilizing cocurrent reactors, in combination, the necessary catalyst volume can be reduced to one-tenth as compared with the usual countercurrent low temperature reactor. 17 refs

Hot vacuum extraction-isotopic dilution mass spectrometry for determination of hydrogen isotopes in zircaloys

International Nuclear Information System (INIS)

Shi, Y.; Leeson, P.K.; Wilkin, D.; Britton, A.; Macleod, R.

2016-01-01

A hot vacuum extraction-isotope dilution mass spectrometry (HVE-IDMS) was studied for determination of hydrogen isotopes in zirconium metal and alloys as nuclear reactor materials. A theoretical assessment of the completeness of the extraction of hydrogen isotopes under the chosen condition was carried out based on the hydrogen and deuterium solubility data for zirconium. The optimal isotopic spiking condition for conventional IDMS was further explored for the special case IDMS where the isotope abundance of the samples is varied and non-natural. Applying the optimal conditions, the accurate IDMS determination was realized. The agreement between the measured values and the certified or prepared values of standard reference materials and homemade standard materials validate the method developed. (author)

Status of research reactors in China. Their utilization and safety upgrading

International Nuclear Information System (INIS)

Xu Hanming; Jin Huajin

2000-01-01

The main research reactors in China basically consist of several old reactors including HWRR, HFETR, SPR, MJTR and MNSR. Except the last one, all the other reactors operate at a high power density and represent themselves as main tools in China for engineering testing, radioactive isotope production, and neutron scattering research. The research and production activities by these reactors are briefed. Main equipment and research topics for neutron scattering are described. The production of radioisotope is summarized. Safety upgrading activities in recent years taken by these old reactors are described, which make the safety feature of each reactor significantly improved and on the whole more close to (even not completely consistent) with the targets set by the modern safety regulation. Since a new multi-purpose research reactor CARR is expected available around the year of 2005, a schedule about the construction of new reactor, reforming or decommissioning of old reactors and smoothly transition of research and production activities from old to new reactor during the coming years has been under careful planning. A suggestion of potential international cooperation items has been preliminarily given. (author)

Almost twenty years' search of transuranium isotopes in effluents discharged to air from nuclear power plants with VVER reactors.

Science.gov (United States)

Hölgye, Z; Filgas, R

2006-04-01

Airborne effluents of 5 stacks (stacks 1-5) of three nuclear power plants, with 9 pressurized water reactors VVER of 4,520 MWe total power, were searched for transuranium isotopes in different time periods. The search started in 1985. The subject of this work is a presentation of discharge data for the period of 1998-2003 and a final evaluation. It was found that 238Pu, 239,240Pu, 241Am, 242Cm, and 244Cm can be present in airborne effluents. Transuranium isotope contents in most of the quarterly effluent samples from stacks 2, 4 and 5 were not measurable. Transuranium isotopes were present in the effluents from stack l during all 9 years of the study and from stack 3 since the 3rd quarter of 1996 as a result of a defect in the fuel cladding. A relatively high increase of transuranium isotopes in effluents from stack 3 occurred in the 3rd quarter of 1999, and a smaller increase occurred in the 3rd quarter of 2003. In each instance 242Cm prevailed in the transuranium isotope mixtures. 238Pu/239,240Pu, 241Am/239,240Pu, 242Cm/239,240Pu, and 244Cm/239,240Pu ratios in fuel for different burn-up were calculated, and comparison of these ratios in fuel and effluents was performed.

Production of transplutonium elements in the high flux isotope reactor (HFIR)

International Nuclear Information System (INIS)

Bigelow, J.E.; Corbett, B.L.; King, L.J.; McGuire, S.C.; Sims, T.M.

1980-01-01

The techniques described have been demonstrated to be adequate to predict the contents of transplutonium element production targets which have been irradiated in the HFIR. The deviations, at least for isotopes of mass 253 or less, are generally within the usual analytical uncertainties, or else are for isiotopes which are of little overall import to the program. Work is especially needed to get a better picture of the production of 250 Cm, 254 Es, 255 Es, and ultimately 257 Fm, since researchers are frequently stating their interest in obtaining larger quantities of these rare and difficult-to-produce nuclides

The possible transmutation of radioactive waste from nuclear reactors

International Nuclear Information System (INIS)

Harries, J.R.

1974-01-01

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

Solution of the isotopic depletion equation using decomposition method and analytical solution

Energy Technology Data Exchange (ETDEWEB)

Prata, Fabiano S.; Silva, Fernando C.; Martinez, Aquilino S., E-mail: fprata@con.ufrj.br, E-mail: fernando@con.ufrj.br, E-mail: aquilino@lmp.ufrj.br [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear

2011-07-01

In this paper an analytical calculation of the isotopic depletion equations is proposed, featuring a chain of major isotopes found in a typical PWR reactor. Part of this chain allows feedback reactions of (n,2n) type. The method is based on decoupling the equations describing feedback from the rest of the chain by using the decomposition method, with analytical solutions for the other isotopes present in the chain. The method was implemented in a PWR reactor simulation code, that makes use of the nodal expansion method (NEM) to solve the neutron diffusion equation, describing the spatial distribution of neutron flux inside the reactor core. Because isotopic depletion calculation module is the most computationally intensive process within simulation systems of nuclear reactor core, it is justified to look for a method that is both efficient and fast, with the objective of evaluating a larger number of core configurations in a short amount of time. (author)

Solution of the isotopic depletion equation using decomposition method and analytical solution

International Nuclear Information System (INIS)

Prata, Fabiano S.; Silva, Fernando C.; Martinez, Aquilino S.

2011-01-01

In this paper an analytical calculation of the isotopic depletion equations is proposed, featuring a chain of major isotopes found in a typical PWR reactor. Part of this chain allows feedback reactions of (n,2n) type. The method is based on decoupling the equations describing feedback from the rest of the chain by using the decomposition method, with analytical solutions for the other isotopes present in the chain. The method was implemented in a PWR reactor simulation code, that makes use of the nodal expansion method (NEM) to solve the neutron diffusion equation, describing the spatial distribution of neutron flux inside the reactor core. Because isotopic depletion calculation module is the most computationally intensive process within simulation systems of nuclear reactor core, it is justified to look for a method that is both efficient and fast, with the objective of evaluating a larger number of core configurations in a short amount of time. (author)

Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

International Nuclear Information System (INIS)

Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V.; Boulyga, S.F.; Becker, J.S.

2005-01-01

An analytical method is described for the estimation of uranium concentrations, of 235 U/ 238 U and 236 U/ 238 U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10 -9 g/g to 2.0 x 10 -6 g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing 235 U/ 238 U and 236 U/ 238 U isotope ratios and the average value amounted to 9.4±0.3 MWd/(kg U). (orig.)

Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

Energy Technology Data Exchange (ETDEWEB)

Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V. [Inst. of Radiobiology, Minsk Univ. (Belarus); Boulyga, S.F. [Inst. of Inorganic Chemistry and Analytical Chemistry, Johannes Gutenberg-Univ. Mainz, Mainz (Germany); Becker, J.S. [Central Div. of Analytical Chemistry, Research Centre Juelich, Juelich (Germany)

2005-07-01

An analytical method is described for the estimation of uranium concentrations, of {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10{sup -9}g/g to 2.0 x 10{sup -6}g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and the average value amounted to 9.4{+-}0.3 MWd/(kg U). (orig.)

Study of isotopic exchange reactors (1961); Etude des reacteurs d'echange isotopique (1961)

Energy Technology Data Exchange (ETDEWEB)

Grandcollot, P; Dirian, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

A study is made of the general case of the theory of first-order isotopic chemical exchange between a gaseous and a liquid phase in a reactor, starting from fundamental reaction kinetics data, and without making any limiting hypothesis concerning the value of the separation factor. The cases of counter-current reactors and of co-current reactors are considered successively. The general deuterium conservation equation requires the definition of the quotient of the reactor; the performances of this reactor are characterised by its overall efficiency. The idea of the ratio is introduced because it represents a convenient intermediary in the calculations. The search for an additive value for reactors in series leads logically to the defining of an exchange capacity, and a total efficiency, or number of theoretical reactors. This method of expressing the performances of a reactor is more general than the efficiency due to Murphee which only has a physical significance in the particular case of homogeneous liquid reactors. The relationships between these various quantities are established, and the representation due to Mc Cabe and Thiele is generalized. The reactor performances are linked to the first - order reaction kinetics by the transfer number. The relationships are given for a certain number of concrete cases. Finally the application of these calculations is given, together with the approximations necessary in the case where, because of the presence of several components in each phase, the exchange reaction no longer obeys a single kinetic law. (authors) [French] On examine dans le cas general la theorie d'un reacteur quelconque pour l'echange chimique isotopique du premier ordre entre une phase gazeuse et une phase liquide, a partir des donnees fondamentales sur la cinetique de la reaction, sans faire aucune hypothese limitative sur le cas des reacteurs a contre ourant, puis celui des reacteurs a co-courant. L'equation generale de conservation du deuterium

Reduce, reuse and recycle: a green solution to Canada's medical isotope shortage.

Science.gov (United States)

Galea, R; Ross, C; Wells, R G

2014-05-01

Due to the unforeseen maintenance issues at the National Research Universal (NRU) reactor at Chalk River and coincidental shutdowns of other international reactors, a global shortage of medical isotopes (in particular technetium-99m, Tc-99m) occurred in 2009. The operation of these research reactors is expensive, their age creates concerns about their continued maintenance and the process results in a large amount of long-lived nuclear waste, whose storage cost has been subsidized by governments. While the NRU has since revived its operations, it is scheduled to cease isotope production in 2016. The Canadian government created the Non-reactor based medical Isotope Supply Program (NISP) to promote research into alternative methods for producing medical isotopes. The NRC was a member of a collaboration looking into the use of electron linear accelerators (LINAC) to produce molybdenum-99 (Mo-99), the parent isotope of Tc-99m. This paper outlines NRC's involvement in every step of this process, from the production, chemical processing, recycling and preliminary animal studies to demonstrate the equivalence of LINAC Tc-99m with the existing supply. This process stems from reusing an old idea, reduces the nuclear waste to virtually zero and recycles material to create a green solution to Canada's medical isotope shortage. © 2013 Published by Elsevier Ltd.

Design and Optimization for the Windowless Target of the China Nuclear Waste Transmutation Reactor

Directory of Open Access Journals (Sweden)

Desheng Cheng

2016-04-01

Full Text Available A windowless spallation target can provide a neutron source and maintain neutron chain reaction for a subcritical reactor, and is a key component of China's nuclear waste transmutation of coupling accelerator and subcritical reactor. The main issue of the windowless target design is to form a stable and controllable free surface that can ensure that energy spectrum distribution is acquired for the neutron physical design when the high energy proton beam beats the lead–bismuth eutectic in the spallation target area. In this study, morphology and flow characteristics of the free surface of the windowless target were analyzed through the volume of fluid model using computational fluid dynamics simulation, and the results show that the outlet cross section size of the target is the key to form a stable and controllable free surface, as well as the outlet with an arc transition. The optimization parameter of the target design, in which the radius of outlet cross section is 60 ± 1 mm, is verified to form a stable and controllable free surface and to reduce the formation of air bubbles. This work can function as a reference for carrying out engineering design of windowless target and for verification experiments.

Isotope analysis of closely adjacent minerals

International Nuclear Information System (INIS)

Smith, M.P.

1990-01-01

This patent describes a method of determining an indicator of at least one of hydrocarbon formation, migration, and accumulation during mineral development. It comprises: searching for a class of minerals in a mineral specimen comprising more than one class of minerals; identifying in the mineral specimen a target sample of the thus searched for class; directing thermally pyrolyzing laser beam radiation onto surface mineral substance of the target sample in the mineral specimen releasing surface mineral substance pyrolysate gases therefrom; and determining isotope composition essentially of the surface mineral substance from analyzing the pyrolysate gases released from the thus pyrolyzed target sample, the isotope composition including isotope(s) selected from the group consisting of carbon, hydrogen, and oxygen isotopes; determining an indicator of at least one of hydrocarbon formation, migration, and accumulation during mineral development of the target mineral from thus determined isotope composition of surface mineral substance pyrolysate

Calculation of radiation production of high specific activity isotopes 192Ir and 60Co

International Nuclear Information System (INIS)

Zhou Quan; Zhong Wenfa; Xu Xiaolin

1997-01-01

The high specific activity isotopes: 192 Ir and 60 Co in the high neutron flux reactor are calculated with the method of reactor physics. The results of calculation are analyzed in two aspects: the production of isotopes and the influence to parameters of the reactor, and hence a better case is proposed as a reference to the production

STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

2014-09-01

Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

Fabrication of 121Sb isotopic targets for the study of nuclear high spin features

Science.gov (United States)

Devi, K. Rojeeta; Kumar, Suresh; Kumar, Neeraj; Abhilash, S. R.; Kabiraj, D.

2018-06-01

Isotopic 121Sb targets with 197Au backing have been prepared by Physical Vapor Deposition (PVD) method using the diffusion pump based coating unit at target laboratory, Inter University Accelerator Centre (IUAC), New Delhi, India. The target thickness was measured by stylus profilo-meter and the purity of the targets was investigated by Energy Dispersive X-ray Analysis (EDXA). One of these targets has been used in an experiment which was performed at IUAC for nuclear structure study through fusion evaporation reaction. The excitation function of the 121Sb(12C, yxnγ) reaction has been performed for energies 58 to 70 MeV in steps of 4 MeV. The experimental results were compared with the calculations of statistical models : PACE4 and CASCADE. The methods adopted to achieve best quality foils and good deposition efficiency are reported in this paper.

Identification of the autotrophic denitrifying community in nitrate removal reactors by DNA-stable isotope probing.

Science.gov (United States)

Xing, Wei; Li, Jinlong; Cong, Yuan; Gao, Wei; Jia, Zhongjun; Li, Desheng

2017-04-01

Autotrophic denitrification has attracted increasing attention for wastewater with insufficient organic carbon sources. Nevertheless, in situ identification of autotrophic denitrifying communities in reactors remains challenging. Here, a process combining micro-electrolysis and autotrophic denitrification with high nitrate removal efficiency was presented. Two batch reactors were fed organic-free nitrate influent, with H 13 CO 3 - and H 12 CO 3 - as inorganic carbon sources. DNA-based stable-isotope probing (DNA-SIP) was used to obtain molecular evidence for autotrophic denitrifying communities. The results showed that the nirS gene was strongly labeled by H 13 CO 3 - , demonstrating that the inorganic carbon source was assimilated by autotrophic denitrifiers. High-throughput sequencing and clone library analysis identified Thiobacillus-like bacteria as the most dominant autotrophic denitrifiers. However, 88% of nirS genes cloned from the 13 C-labeled "heavy" DNA fraction showed low similarity with all culturable denitrifiers. These findings provided functional and taxonomical identification of autotrophic denitrifying communities, facilitating application of autotrophic denitrification process for wastewater treatment. Copyright © 2017 Elsevier Ltd. All rights reserved.

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

Production of high-specific activity radionuclides using SM high-flux reactor

International Nuclear Information System (INIS)

Karelin, Ye.A.; Toporov, Yu.G.; Filimonov, V.T.; Vakhetov, F.Z.; Tarasov, V.A.; Kuznetsov, R.A.; Lebedev, V.M.; Andreev, O.I.; Melnik, M.I.; Gavrilov, V.D.

1997-01-01

The development of High Specific Activity (HSA) radionuclides production technologies is one of the directions of RIAR activity, and the high flux research reactor SM, having neutron flux density up to 2.10 15 cm -2 s 1 in a wide range of neutron spectra hardness, plays the principal role in this development. The use of a high-flux reactor for radionuclide production provides the following advantages: production of radionuclides with extremely high specific activity, decreasing of impurities content in irradiated targets (both radioactive and non-radioactive), cost-effective use of expensive isotopically enriched target materials. The production technologies of P-33, Gd-153, W-188, Ni-63, Fe-55,59, Sn-113,117m,119m, Sr- 89, applied in industry, nuclear medicine, research, etc, were developed by RIAR during the last 5-10 years. The research work included the development of calculation procedures for radionuclide reactor accumulation forecast, experimental determination of neutron cross-sections, the development of irradiated materials reprocessing procedures, isolation and purification of radionuclides. The principal results are reviewed in the paper. (authors)

Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

Energy Technology Data Exchange (ETDEWEB)

Aleksandrova, I. V.; Koresheva, E. R., E-mail: elena.koresheva@gmail.com; Krokhin, O. N. [Russian Academy of Sciences, Lebedev Physical Institute (Russian Federation); Osipov, I. E. [Power Efficiency Centre, Inter RAO UES (Russian Federation)

2016-12-15

In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.

Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

International Nuclear Information System (INIS)

Primm, Trent; Guida, Tracey

2010-01-01

Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration/Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

Long-lived isotopes production in Pb-Bi target irradiated by high energy protons

Energy Technology Data Exchange (ETDEWEB)

Korovin, Y.A.; Konobeyev, A.Y.; Pereslavtsev, P.E. [Obninsk Institute of Nuclear Power Engineering, Obninsk (Russian Federation)

1995-10-01

Concentration of long-lived isotopes has been calculated for lead and lead-bismuth targets irradiated by protons with energy 0.4, 0.8, 1.0 and 1.6 GeV. The time of irradiation is equal from 1 month up to 2 years. The data libraries BROND, ADL and MENDL have been used to obtain the rate of nuclide transmutation. All calculations have been performed using the SNT code.

Hydrogen isotope separation experience at the Savannah River Site

International Nuclear Information System (INIS)

Lee, M.W.

1993-01-01

Savannah River Site (SRS) is a sole producer of tritium for US Weapons Program. SRS has built Facilities, developed the tritium handling processes, and operated safely for the last forty years. Tritium is extracted from the irradiated reactor target, purified, mixed with deuterium, and loaded to the booster gas bottle in the weapon system for limited lifetime. Tritium is recovered from the retired bottle and recycled. Newly produced tritium is branded into the recycled tritium. One of the key process is the hydrogen isotope separation that tritium is separated from deuterium and protium. Several processes have been used for the hydrogen isotope separation at SRS: Thermal Diffusion Column (TD), Batch Cryogenic Still (CS), and Batch Chromatography called Fractional Sorption (FS). TD and CS requires straight vertical columns. The overall system separation factor depends on the length of the column. These are three story building high and difficult to put in glove box. FS is a batch process and slow operation. An improved continuous chromatographic process called Thermal Cycling Absorption Process (TCAP) has been developed. It is small enough to be about to put in a glove box yet high capacity comparable to CS. The SRS tritium purification processes can be directly applicable to the Fusion Fuel Cycle System of the fusion reactor

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

International Nuclear Information System (INIS)

Grossbeck, M. L.; Renier, J-P.A.; Bigelow, Tim

2003-01-01

Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

Production of stable isotopes utilizing the plasma separation process

Science.gov (United States)

Bigelow, T. S.; Tarallo, F. J.; Stevenson, N. R.

2005-12-01

A plasma separation process (PSP) is being operated at Theragenics Corporation's®, Oak Ridge, TN, facility for the enrichment of stable isotopes. The PSP utilizes ion cyclotron mass discrimination to separate isotopes on a relatively large scale. With a few exceptions, nearly any metallic element could be processed with PSP. Output isotope enrichment factor depends on natural abundance and mass separation and can be fairly high in some cases. The Theragenics™ PSP facility is believed to be the only such process currently in operation. This system was developed and formerly operated under the US Department of Energy Advanced Isotope Separation program. Theragenics™ also has a laboratory at the PSP site capable of harvesting the isotopes from the process and a mass spectrometer system for analyzing enrichment and product purity. Since becoming operational in 2002, Theragenics™ has utilized the PSP to separate isotopes of several elements including: dysprosium, erbium, gadolinium, molybdenum and nickel. Currently, Theragenics™ is using the PSP for the separation of 102Pd, which is used as precursor for the production of 103Pd. The 103Pd radioisotope is the active ingredient in TheraSeed®, which is used in the treatment of early stage prostate cancer and being investigated for other medical applications. New industrial, medical and research applications are being investigated for isotopes that can be enriched on the PSP. Pre-enrichment of accelerator or reactor targets offers improved radioisotope production. Theragenics operates 14 cyclotrons for proton activation and has access to HFIR at ORNL for neutron activation of radioisotopes.

Reactor, radioactive isotopes and nuclear energy: their avatars in Venezuela

Energy Technology Data Exchange (ETDEWEB)

Roche, M

1981-03-01

The decision to bring a fair sized (3MW) research reactor to Venezuela, made in 1954 by a single, ambitious and prestige seeking individual working with a dictatorial government, is a clear case of cargo cult, an implicit desire to import industralized countries' science and technology by purchasing key in hand their expensive machine. The reactor has never ceased to experience difficulties since then, not so much of a physical or mechanical, but rather of a human nature and due to the almost grotesque distance between the machine's potentialities and the quantity and quality of personnel available. Demand and motivation have been scarce, because fossil and hydro energy have been so far plentiful. Military motivation was in theory absent. Perspectives have apparently improved, not that a scientific community has been trained and an infrastructure exists. Radioactive isotopes have been widely used in Venezuela, beginning in 1953, for medical practice and biological research. At present about 2.5 million bolivars worth of radioisotopes are imported annually, mostly from the US and to a lesser extent, from UK. Steps are being taken to train nuclear engineers, since most studies thus far indicate the last few years of the century as the time when nuclear energy will begin to enter the picture, and since a period of at least ten years is needed between the decision to build an atomic power plant and the time it goes into operation. Choice of technique has not been made, but an active, although still small, uranium prospecting program has been initiated. It seems as if, by the end of the century, either nuclear energy will have to supplement other sources, or standard of living of Venezuelans - at least that relative minority who can afford to live well - will drop. 2 figures, 2 tables.

Developing the Sandia National Laboratories transportation infrastructure for isotope products and wastes

International Nuclear Information System (INIS)

Trennel, A.J.

1995-01-01

Certain radioactive isotopes for North American and especially the United States' needs are enormously important to the medical community and their numerous patients. The most important medical isotope is 99 Mo, which is currently manufactured by Nordion International Inc. in a single, aging reactor operated by Atomic Energy of Canada, Ltd. The reactor's useful life is expected to end at the turn of the century. Production loss because of reactor shutdown possibilities prompted the US Congress to direct the DOE to provide for a US backup source for this crucial isotope. The SNL Annular Core Research Reactor (ACRR) was evaluated as a site to provide 99 Mo initially and other isotopes that can be economically extracted from the process. Medical isotope production at SNL is a new venture in manufacturing. Should SNL be selected and the project reach the manufacturing stage, SNL would expect to service up to 30% of the US market under normal circumstances as a backup to the Canadian supply with the capability to service 100% should the need arise. The demand for 99 Mo increases each year; hence, the proposed action accommodates growth in demand to meet this increase. The proposed project would guarantee the supply of medical isotopes would continue if either the irradiation or processing activities in Canada were interrupted

Investigation of mechanisms of production of argon, krypton and xenon isotopes formed in heavy targets by protons with an energy ranging from 0.15 to 24 GeV

International Nuclear Information System (INIS)

Sauvageon, Henri

1981-01-01

As experimental results of the investigation of interactions between high-energy protons and nucleus generally lead to the distinction between four types of reaction mechanisms (spallation, fission, fragmentation and isotope production), this research thesis reports the study of this mechanisms by using the so-called 'thick target - thick collector' experiment and by studying the production of various isotopes of rare gases (argon, krypton, xenon). These isotopes are produced by using platinum, gold, bismuth and thorium targets bombarded by protons with an energy ranging from 0.15 to 24 GeV. The author presents the experimental methods (target preparation and irradiation, rare gas analysis system), reports the analysis of thick target - thick-collector experiments (vector-based representation, path determination, path-curve energy, corrections of experimental data, excitation energy of the intermediate nucleus), presents the experimental results, and discusses their interpretation (two-stage model of high energy nuclear reactions, isotopes produced by spallation and by fission, isotopes produced by deep spallation, representations of mechanisms of fragmentation and deep spallation)

Study of the properties of the Am-O system in view of the transmutation of Am 241 in fast reactors

International Nuclear Information System (INIS)

Casalta, S.

1996-04-01

To reduce the long term toxicity of Am 241 it was considered to transmute this isotope in fast reactor. The first part of this thesis is an introduction at this problem. In the second part we give the experimental techniques used for the realisation of an AmO 2 -MgO target (powder metallurgy under inert, oxidizing or reducing atmosphere). The properties of the Am-O system has been analyzed by X diffraction, thermodynamic and ceramography, in the Am 2 O 3 -AmO 2 field. In the third part we study the external exposure risk created by the manufacturing of this target and in the last part the behavior of this target in a fast reactor. 66 refs., 28 figs., 25 tabs., 1 append

Dating of the Francevillian sedimentary series and mineralogic and isotopic (Sm, Nd, Rb, Sr, K, Ar, U, O and C) characterization of the gangue of the reactors 10 and 13. Preliminary report

International Nuclear Information System (INIS)

Gautier-Lafaye, F.; Stille, P.; Bros, R.; Taieb, R.

1993-01-01

This paper summarizes the various ages reported for the diagenetic events in the Francevillian sedimentary series (Precambrian era) and the fission reactors of Oklo. Obviously, differences exist between the ages obtained on the silicate minerals and the ages obtained on the Uranium ores and on the reactors. Clay minerals which crystallized during the fission reactions yield younger ages than the reactors themselves. Similarly, the diagenetic clays (1870 Ma) show younger ages than the Uranium ores (2000 Ma). This is in contrast to mineralogical and field evidence indicating that Uranium mineralization occurred during diagenesis of the Francevillian sediments. These antithetical results give rise to several questions. Does the age obtained on the diagenetic clays date a late thermal event or does the age of the Uranium mineralization reflect a multistage U-Pb history. This work tries to bring answers with the help of new isotopic analysis and studies mineralogy of the gangue of reactors and isotopic compositions in Uranium ores. 8 refs., 4 figs

Actinides burnup in a sodium fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2017-09-15

The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)

A method for evaluation the activity of the reactor components

International Nuclear Information System (INIS)

Gugiu, E.D.; Roth, Cs.

2003-01-01

The ability to predict the radioactivity levels of the reactor components is an important aspect from waste management point of view, as well as from radioprotection purposes. A special case is represented by the research reactors where, one of the major contributions to the radioactivity inventory is due to the experimental devices involved in various research works during reactor life. Generally, aluminum and aluminum alloys are used in manufacturing these devices; as a result, the work presented in this paper is focused on the qualitative and quantitative analysis of the radioactive isotopes contained in these materials. A device used for silicon doping by neutron transmutation that was placed near TRIGA reactor core is investigated. The isotopic content of various samplings drawn from various points of the device was analyzed by gamma spectrometry using a HPGe detector. Computations, using the MCNP5 code, are also performed in order to evaluate the reaction rates for all the isotopes and their reactions. The Monte Carlo simulations are performed for a detailed geometry and material composition of the reactor core and the device. The Origen-S code is also used in order to evaluate the isotopic inventory and the activity values. A detailed analysis regarding the possibility to estimate by computations and/or by gamma spectrometry the activity values of the isotopes which are of interest for decommissioning is presented in the paper. (authors)

Synthesis of neutron-rich transuranic nuclei in fissile spallation targets

Energy Technology Data Exchange (ETDEWEB)

Mishustin, Igor, E-mail: mishustin@fias.uni-frankfurt.de [Frankfurt Institute for Advanced Studies, J.-W. Goethe University, 60438 Frankfurt am Main (Germany); “Kurchatov Institute”, National Research Center, 123182 Moscow (Russian Federation); Malyshkin, Yury, E-mail: malyshkin@fias.uni-frankfurt.de [Frankfurt Institute for Advanced Studies, J.-W. Goethe University, 60438 Frankfurt am Main (Germany); Institute for Nuclear Research, Russian Academy of Sciences, 117312 Moscow (Russian Federation); Pshenichnov, Igor, E-mail: pshenich@fias.uni-frankfurt.de [Frankfurt Institute for Advanced Studies, J.-W. Goethe University, 60438 Frankfurt am Main (Germany); Institute for Nuclear Research, Russian Academy of Sciences, 117312 Moscow (Russian Federation); Greiner, Walter [Frankfurt Institute for Advanced Studies, J.-W. Goethe University, 60438 Frankfurt am Main (Germany)

2015-04-15

A possibility of synthesizing neutron-rich superheavy elements in spallation targets of Accelerator Driven Systems (ADS) is considered. A dedicated software called Nuclide Composition Dynamics (NuCoD) was developed to model the evolution of isotope composition in the targets during a long-time irradiation by intense proton and deuteron beams. Simulation results show that transuranic elements up to {sup 249}Bk can be produced in multiple neutron capture reactions in macroscopic quantities. However, the neutron flux achievable in a spallation target is still insufficient to overcome the so-called fermium gap. Further optimization of the target design, in particular, by including moderating material and covering it by a reflector could turn ADS into an alternative source of transuranic elements in addition to nuclear fission reactors.

Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

2011-07-01

The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

HTCAP-1: a program for calcuating operating temperatures in HFIR target irradiation experiments

International Nuclear Information System (INIS)

Kania, M.J.; Howard, A.M.

1980-06-01

The thermal modeling code, HTCAP-1, calculates in-reactor operating temperatures of fueled specimens contained in the High Flux Isotope Reactor (HFIR) target irradiation experiments (HT-series). Temperature calculations are made for loose particle and bonded fuel rod specimens. Maximum particle surface temperatures are calculated for the loose particles and centerline and surface temperatures for the fuel rods. Three computational models are employed to determine fission heat generation rates, capsule heat transfer analysis, and specimen temperatures. This report is also intended to be a users' manual, and the application of HTCAP-1 to the HT-34 irradiation capsule is presented

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Undercovering the hidden links. Nuclear and isotope techniques target nutritional needs

International Nuclear Information System (INIS)

Iyengar, Venkatesh

2001-01-01

Global nutrition problems raise a host of questions and warrant action by the international community of scientists, nutritionists, physicians and other medical professionals. What steps should be taken to remedy this situation? How can this be accomplished economically? How can progress be monitored? What is the role of technology in the overall monitoring process? The last question, which is most relevant to this article, is of particular importance to the IAEA and its support of nutrition programmes. The IAEA's activities in human nutrition were initiated to apply nuclear and related isotopic techniques for solving problems prevalent in developing countries. Among the numerous applications available, isotopic techniques are uniquely well suited to targeting and tracking progress in food and nutrition development programmes. These are tools that help evaluate nutritional status of individuals and populations, measure nutrient requirements and the uptake and bio-availability of vitamins and minerals. The IAEA's efforts help to: verify the nature of the nutrition problem and the efficacy of specific interventions; implement nutrition intervention programmes by monitoring effectiveness and reducing programme costs; guide in the processing of local foods for optimal nutritional value; serve as early indicators of important long-term health improvements; and strengthen capacity building in developing countries. Among the numerous applications available, isotopic techniques are uniquely well suited to targeting and tracking progress in food and nutrition development programmes. These are tools that help evaluate nutritional status of individuals and populations, measure nutrient requirements and the uptake and bio-availability of vitamins and minerals. The IAEA's efforts help to: verify the nature of the nutrition problem and the efficacy of specific interventions; implement nutrition intervention programmes by monitoring effectiveness and reducing programme costs

Influence of neutron energy on formation of radioisotopes during the irradiation of targets in reactor

Directory of Open Access Journals (Sweden)

P. M. Vorona

2011-09-01

Full Text Available Method of calculation of nuclear transformations in irradiated targets is realized for selection of optimal conditions for accumulation of radioisotopes in reactor, taking into account contributions of different energy neutrons (thermal, resonance and fast. Wide potentialities of program complex MCNP-4C based on the method of statistical testing (Monte Carlo method were used. Positive in proposed method is that all calculations starting from spectra and fluxes of neutrons in reactor and completing by quantity of accumulating nuclei carry out within the framework of the same methodological approach. It was shown by the example of radioactive 98Mo production in Mo98Mo(n, γ99Mo reaction that for achievement of maximal yield of target radionuclide. it is necessary to irradiate start targets of Molybdenum in hard spectrum with essential contribution of resonance neutrons.

Transportation of medical isotopes

International Nuclear Information System (INIS)

Nielsen, D.L.

1997-01-01

A Draft Technical Information Document (HNF-1855) is being prepared to evaluate proposed interim tritium and medical isotope production at the Fast Flux Test Facility (FFTF). This assessment examines the potential health and safety impacts of transportation operations associated with the production of medical isotopes. Incident-free and accidental impacts are assessed using bounding source terms for the shipment of nonradiological target materials to the Hanford Site, the shipment of irradiated targets from the FFTF to the 325 Building, and the shipment of medical isotope products from the 325 Building to medical distributors. The health and safety consequences to workers and the public from the incident-free transportation of targets and isotope products would be within acceptable levels. For transportation accidents, risks to works and the public also would be within acceptable levels. This assessment is based on best information available at this time. As the medical isotope program matures, this analysis will be revised, if necessary, to support development of a final revision to the Technical Information Document

Transportation of medical isotopes

Energy Technology Data Exchange (ETDEWEB)

Nielsen, D.L.

1997-11-19

A Draft Technical Information Document (HNF-1855) is being prepared to evaluate proposed interim tritium and medical isotope production at the Fast Flux Test Facility (FFTF). This assessment examines the potential health and safety impacts of transportation operations associated with the production of medical isotopes. Incident-free and accidental impacts are assessed using bounding source terms for the shipment of nonradiological target materials to the Hanford Site, the shipment of irradiated targets from the FFTF to the 325 Building, and the shipment of medical isotope products from the 325 Building to medical distributors. The health and safety consequences to workers and the public from the incident-free transportation of targets and isotope products would be within acceptable levels. For transportation accidents, risks to works and the public also would be within acceptable levels. This assessment is based on best information available at this time. As the medical isotope program matures, this analysis will be revised, if necessary, to support development of a final revision to the Technical Information Document.

Reduce, reuse and recycle: A green solution to Canada's medical isotope shortage

International Nuclear Information System (INIS)

Galea, R.; Ross, C.; Wells, R.G.

2014-01-01

Due to the unforeseen maintenance issues at the National Research Universal (NRU) reactor at Chalk River and coincidental shutdowns of other international reactors, a global shortage of medical isotopes (in particular technetium-99m, Tc-99m) occurred in 2009. The operation of these research reactors is expensive, their age creates concerns about their continued maintenance and the process results in a large amount of long-lived nuclear waste, whose storage cost has been subsidized by governments. While the NRU has since revived its operations, it is scheduled to cease isotope production in 2016. The Canadian government created the Non-reactor based medical Isotope Supply Program (NISP) to promote research into alternative methods for producing medical isotopes. The NRC was a member of a collaboration looking into the use of electron linear accelerators (LINAC) to produce molybdenum-99 (Mo-99), the parent isotope of Tc-99m. This paper outlines NRC’s involvement in every step of this process, from the production, chemical processing, recycling and preliminary animal studies to demonstrate the equivalence of LINAC Tc-99m with the existing supply. This process stems from reusing an old idea, reduces the nuclear waste to virtually zero and recycles material to create a green solution to Canada's medical isotope shortage. - Highlights: • Commercial power electron accelerators are realistic option to produce 99 Mo. • Could cover national demand of Canada. • Demonstrate LINAC- 99 Mo as environmental and economical solution to isotope crisis. • Demonstrate LINAC- 99m Tc to be clinically equivalent to current fission- 99m Tc supply

Multi purpose research reactor

International Nuclear Information System (INIS)

Raina, V.K.; Sasidharan, K.; Sengupta, Samiran; Singh, Tej

2006-01-01

At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor

SFCOMPO 2.0 - A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

Science.gov (United States)

Michel-Sendis, Franco; Martinez-González, Jesus; Gauld, Ian

2017-09-01

SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF) samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA). The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99 mTc for medical purposes is currently produced from the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers. (author)

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99m Tc for medical purposes is currently produced form the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers

Experimental observation and investigation of reactor Cs-137 isotope deactivation in biological cells

International Nuclear Information System (INIS)

Vysotskii, V.I.; Tashyrev, A.B.; Kornilova, A.A.

2007-01-01

Complete text of publication follows. The problem of natural accelerated deactivation of radioactive waste (including deactivation in environmental) is studied. In the work the process of direct controlled deactivation of water mixture of selected different longlived radioactive isotopes in growing microbiological cultures has been studied. The process was connected with transmutation of long-lived active nuclei to non-radioactive isotopes during growth and metabolism of special microbiological MCT ('microbial catalyst-transmutator'). The MCT is the special granules that include: concentrated biomass of metabolically active microorganisms, sources of carbon and energy, phosphorus, nitrogen, etc., and gluing substances that keep all components in the form of granules stable in water solutions for a long period of time at any external conditions. The base of the MCT is microbe syntrophin associations of thousands different microorganism kinds that are in the state of complete symbiosis. These microorganisms appertain to different physiological groups that represent practically the whole variety of the microbe metabolism and relevantly all kinds of microbe accumulation mechanisms. The state of complete symbiosis of the syntrophin associations results on the possibility of maximal adaptation of the microorganisms' association to any external conditions change. The mechanism of nuclear transmutation in growing biological system is described in details in the book. The research has been carried out on the basis of the same distilled water that contained different long-lived reactor isotopes (e.g., Eu 154 , Eu 155 , Cs 137 , Am 241 ). In our experiments 8 identical closed glass flasks with 10 ml of the same active water in each were used. The 'microbial catalyst-transmutator' was placed in 7 glass flasks. In six different flasks different pure K, Ca, Mg, Na, Fe and P salts as single admixture were added to the active water. These chemical elements are vitally necessary

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

Progress in ISOL target-ion source systems

Energy Technology Data Exchange (ETDEWEB)

Koester, U. [Institut Laue Langevin, 6 Rue Jules Horowitz, F-38042 Grenoble Cedex 9 (France); ISOLDE, CERN, CH-1211 Geneve 23 (Switzerland)], E-mail: koester@ill.fr; Arndt, O. [HGF VISTARS and Institut fuer Kernchemie, Johannes-Gutenberg Universitaet Mainz, D-55128 Mainz (Germany); Bouquerel, E.; Fedoseyev, V.N. [ISOLDE, CERN, CH-1211 Geneve 23 (Switzerland); Franberg, H. [ISOLDE, CERN, CH-1211 Geneve 23 (Switzerland); Laboratory for Radio- and Environmental Chemistry, Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Joinet, A. [ISOLDE, CERN, CH-1211 Geneve 23 (Switzerland); Centre d' Etude Spatiale des Rayonnements, 9 Av. du Colonel Roche, F-31028 Toulouse Cedex 4 (France); Jost, C. [HGF VISTARS and Institut fuer Kernchemie, Johannes-Gutenberg Universitaet Mainz, D-55128 Mainz (Germany); Kerkines, I.S.K. [Laboratory of Physical Chemistry, National and Kapodistrian University of Athens, Department of Chemistry, Zografou 157 71, GR (Greece); Cherry L. Emerson Center for Scientific Computation and Department of Chemistry, Emory University, Atlanta, GA 30322 (United States); Kirchner, R. [Gesellschaft fuer Schwerionenforschung, Planckstr. 1, D-64291 Darmstadt (Germany)

2008-10-15

The heart of every ISOL (isotope separation on-line) facility is its target and ion source system. Its efficiency, selectivity and rapidity is decisive for the production of intense and pure ion beams of short-lived isotopes. Recent progress in ISOL target and ion source technology is discussed at the examples of radioactive ion beams of exotic zinc and tin isotopes that were purified by isothermal chromatography and molecular sideband separation respectively. An outlook is given to which other elements these purification methods are applicable.

Progress in ISOL target-ion source systems

International Nuclear Information System (INIS)

Koester, U.; Arndt, O.; Bouquerel, E.; Fedoseyev, V.N.; Franberg, H.; Joinet, A.; Jost, C.; Kerkines, I.S.K.; Kirchner, R.

2008-01-01

The heart of every ISOL (isotope separation on-line) facility is its target and ion source system. Its efficiency, selectivity and rapidity is decisive for the production of intense and pure ion beams of short-lived isotopes. Recent progress in ISOL target and ion source technology is discussed at the examples of radioactive ion beams of exotic zinc and tin isotopes that were purified by isothermal chromatography and molecular sideband separation respectively. An outlook is given to which other elements these purification methods are applicable.

Identification of tertiary butyl alcohol (TBA)-utilizing organisms in BioGAC reactors using 13C-DNA stable isotope probing.

Science.gov (United States)

Aslett, Denise; Haas, Joseph; Hyman, Michael

2011-09-01

Biodegradation of the gasoline oxygenates methyl tertiary-butyl ether (MTBE) and ethyl tertiary-butyl ether (ETBE) can cause tertiary butyl alcohol (TBA) to accumulate in gasoline-impacted environments. One remediation option for TBA-contaminated groundwater involves oxygenated granulated activated carbon (GAC) reactors that have been self-inoculated by indigenous TBA-degrading microorganisms in ground water extracted from contaminated aquifers. Identification of these organisms is important for understanding the range of TBA-metabolizing organisms in nature and for determining whether self-inoculation of similar reactors is likely to occur at other sites. In this study (13)C-DNA-stable isotope probing (SIP) was used to identify TBA-utilizing organisms in samples of self-inoculated BioGAC reactors operated at sites in New York and California. Based on 16S rRNA nucleotide sequences, all TBA-utilizing organisms identified were members of the Burkholderiales order of the β-proteobacteria. Organisms similar to Cupriavidus and Methylibium were observed in both reactor samples while organisms similar to Polaromonas and Rhodoferax were unique to the reactor sample from New York. Organisms similar to Hydrogenophaga and Paucibacter strains were only detected in the reactor sample from California. We also analyzed our samples for the presence of several genes previously implicated in TBA oxidation by pure cultures of bacteria. Genes Mpe_B0532, B0541, B0555, and B0561 were all detected in (13)C-metagenomic DNA from both reactors and deduced amino acid sequences suggested these genes all encode highly conserved enzymes. One gene (Mpe_B0555) encodes a putative phthalate dioxygenase-like enzyme that may be particularly appropriate for determining the potential for TBA oxidation in contaminated environmental samples.

Isotope correlations for safeguards surveillance and accountancy methods

International Nuclear Information System (INIS)

Persiani, P.J.; Kalimullah.

1982-01-01

Isotope correlations corroborated by experiments, coupled with measurement methods for nuclear material in the fuel cycle have the potential as a safeguards surveillance and accountancy system. The ICT allows the verification of: fabricator's uranium and plutonium content specifications, shipper/receiver differences between fabricator output and reactor input, reactor plant inventory changes, reprocessing batch specifications and shipper/receiver differences between reactor output and reprocessing plant input. The investigation indicates that there exist predictable functional relationships (i.e. correlations) between isotopic concentrations over a range of burnup. Several cross-correlations serve to establish the initial fuel assembly-averaged compositions. The selection of the more effective correlations will depend not only on the level of reliability of ICT for verification, but also on the capability, accuracy and difficulty of developing measurement methods. The propagation of measurement errors through the correlations have been examined to identify the sensitivity of the isotope correlations to measurement errors, and to establish criteria for measurement accuracy in the development and selection of measurement methods. 6 figures, 3 tables

Massive mercury target for thallium isotope production on the beam of high energy protons

International Nuclear Information System (INIS)

Novgorodov, A.F.; Kolachkovski, A.; Nguen Kong Chang.

1980-01-01

The yields of thallium radioisotopes in a massive mercury target irradiated with 660 MeV protons have been determined. The constancy of isotopic composition of radiothallium along the whole length (40 cm) of the target has been found. The yields of 200 Tl, 201 Tl and 202 Tl amount to 22.9+-2.8; 3.42+-0.45 and 0.459+-0.61 mCu/mkA h, respectively. It has been shown that the extraction of radioisotopes of thallium and some other elements from large amounts of mercury as well as their subsequent concentration may be carried out fully and relatavely fast when using dilute solutions of acetic acid

Study of short-lived fission products with the aid of an isotope separator connected to reactor R2-0

International Nuclear Information System (INIS)

Rudstam, G.

1976-01-01

This report constitutes a final report on project 74-3289 together with a preliminary report for project 75-3332. These projects have been included in the budget years 1974/75 and 1975/76 as a contribution to the operating costs of reactor R2-0 at Studsvik. The reactor was used for experimental studies on short-lived fission products with OSIRIS isotope-separator equipment. The scientific programme is very broad. It comprises, in the first place, characterisation of fission products (a study of their excitation levels, measurement of decay properties such as half-life and emission of delayed neutrons, determination of neutron energy spectrum, determination of total decay energy, etc.). An important application of this field of research is the determination of decay heat in nuclear fuel. The programme thus comprises research of a fundamental character and applied research. (H.E.G.)

Report of the ANS Project Feasibility Workshop for a High Flux Isotope Reactor-Center for Neutron Research Facility

International Nuclear Information System (INIS)

Peretz, F.J.; Booth, R.S.

1995-07-01

The Advanced Neutron Source (ANS) Conceptual Design Report (CDR) and its subsequent updates provided definitive design, cost, and schedule estimates for the entire ANS Project. A recent update to this estimate of the total project cost for this facility was $2.9 billion, as specified in the FY 1996 Congressional data sheet, reflecting a line-item start in FY 1995. In December 1994, ANS management decided to prepare a significantly lower-cost option for a research facility based on ANS which could be considered during FY 1997 budget deliberations if DOE or Congressional planners wished. A cost reduction for ANS of about $1 billion was desired for this new option. It was decided that such a cost reduction could be achieved only by a significant reduction in the ANS research scope and by maximum, cost-effective use of existing High Flux Isotope Reactor (HFIR) and ORNL facilities to minimize the need for new buildings. However, two central missions of the ANS -- neutron scattering research and isotope production-were to be retained. The title selected for this new option was High Flux Isotope Reactor-Center for Neutron Research (HFIR-CNR) because of the project's maximum use of existing HFIR facilities and retention of selected, central ANS missions. Assuming this shared-facility requirement would necessitate construction work near HFIR, it was specified that HFIR-CNR construction should not disrupt normal operation of HFIR. Additional objectives of the study were that it be highly credible and that any material that might be needed for US Department of Energy (DOE) and Congressional deliberations be produced quickly using minimum project resources. This requirement made it necessary to rely heavily on the ANS design, cost, and schedule baselines. A workshop methodology was selected because assessment of each cost and/or scope-reduction idea required nearly continuous communication among project personnel to ensure that all ramifications of propsed changes

Accurate isotope ratio mass spectrometry. Some problems and possibilities

International Nuclear Information System (INIS)

Bievre, P. de

1978-01-01

The review includes reference to 190 papers, mainly published during the last 10 years. It covers the following: important factors in accurate isotope ratio measurements (precision and accuracy of isotope ratio measurements -exemplified by determinations of 235 U/ 238 U and of other elements including 239 Pu/ 240 Pu; isotope fractionation -exemplified by curves for Rb, U); applications (atomic weights); the Oklo natural nuclear reactor (discovered by UF 6 mass spectrometry at Pierrelatte); nuclear and other constants; isotope ratio measurements in nuclear geology and isotope cosmology - accurate age determination; isotope ratio measurements on very small samples - archaeometry; isotope dilution; miscellaneous applications; and future prospects. (U.K.)

Probability of fracture and life extension estimate of the high-flux isotope reactor vessel

International Nuclear Information System (INIS)

Chang, S.J.

1998-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in ductile-brittle transition temperature (DBTT) for fracture, often denoted by RT NDT for carbon steel. This transition temperature can be calibrated by the drop-weight test and, sometimes, by the Charpy impact test. The life extension for the high-flux isotope reactor (HFIR) vessel is calculated by using the method of fracture mechanics that is incorporated with the effect of the DBTT change. The failure probability of the HFIR vessel is limited as the life of the vessel by the reactor core melt probability of 10 -4 . The operating safety of the reactor is ensured by periodic hydrostatic pressure test (hydrotest). The hydrotest is performed in order to determine a safe vessel static pressure. The fracture probability as a result of the hydrostatic pressure test is calculated and is used to determine the life of the vessel. Failure to perform hydrotest imposes the limit on the life of the vessel. The conventional method of fracture probability calculations such as that used by the NRC-sponsored PRAISE CODE and the FAVOR CODE developed in this Laboratory are based on the Monte Carlo simulation. Heavy computations are required. An alternative method of fracture probability calculation by direct probability integration is developed in this paper. The present approach offers simple and expedient ways to obtain numerical results without losing any generality. In this paper, numerical results on (1) the probability of vessel fracture, (2) the hydrotest time interval, and (3) the hydrotest pressure as a result of the DBTT increase are obtained

Fast Reactors and Nuclear Nonproliferation

International Nuclear Information System (INIS)

Avrorina, E.N.; Chebeskovb, A.N.

2013-01-01

Conclusion remarks: 1. Fast reactor start-up with U-Pu fuel: – dependent on thermal reactors, – no needs in U enrichment, – needs in SNF reprocessing, – Pu is a little suitable for NED, – practically impossible gun-type NED, – difficulties for implosion-type NED: necessary tests, advanced technologies, etc. – Pu in blankets is similar to WPu by isotopic composition, – Use of blanket for production isotopes (e.g. 233 U), – Combined reprocessing of SNF: altogether blanket and core, – Blanket elimination: decrease in Pu production – No pure Pu separation. 2. Fast reactor start-up with U fuel: - Needs in both U enrichment and SNF reprocessing, - Independent of thermal reactors, - Good Pu bred in the core let alone blankets, - NED of simple gun-type design, - Increase of needs in SWU, - Increased demands in U supply. 3. Fast reactors for export: - Uranium shortage, - To replace thermal reactors in future, - No blankets (depends on the country, though), - Fuel supply and SNF take back, - International centers for rendering services of NFC. Time has come to remove from FRs and their NFC the label unfairly identifying them as the most dangerous installations of nuclear power from the standpoint of being a proliferation problem

Nuclear Burning Wave Modular Fast Reactor Concept

International Nuclear Information System (INIS)

Kodochigov, N.G.; Sukharev, Yu.P.

2014-01-01

The necessity to provide nuclear power industry, comparable in a scope with power industry based on a traditional fuel, inspired studies of an open-cycle fast reactor aimed at: - solution of the problem of fuel provision by implementing the highest breeding characteristics of new fissile materials of raw isotopes in a fast reactor and applying accumulated fissile isotopes in the same reactor, independently on a spent fuel reprocessing rate in the external fuel cycle; - application of natural or depleted uranium for makeup fuel, which, with no spent fuel reprocessing, forms the most favorable non-proliferation conditions; - application of inherent properties of the core and reactor for safety provision. The present report, based on previously published papers, gives the theoretical backgrounds of the concept of the reactor with a nuclear burning wave, in which an enriched-fuel core (driver) is replaced by a blanket, and basic conditions for nuclear burning wave initiating and keeping are shown. (author)

Isotope production

Energy Technology Data Exchange (ETDEWEB)

Lewis, Dewi M.

1995-07-15

Some 2 0% of patients using radiopharmaceuticals receive injections of materials produced by cyclotrons. There are over 200 cyclotrons worldwide; around 35 are operated by commercial companies solely for the production of radio-pharmaceuticals with another 25 accelerators producing medically useful isotopes. These neutron-deficient isotopes are usually produced by proton bombardment. All commonly used medical isotopes can be generated by 'compact' cyclotrons with energies up to 40 MeV and beam intensities in the range 50 to 400 microamps. Specially designed target systems contain gram-quantities of highly enriched stable isotopes as starting materials. The targets can accommodate the high power densities of the proton beams and are designed for automated remote handling. The complete manufacturing cycle includes large-scale target production, isotope generation by cyclotron beam bombardment, radio-chemical extraction, pharmaceutical dispensing, raw material recovery, and labelling/packaging prior to the rapid delivery of these short-lived products. All these manufacturing steps adhere to the pharmaceutical industry standards of Good Manufacturing Practice (GMP). Unlike research accelerators, commercial cyclotrons are customized 'compact' machines usually supplied by specialist companies such as IBA (Belgium), EBCO (Canada) or Scanditronix (Sweden). The design criteria for these commercial cyclotrons are - small magnet dimensions, power-efficient operation of magnet and radiofrequency systems, high intensity extracted proton beams, well defined beam size and automated computer control. Performance requirements include rapid startup and shutdown, high reliability to support the daily production of short-lived isotopes and low maintenance to minimize the radiation dose to personnel. In 1987 a major step forward in meeting these exacting industrial requirements came when IBA, together with the University of Louvain-La-Neuve in Belgium, developed the Cyclone-30

Validation of a new design of tellurium dioide-irradiated target

Energy Technology Data Exchange (ETDEWEB)

Fllaoui, Aziz; Ghamad, Younes; Zoubir, Brahim; Ayaz, Zinel Abidine; El Morabiti, Aissam; Amayoud, Hafid [Centre National de l' Energie des Sciences et des Techniques Nucleaires, Rabat (Morocco); Chakir, El Mahjoub [Nuclear Physics Department, University Ibn Toufail, Kenitra (Morocco)

2016-10-15

Production of iodine-131 by neutron activation of tellurium in tellurium dioxide (TeO{sub 2}) material requires a target that meets the safety requirements. In a radiopharmaceutical production unit, a new lid for a can was designed, which permits tight sealing of the target by using tungsten inert gas welding. The leakage rate of all prepared targets was assessed using a helium mass spectrometer. The accepted leakage rate is ≤ 10 - 4 mbr.L/s, according to the approved safety report related to iodine-131 production in the TRIGA Mark II research reactor (TRIGA: Training, Research, Isotopes, General Atomics). To confirm the resistance of the new design to the irradiation conditions in the TRIGA Mark II research reactor's central thimble, a study of heat effect on the sealed targets for 7 hours in an oven was conducted and the leakage rates were evaluated. The results show that the tightness of the targets is ensured up to 600 .deg. C with the appearance of deformations on lids beyond 450 .deg. C. The study of heat transfer through the target was conducted by adopting a one-dimensional approximation, under consideration of the three transfer modes-convection, conduction, and radiation. The quantities of heat generated by gamma and neutron heating were calculated by a validated computational model for the neutronic simulation of the TRIGA Mark II research reactor using the Monte Carlo N-Particle transport code. Using the heat transfer equations according to the three modes of heat transfer, the thermal study of I-131 production by irradiation of the target in the central thimble showed that the temperatures of materials do not exceed the corresponding melting points. To validate this new design, several targets have been irradiated in the central thimble according to a preplanned irradiation program, going from 4 hours of irradiation at a power level of 0.5 MW up to 35 hours (7 h/d for 5 days a week) at 1.5 MW. The results show that the irradiated targets are

Validation of a New Design of Tellurium Dioxide-Irradiated Target

Directory of Open Access Journals (Sweden)

Aziz Fllaoui

2016-10-01

Full Text Available Production of iodine-131 by neutron activation of tellurium in tellurium dioxide (TeO2 material requires a target that meets the safety requirements. In a radiopharmaceutical production unit, a new lid for a can was designed, which permits tight sealing of the target by using tungsten inert gas welding. The leakage rate of all prepared targets was assessed using a helium mass spectrometer. The accepted leakage rate is ≤ 10−4 mbr.L/s, according to the approved safety report related to iodine-131 production in the TRIGA Mark II research reactor (TRIGA: Training, Research, Isotopes, General Atomics. To confirm the resistance of the new design to the irradiation conditions in the TRIGA Mark II research reactor's central thimble, a study of heat effect on the sealed targets for 7 hours in an oven was conducted and the leakage rates were evaluated. The results show that the tightness of the targets is ensured up to 600°C with the appearance of deformations on lids beyond 450°C. The study of heat transfer through the target was conducted by adopting a one-dimensional approximation, under consideration of the three transfer modes—convection, conduction, and radiation. The quantities of heat generated by gamma and neutron heating were calculated by a validated computational model for the neutronic simulation of the TRIGA Mark II research reactor using the Monte Carlo N-Particle transport code. Using the heat transfer equations according to the three modes of heat transfer, the thermal study of I-131 production by irradiation of the target in the central thimble showed that the temperatures of materials do not exceed the corresponding melting points. To validate this new design, several targets have been irradiated in the central thimble according to a preplanned irradiation program, going from 4 hours of irradiation at a power level of 0.5 MW up to 35 hours (7 h/d for 5 days a week at 1.5 MW. The results show that the irradiated targets are

Comparison of short-lived medical isotopes activation by laser thin target induced protons and conventional cyclotron proton beams

Science.gov (United States)

Murray, Joseph; Dudnikova, Galina; Liu, Tung-Chang; Papadopoulos, Dennis; Sagdeev, Roald; Su, J. J.; UMD MicroPET Team

2014-10-01

Production diagnostic or therapeutic nuclear medicines are either by nuclear reactors or by ion accelerators. In general, diagnostic nuclear radioisotopes have a very short half-life varying from tens of minutes for PET tracers and few hours for SPECT tracers. Thus supplies of PET and SPECT radiotracers are limited by regional production facilities. For example 18F-fluorodeoxyglucose (FDG) is the most desired tracer for positron emission tomography because its 110 minutes half-life is sufficient long for transport from production facilities to nearby users. From nuclear activation to completing image taking must be done within 4 hours. Decentralized production of diagnostic radioisotopes will be idea to make high specific activity radiotracers available to researches and clinicians. 11 C, 13 N, 15 O and 18 F can be produced in the energy range from 10-20 MeV by protons. Protons of energies up to tens of MeV generated by intense laser interacting with hydrogen containing targets have been demonstrated by many groups in the past decade. We use 2D PIC code for proton acceleration, Geant4 Monte Carlo code for nuclei activation to compare the yields and specific activities of short-lived isotopes produced by cyclotron proton beams and laser driven protons.

Sensitivity of reactor integral parameters to #betta##betta# parameter of resolved resonances of fertile isotopes and to the α values, in thermal and epithermal spectra

International Nuclear Information System (INIS)

Barroso, D.E.G.

1982-01-01

A sensitivity analysis of reactor integral parameter to more 10% variation in the resolved resonance parameters #betta##betta# of the fertile isotope and the variations of more 10% in the α values (#betta# sub(#betta#)/#betta# sub(f)) of fissile isotopes of PWR fuel elements, is done. The analysis is made with thermal and epithermal spectra, those last generated in a fuel cell with low V sub(M)/V sub(F). The HAMMER system, the interface programs HELP and LITHE and the HAMMER computer codes, were used as a base for this study. (E.G.) [pt

Probabilistic fracture mechanics analysis for the life extension estimate of the high flux isotope reactor vessel

International Nuclear Information System (INIS)

Chang, S.J.

1997-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A new method of fracture probability calculation is presented in this paper. The fracture probability as a result of the hydrostatic pressure test (hydrotest) is used to determine the life of the vessel. The hydrotest is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the safe reactor operation time from the time of the satisfactory hydrostatic test. The life extension calculation provides the following information on the remaining life of the reactor as a function of the NDT increase: (1) the life of the vessel is determined by the probability of vessel fracture as a result of hydrotest at several hydrotest pressures and vessel embrittlement conditions, (2) the hydrotest time interval vs the NDT increase rate, and (3) the hydrotest pressure vs the NDT increase rate. It is understood that the use of a complete range of uncertainties of the NDT increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85 degrees F

The Texts of the Instruments connected with the Agency's Assistance to Argentina in Establishing a Research and Isotope Production Reactor Project

Energy Technology Data Exchange (ETDEWEB)

NONE

1965-11-04

The texts of the Title Transfer Agreement between the Agency and the Governments of Argentina and the United States of America, and of the Project Agreement between the Agency and the Government of Argentina, in connection with the Agency's assistance to that Government in establishing a research and isotope production reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 2 December 1964.

The Texts of the Instruments connected with the Agency's Assistance to Argentina in Establishing a Research and Isotope Production Reactor Project

International Nuclear Information System (INIS)

1965-01-01

The texts of the Title Transfer Agreement between the Agency and the Governments of Argentina and the United States of America, and of the Project Agreement between the Agency and the Government of Argentina, in connection with the Agency's assistance to that Government in establishing a research and isotope production reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 2 December 1964

Delivery of completed irradiation vehicles and the quality assurance document to the High Flux Isotope Reactor for irradiation

Energy Technology Data Exchange (ETDEWEB)

Petrie, Christian M. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); McDuffee, Joel Lee [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

2015-10-01

This report details the initial fabrication and delivery of two Fuel Cycle Research and Development (FCRD) irradiation capsules (ATFSC01 and ATFSC02), with associated quality assurance documentation, to the High Flux Isotope Reactor (HFIR). The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0202268. These irradiation experiments are testing silicon carbide composite tubes in order to obtain experimental validation of thermo-mechanical models of stress states in SiC cladding irradiated under a prototypic high heat flux. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulics and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division (RNSD). A complete fabrication package record is maintained by the THIEG and is available upon request.

Identification of Thioredoxin Disulfide Targets Using a Quantitative Proteomics Approach Based on Isotope-Coded Affinity Tags

DEFF Research Database (Denmark)

Hägglund, Per; Bunkenborg, Jakob; Maeda, Kenji

2008-01-01

Thioredoxin (Trx) is a ubiquitous protein disulfide reductase involved in a wide range of cellular redox processes. A large number of putative target proteins have been identified using proteomics approaches, but insight into target specificity at the molecular level is lacking since the reactivity...... of Trx toward individual disulfides has not been quantified. Here, a novel proteomics procedure is described for quantification of Trx-mediated target disulfide reduction based on thiol-specific differential labeling with the iodoacetamide-based isotope-coded affinity tag (ICAT) reagents. Briefly......, protein extract of embryos from germinated barley seeds was treated +/- Trx, and thiols released from target protein disulfides were irreversibly blocked with iodoacetamide. The remaining cysteine residues in the Trx-treated and the control (-Trx) samples were then chemically reduced and labeled...

The High Flux Isotope Reactor (HFIR) cold source project at ORNL

International Nuclear Information System (INIS)

Selby, D.L.; Lucas, A.T.; Chang, S.J.; Freels, J.D. . E-mail-yb2@ornl.gov

1998-01-01

Following the decision to cancel the Advanced Neutron Source (ANS) Project at Oak Ridge National Laboratory (ORNL), it was determined that a hydrogen cold source should be retrofitted into an existing beam tube of the High Flux Isotope Reactor (HFIR) at ORNL. The preliminary design of this system has been completed and an 'approval in principle' of the design has been obtained from the internal ORNL safety review committees and the U.S. Department of Energy (DOE) safety review committee. The cold source concept is basically a closed loop forced flow supercritical hydrogen system. The supercritical approach was chosen because of its enhanced stability in the proposed high heat flux regions. Neutron and gamma physics of the moderator have been analyzed using the 3D Monte Carlo code MCNP 1 A D structural analysis model of the moderator vessel, vacuum tube, and beam tube was completed to evaluate stress loadings and to examine the impact of hydrogen detonations in the beam tube. A detailed ATHENA 2 system model of the hydrogen system has been developed to simulate loop performance under normal and off-normal transient conditions. Semi-prototypic hydrogen loop tests of the system have been performed at the Arnold Engineering Design Center (AEDC) located in Tullahoma, Tennessee to verify the design and benchmark the analytical system model. A 3.5 kW refrigerator system has been ordered and is expected to be delivered to ORNL by the end of this calendar year. Our present schedule shows the assembling of the cold source loop on site during the fall of 1999 for final testing before insertion of the moderator plug assembly into the reactor beam tube during the end of the year 2000. (author)

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

Energy Technology Data Exchange (ETDEWEB)

Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

2006-02-01

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

International Nuclear Information System (INIS)

Primm, R.T. III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

2006-01-01

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U 3 O 8 mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties

Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

International Nuclear Information System (INIS)

Behar, Christophe

2014-01-01

Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

New nuclear technologies will help to ensure the public trust and further development of research reactors

International Nuclear Information System (INIS)

Miasnikov, S.V.

2001-01-01

Decrease of public trust to research reactors causes the concern of experts working in this field. In the paper the reasons of public mistrust to research reactors are given. A new technology of 99 Mo production in the 'Argus' solution reactor developed in the Russian Research Centre 'Kurchatov Institute' is presented as an example assisting to eliminate these reasons. 99 Mo is the most widespread and important medical isotope. The product received employing a new technology completely meets the international specifications. Besides, the proposed technology raises the efficiency of 235 U consumption practically up to 100% and allows using a reactor with power 10 and more times lower than that in the target technology. The developed technology meets the requirements of the community to nuclear safety of manufacture, reduction of radioactive waste and non-proliferation of nuclear materials. (author)

Pallas: the new nuclear reactor in the Netherlands

International Nuclear Information System (INIS)

De Jong, P.G.T.; Van Der Schaaf, B.; Schrijver, J.M.

2010-01-01

In the European Union, the first generation research reactors are approaching necessary operational retirement. Maintenance costs are increasing and continuity of operations is compromised by the aging of materials and components. The High Flux Reactor (HFR) in Petten, The Netherlands, is one such reactor. Nuclear Research and Consultancy Group (NRG), the current licence holder and operator of the HFR, therefore plans to build a new research reactor called PALLAS. This will be a state-of-the-art reactor equipped to meet the growing world demand for both nuclear knowledge and services and the production of essential medical isotopes. It will have the capacity to be the world's biggest producer of such isotopes. The tender process for PALLAS began in 2007 and will continue through 2010- 2011, following the EU rules for competitive tendering of complex, one-off design and construction projects. NRG is currently still actively pursuing the acquisition of the funding for the project. In the exploitation of PALLAS there will be both public and private interests. Public interests have to do with research for sustainable energy and with guaranteed availability of medical isotopes for the treatment of patients. Private interests are focused on commercial irradiations and the production of isotopes. Currently it is expected that the design phase will have to be almost fully public funded NRG welcomes the cabinet-council's recent support for the building of a new reactor and is fortunate in having fast growing public acceptance and support for it too. The licensing process began in autumn 2009 with a, so called, Notification of Intent to conduct an Environmental Impact Assessment (EIA) for PALLAS. Public hearings have been held to inform the national EIA committee's approach to consideration of the Impact Assessment. The PALLAS project team in Petten will guide the design and construction processes, is responsible for the licensing and commissioning and will manage the design

Medical Isotopes Production Project: Molybdenum-99 and related isotopes: Environmental Impact Statement, Volume I

International Nuclear Information System (INIS)

1996-04-01

This Environmental Impact Statement (EIS) provides environmental and technical information concerning the U.S. Department of Energy's (DOE) proposal to establish a domestic source to produce molybdenum-99 (Mo-99) and related medical isotopes (iodine-131, xenon-133 and iodine-125). Mo-99, a radioactive isotope of the element molybdenum, decays to form metastable technetium-99 (Tc-99m), a radioactive isotope used thousands of times daily in medical diagnostic procedures in the U.S. Currently, all Mo-99 used in the U.S. is obtained from a single Canadian source. DOE is pursuing the Medical Isotopes Production Project in order to ensure that a reliable supply of Mo-99 is available to the U.S. medical community. Under DOE's preferred alternative, the Chemistry and Metallurgy Research Facility at the Los Alamos National Laboratory (LANL) and the Annular Core Research Reactor and Hot Cell Facility at Sandia National Laboratories/New Mexico (SNL/NM) would be used for production of the medical isotopes. In addition to the preferred alternative, three other reasonable alternatives and a no action alternative are analyzed in detail. The sites for the three reasonable alternatives are LANL, Oak Ridge National Laboratory (ORNL), and Idaho National Engineering Laboratory (INEL). The analyses in this EIS indicate no significant difference in the potential environmental impacts among the alternatives. Each of the alternatives would use essentially the same technology for the production of the medical isotopes. Minor differences in environmental impacts among alternatives relate to the extent of activity necessary to modify and restart (as necessary) existing reactors and hot cell facilities at each of the sites, the quantities, of low-level radioactive waste generated, how such waste would be managed, and the length of time needed for initial and full production capacity

SFCOMPO 2.0 – A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

Directory of Open Access Journals (Sweden)

Michel-Sendis Franco

2017-01-01

Full Text Available SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA. The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

Broad-Application Test Reactor

International Nuclear Information System (INIS)

Motloch, C.G.

1992-05-01

This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

About the possibility of use of different types of targets as a neutron source for subcritical nuclear reactor driven by particle beam accelerator

Energy Technology Data Exchange (ETDEWEB)

Avdeev, E.F.; Dorokhovich, S.L.; Chusov, I.A. [Obninsk Institute of Nuclear Power Engineering (Russian Federation)

1995-10-01

The schemes of jet gas and liquid targets as well as the gastargets with a solid phase dispersion are introduced to use to receive the neutrons admitted to a subcritical reactor core. The possible variants of target position in the reactor are considered, target characteristics are calculated. The authors pay a great attention to the estimation of radioactive products yield receiving due to the interaction of the beam with the target.

Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

International Nuclear Information System (INIS)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S.

2011-01-01

A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched ( 235 U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1985-01-01

This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

The centenary of the discovery of isotopes

International Nuclear Information System (INIS)

Soulie, Edgar

2013-01-01

This article recalls works performed by different scientists (Marckwald and Keetman, Stromholm and Svedberg, Soddy, Thompson, Aston) which resulted in the observation and identification of the existence of isotopes. The author also recalls various works related to mechanisms of production of isotopes, the discovery of uranium fission and the principle of chain reaction. The author notably evokes French scientists involved in the development of mass spectroscopy and in the research and applications on isotopes within the CEA after the Second World War. A bibliography of article and books published by one of them, Etienne Roth, is provided. References deal with nuclear applications of chemical engineering (heavy water and its production, chemical processes in fission reactors, tritium extraction and enrichment), isotopic fractioning and physical-chemical processes, mass spectrometry and isotopic analysis, isotopic geochemistry (on 07;Earth, search for deuterium in moon rocks and their consequences), first dating and the Oklo phenomenon, radioactive dating, water and climate (isotopic hydrology, isotopes and hailstone formation, the atmosphere), and miscellaneous scientific fields (nuclear measurements and radioactivity, isotopic abundances and atomic weight, isotopic separation and use of steady isotopes)

Isotope products manufacture in Russia and its prospects

International Nuclear Information System (INIS)

Malyshev, S.V.; Okhotina, I.A.; Kalelin, E.A.; Krasnov, N.N.; Kuzin, V.V.; Malykh, J.A.; Makarovsky, S.B.

1997-01-01

At the present stage of the world economy development, stable and radioactive isotopes,preparations and products on their base are widely used in many fields of the national economy, medicine and scientific researches. The Russian Federation is one of the largest worldwide producers of a variety of nuclide products on the base of more than 350 isotopes, as follows: stable isotopes reactor, cyclotron, fission product radioactive isotopes, ion-radiation sources compounds, labelled with stable and radioactive isotopes, radionuclide short-lived isotope generators, radiopharmaceuticals, radionuclide light and heat sources; luminous paints on base of isotopes. The Russian Ministry for Atomic Energy coordinates activity for development and organization of manufacture and isotope products supply in Russia as well as for export. Within many years of isotope industry development, there have appeared some manufacturing centres in Russia, dealing with a variety of isotope products. The report presents the production potentialities of these centres and also an outlook on isotope production development in Russia in the next years

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

Energy Technology Data Exchange (ETDEWEB)

Travis, Adam R [ORNL

2014-05-01

A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

Feasibility neutronic conceptual design for the core configuration of a 75 kWth Aqueous Homogeneous Reactor for 99Mo production

International Nuclear Information System (INIS)

Milian, D.; Milian, D. E.; Rodriguez, L. P.; Salomon, J.; Cadavid, N.

2015-01-01

99m Tc is a very useful radioisotope, which is used in nearly 80% of all nuclear medicine procedures. 99m Tc is produced from 99 Mo decay. Since 2007 the medical community has been plagued by 99 Mo shortages due to aging reactors, such as the National Research Universal reactor in Canada and the High Flux Reactor in Petten, The Netherlands. At present, most of the world's supply of 99 Mo for medical isotope production involves the neutron fission of 235 U in multipurpose research reactors. 99 Mo mostly results from the fission reaction of 235 U targets with a fission yield about 6.1%. After irradiation in the reactor, the target is digested in acid or alkaline solutions and 99 Mo is recovered through a series of extraction (separation) and purification steps. 99 Mo production system in an Aqueous Homogeneous Reactor (AHR) offers a better method, because all of the 99 Mo can be extracted from the fuel solution. Over 30 AHRs has been built and operated around the world with 149 years of combined experience. In this paper, an AHR conceptual design using LEU (Low Enriched Uranium) is optimized to meet the South American demand for 99 Mo for the coming years. Aspect related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotope production and others are evaluated. The neutronic calculations have been performed with the well-known MCNPX computational code. A benchmarking experiments performed at the Russian Research Center 'Kurchatov Institute' in order to validate that the developed models of AHRs with MCNPX code and the available library in XSDIR, ENDF/B VI.2, are adequate for studies of aqueous fuel solutions. (Author)

Overview on recent developments: alternative isotope production methods in Canada

International Nuclear Information System (INIS)

Huynh, K.

2012-01-01

The purpose of this paper is to provide an update on the Government of Canada's programs in alternative isotope production methods for securing supply of technetium 99m for Canadians. The supply disruptions of isotopes in 2007 and 2009/2010 caused by unplanned outages at AECL's National Research Universal (NRU) reactor highlighted the fragility of the supply chain that delivers medical isotopes, specifically Technetium 99m (Tc99m) to patients in Canada and globally. Tc99m, which is derived from its parent, molybdenum99 (Mo99) is the most widely used medical isotope for imaging, and accounts for 80 percent of nuclear medicine diagnostic procedures. Prior to the outage, nearly all the Mo99 produced for the world market came from five aging government owned research reactors in Canada, France, the Netherlands, Belgium and South Africa. The NRU, the largest of these, produced about 30 to 40 percent of the world supply of isotopes prior to 2009 - since its return to service in 2010, its world market share is estimated at 15 to 20%.

Efficiency of an LBE spallation target in an accelerator-driven molten salt subcritical reactor

Energy Technology Data Exchange (ETDEWEB)

Bak, Sang-In [Sungkyunkwan University, Suwon (Korea, Republic of); Hong, Seung-Woo [Sungkyunkwan University, Suwon (Korea, Republic of); Kadi, Yacine [CERN, Geneva (Switzerland)

2016-10-15

An Accelerator-Driven System (ADS) combined with a subcritical Molten Salt Reactor (MSR) is a type of hybrid reactor originally designed to breed uranium from thorium or to incinerate long-lived minor actinides in nuclear wastes. In an MSR, the salt material is used not only as a nuclear fuel but also as a primary coolant. In addition, this material is used as a target for inducing spallation neutrons in most AD-MSR concepts. A high energy proton beam impinges on a heavy metal target to induce spallation reactions and produces neutrons. Accordingly, a reliable proton accelerator is needed to feed the source neutrons. As ADSs have been criticized for requiring high power accelerators, minimization of beam power is an important aspect of ADS design. A primary concern associated with ADS development is stable high-power accelerators. We therefore studied the neutron source efficiencies of an AD-MSR involving chloride fuels by including a Pb-Bi eutectic (LBE) spallation target. The proton source efficiency and the accelerator beam power required have been studied for an AD-MSR. Adoption of an LBE spallation target induces an increase in proton source efficiencies in comparison to the case without a spallation target. Thus the presence of an efficient spallation target is useful in the reduction of the beam power of an accelerator. Almost 33 % of the beam power can be reduced in comparison to the case without the target for NaCl-Th/{sup 233}U fuel, and about 16 % for NaCl-U/TRU fuel. The beam power amplifications increase by 1.5 times for NaCl-Th/{sup 233}U and 1.2 times for NaCl-U/TRU in comparison with the no target AD-MSR.

Experiences in controlling the upgrading of TRIGA 2000 Bandung reactor

International Nuclear Information System (INIS)

Huda, K.; Wibowo, Y.W.; Suprawhardana, M.S.

2001-01-01

TRIGA 2000 Bandung Reactor was established in 1961 for research, education and isotope production purposes. The reactor reached its first criticality in October 1964 and operated at nominal power of 250 kW until 1971. In 1971 the reactor was upgraded to the power level of 1000 kW. In order to raise the capacity of isotope production, the reactor has been upgraded again to the power level of 2000 kW. During the modification of the reactor, the Center for Research and Development of Nuclear Techniques (CRDNT) was management of the reactor as it faced many problems, either technical or non-technical ones. This caused the upgrading activities to take a long time. At this time, the reactor upgrading has almost finished, and the nuclear commissioning is going on. Several aspects and problems associated with the upgrading process have been reviewed and the results are discussed in the present paper. (author)

Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched (<20% {sup 235}U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

Radiative capture on $^{242}$Pu for MOX fuel reactors

CERN Multimedia

The use of MOX fuel (mixed-oxide fuel made of UO$_{2}$ and PuO$_{2}$) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. Indeed around 66% of the plutonium from spent fuel is made of $^{239}$Pu and $^{241}$Pu, which are fissile in thermal reactors. A typical reactor of this type uses a fuel with 7% reprocessed Pu and 93% depleted U, thus profiting from both the spent fuel and the remaining $^{238}$U following the $^{235}$U enrichment. With the use of such new fuel compositions rich in Pu the better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. This is clearly stated in the recent OECD NEA’s “High Priority Request List” and in the WPEC-26 “Uncertainty and target accuracy assessment for innovative systems using recent covariance data evaluations” report. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United ...

Verification of maximum radial power peaking factor due to insertion of FPM-LEU target in the core of RSG-GAS reactor

Energy Technology Data Exchange (ETDEWEB)

Setyawan, Daddy, E-mail: d.setyawan@bapeten.go.id [Center for Assessment of Regulatory System and Technology for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia); Rohman, Budi [Licensing Directorate for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia)

2014-09-30

Verification of Maximum Radial Power Peaking Factor due to insertion of FPM-LEU target in the core of RSG-GAS Reactor. Radial Power Peaking Factor in RSG-GAS Reactor is a very important parameter for the safety of RSG-GAS reactor during operation. Data of radial power peaking factor due to the insertion of Fission Product Molybdenum with Low Enriched Uranium (FPM-LEU) was reported by PRSG to BAPETEN through the Safety Analysis Report RSG-GAS for FPM-LEU target irradiation. In order to support the evaluation of the Safety Analysis Report incorporated in the submission, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the maximum radial power peaking factor change due to the insertion of FPM-LEU target in RSG-GAS Reactor by computational method using MCNP5and ORIGEN2. From the results of calculations, the new maximum value of the radial power peaking factor due to the insertion of FPM-LEU target is 1.27. The results of calculations in this study showed a smaller value than 1.4 the limit allowed in the SAR.

Medical Isotopes Production Project: Molybdenum-99 and related isotopes - environmental impact statement. Volume II, comment response document

International Nuclear Information System (INIS)

1996-04-01

This Environmental Impact Statement (EIS) provides environmental and technical information concerning the U.S. Department of Energy's (DOE) proposal to establish a domestic source to produce molybdenum-99 (Mo-99) and related isotopes (iodine-131, xenon-133, and iodine-125). Mo-99, a radioactive isotope of the element molybdenum, decays to form metastable technetium-99 (Tc-99m), a radioactive isotope used thousands of times daily in medical diagnostic procedures in the U.S. Currently, all Mo-99 used in the U.S. is obtained from a single Canadian source. DOE is pursuing the Medical Isotopes Production Project in order to ensure that a reliable supply of Mo-99 is available to the U.S. medical community as soon as practicable. Under DOE's preferred alternative, the Chemistry and Metallurgy Research Facility at the Los Alamos National Laboratory (LANL) and the Annular Core Research Reactor and Hot Cell Facility at Sandia National Laboratories/New Mexico (SNL/NM) would be used for production of the medical isotopes. In addition, three other reasonable alternatives and a No Action alternative are analyzed in detail, The sites for these three reasonable alternatives are LANL, Oak Ridge National Laboratory (ORNL), and Idaho National Engineering Laboratory (INEL). The analyses in this EIS indicate no significant difference in the potential environmental impacts among the alternatives. Each of the alternatives would use essentially the same technology for the production of the medical isotopes. Minor differences in environmental impacts among alternatives relate to the extent of activity necessary to modify and restart (as necessary) existing reactors and hot cell facilities at each of the sites, the quantities of low-level radioactive waste generated, how such waste would be managed, and the length of time needed for initial and full production capacity. This document contains comments recieved from meetings held regarding the site selection for isotope production

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

Energy Technology Data Exchange (ETDEWEB)

Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

2011-02-01

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

Solid targets for 99mTc production on medical cyclotrons

International Nuclear Information System (INIS)

Hanemaayer, V.; Buckley, K.R.; Klug, J.; Ruth, T.J.; Schaffer, P.; Zeisler, S.K.; Benard, F.; Kovacs, M.; Leon, C.

2014-01-01

Recent disruptions in the molybdenum-technetium generator supply chain prompted a review of non-reactor based production methods for both 99 Mo and 99m Tc. Small medical cyclotrons (E p ∼ 16-24 MeV) are capable of producing Curie quantities of 99m Tc from isotopically enriched 100 Mo using the 100 Mo(p,2n) 99m Tc reaction. Unlike most other metallic target materials for routine production of medical radioisotopes, molybdenum cannot be deposited by reductive electroplating from aqueous salt solutions. To overcome this issue, we developed a new process for solid molybdenum targets based on the electrophoretic deposition of fine 100 Mo powder onto a tantalum plate, followed by high temperature sintering. The targets obtained were mechanically robust and thermally stable when irradiated with protons at high power density. (author)

MUICYCL and MUIFAP: models tracking minor uranium isotopes in the nuclear fuel cycle

International Nuclear Information System (INIS)

Blum, S.R.; McLaren, R.A.

1979-10-01

Two computer programs have been written to provide information on the buildup of minor uranium isotopes in the nuclear fuel cycle. The Minor Uranium Isotope Cycle Program, MUICYCL, tracks fuel through a multiyear campaign cycle of enrichment, reactor burnup, reprocessing, enrichment, etc. MUICYCL facilities include preproduction stockpiles, U 235 escalation, and calculation of losses. The Minor Uranium Isotope Flowsheet Analyzer Program, MUIFAP, analyzes one minor isotope in one year of an enrichment operation. The formulation of the enrichment cascade, reactors, and reprocessing facility is presented. Input and output descriptions and sample cases are presented. The programs themselves are documented by short descriptions of each routine, flowcharts, definitions of common blocks and variables, and internal documentation. The programs are written in FORTRAN for use in batch mode

CRDIAC: Coupled Reactor Depletion Instrument with Automated Control

International Nuclear Information System (INIS)

Logan, Steven K.

2012-01-01

When modeling the behavior of a nuclear reactor over time, it is important to understand how the isotopes in the reactor will change, or transmute, over that time. This is especially important in the reactor fuel itself. Many nuclear physics modeling codes model how particles interact in the system, but do not model this over time. Thus, another code is used in conjunction with the nuclear physics code to accomplish this. In our code, Monte Carlo N-Particle (MCNP) codes and the Multi Reactor Transmutation Analysis Utility (MRTAU) were chosen as the codes to use. In this way, MCNP would produce the reaction rates in the different isotopes present and MRTAU would use cross sections generated from these reaction rates to determine how the mass of each isotope is lost or gained. Between these two codes, the information must be altered and edited for use. For this, a Python 2.7 script was developed to aid the user in getting the information in the correct forms. This newly developed methodology was called the Coupled Reactor Depletion Instrument with Automated Controls (CRDIAC). As is the case in any newly developed methodology for modeling of physical phenomena, CRDIAC needed to be verified against similar methodology and validated against data taken from an experiment, in our case AFIP-3. AFIP-3 was a reduced enrichment plate type fuel tested in the ATR. We verified our methodology against the MCNP Coupled with ORIGEN2 (MCWO) method and validated our work against the Post Irradiation Examination (PIE) data. When compared to MCWO, the difference in concentration of U-235 throughout Cycle 144A was about 1%. When compared to the PIE data, the average bias for end of life U-235 concentration was about 2%. These results from CRDIAC therefore agree with the MCWO and PIE data, validating and verifying CRDIAC. CRDIAC provides an alternative to using ORIGEN-based methodology, which is useful because CRDIAC's depletion code, MRTAU, uses every available isotope in its depletion

Optimization of neutron flux distribution in Isotope Production Reactor

International Nuclear Information System (INIS)

Valladares, G.L.

1988-01-01

In order to optimize the thermal neutrons flux distribution in a Radioisotope Production and Research Reactor, the influence of two reactor parameters was studied, namely the Vmod / Vcomb ratio and the core volume. The reactor core is built with uranium oxide pellets (UO 2 ) mounted in rod clusters, with an enrichment level of ∼3 %, similar to LIGHT WATER POWER REATOR (LWR) fuel elements. (author) [pt

Candidate processes for diluting the 235U isotope in weapons-capable highly enriched uranium

International Nuclear Information System (INIS)

Snider, J.D.

1996-02-01

The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile 235 U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile 235 U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel

Enrichment of 15N and 10B isotopes by chemical exchange process

International Nuclear Information System (INIS)

D'Souza, A.B.; Sonwalkar, A.S.; Subrahmanyam, B.V.; Valladares, B.A.

1994-01-01

Many processes are available for separation of stable isotopes like distillation, chemical exchange, thermal diffusion, gaseous diffusion, centrifuge etc. Chemical exchange process is eminently suitable for separation of isotopes of light elements. Work done on separation and enrichment of two of the stable isotopes viz. 15 N and 10 B in Chemical Engineering Division is presented. 15 N is widely used as a tracer in agricultural research and 10 B is used in nuclear industry as control rod material, soluble reactor poison, neutron detector etc. The work on 15 N isotope resulted in a pilot plant, which was the only source of this material in the country for many years and later it was translated into a production plant as M/s. RCF Ltd. The work done on the ion-exchange process for enrichment of 10 B isotope which is basically a chemical exchange process, is now being updated into a pilot plant to produce enriched 10 B to be used as soluble reactor poison. (author)

Infrastructure for thulium-170 isotope power systems for autonomous underwater vehicle fleets

International Nuclear Information System (INIS)

Walter, C.E.

1991-07-01

The radioisotope thulium-170 is a safe and environmentally benign heat source for providing the high endurance and energy densities needed by advanced power systems for autonomous underwater vehicles (AUV). Thulium Isotope Power (TIP) systems have an endurance of ∼3000 h, and gravimetric and volumetric energy densities of 3 x 10 4 Wh/kg and 3 x 10 8 Wh/m 3 , respectively. These energy densities are more than 200 times higher than those currently provided by Ag-Zn battery technology. In order to capitalize on these performance levels with about one hundred AUVs in continuous use, it will be necessary to establish an infrastructure for isotope production and heat-source refurbishment. The infrastructure cost is not trivial, and studies are needed to determine its optimum configuration. The major component of the projected infrastructure is the nuclear reactor used to produce Tm- 170 by neutron absorption in Tm-169. The reactor design should ideally be optimized for TM-170 production. Using the byproduct ''waste'' heat beneficially would help defray the cost of isotope production. However, generating electric power with the reactor would compromise both the cost of electricity and the isotope production capacity. A coastal location for the reactor would be most convenient from end-use considerations, and the ''waste'' heat could be used to desalinate seawater in water-thirsty states. 13 refs., 6 figs., 2 tabs

The irradiation test program for transmutation in the French Phenix fast reactor

International Nuclear Information System (INIS)

Guidez, J.; Chaucheprat, P.; Fontaine, B.; Brunon, E.

2004-01-01

Put on commercial operation in July 1974, the French fast reactor Phenix reached a 100 000 hours operation time in september 2003. When the French law relative to long lived radioactive waste management was promulgated on December 1991, priority was given to Phenix to be run as a research reactor and to carry on a wide irradiation program dedicated to study transmutation of minor actinides and long-lived fission products. After a major renovation program required to extend the reactor lifetime, Phenix power buildup took place in 2003. Experimental irradiations have been loaded in the core, involving components for heterogeneous and homogeneous transmutation modes, americium targets, technetium 99 metal pins and isolated isotopes for integral cross-sections measurements. Associated post- irradiated examination programs are already underway or planned. With new experiments to be loaded in the core in 2006 the Phenix reactor remains to be a powerful tool providing an important experimental data on fast reactors and on transmutation of minor actinides and long-lived fission products, as well as it will contribute to gain further experience in the framework of the GENERATION IV International Forum. (authors)

Transmutation of Americium in Light and Heavy Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Hyland, B.; Dyck, G.R.; Edwards, G.W.R. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada); Ellis, R.J.; Gehin, J.C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee (United States); Maldonado, G.I. [University of Tennessee (Knoxville)/ORNL, Tennessee (United States)

2009-06-15

There is interest worldwide in reducing the burden on geological nuclear fuel disposal sites. In most disposal scenarios the decay heat loading of the surrounding rock limits the capacity of these sites. On the long term, this decay heat is generated primarily by actinides, and a major contributor 100 to 1000 years after discharge from the reactor is {sup 241}Am. One possible approach to reducing the decay-heat burden is to reprocess spent reactor fuel and use thermal spectrum reactors to 'burn' the Am nuclides. The viability of this approach is dependent upon the detailed changes in chemical and isotopic composition of actinide-bearing fuels after irradiation in thermal reactor spectra. The currently available thermal spectrum reactor options include light water-reactors (LWRs) and heavy-water reactors (HWRs) such as the CANDU{sup R} designs. In addition, as a result of the recycle of spent LWR fuel, there would be a considerable amount of potential recycled uranium (RU). One proposed solution for the recycled uranium is to use it as fuel in Candu reactors. This paper investigates the possibilities of transmuting americium in 'spiked' bundles in pressurized water reactors (PWRs) and in boiling water reactors (BWRs). Transmutation of Am in Candu reactors is also examined. One scenario studies a full core fuelled with homogeneous bundles of Am mixed with recycled uranium, while a second scenario places Am in an inert matrix in target channels in a Candu reactor, with the rest of the reactor fuelled with RU. A comparison of the transmutation in LWRs and HWRs is made, in terms of the fraction of Am that is transmuted and the impact on the decay heat of the spent nuclear fuel. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). (authors)

Reactor utilization, Part 1

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1981-01-01

The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

Search for other natural fission reactors

International Nuclear Information System (INIS)

Apt, K.E.; Balagna, J.P.; Bryant, E.A.; Cowan, G.A.; Daniels, W.R.; Vidale, R.J.

1977-01-01

Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m 3 and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater than 0.6 b.a. old and if they contained sufficient water but lacked neutron-absorbing impurities. While uranium deposits of northern Canada and northern Australia have received most attention, ore samples have been obtained from the following worldwide locations: the Shinkolobwe and Katanga regions of Zaire; Southwest Africa; Rio Grande do Norte, Brazil; the Jabiluka, Nabarlek, Koongarra, Ranger, and El Sharana ore bodies of the Northern Territory, Australia; the Beaverlodge, Maurice Bay, Key Lake, Cluff Lake, and Rabbit Lake ore bodies and the Great Bear Lake region, Canada. The ore samples were tested for isotopic variations in uranium, neodymium, samarium, and ruthenium which would indicate natural fission. Isotopic anomalies were not detected. Criticality was not achieved in these deposits because they did not have sufficient 235 U content (a function of age and total uranium content) and/or because they had significant impurities and insufficient moderation. A uranium mill monitoring technique has been considered where the ''yellowcake'' output from appropriate mills would be monitored for isotopic alterations indicative of the exhumation and processing of a natural reactor

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Apparatus for isotopic alteration of mercury vapor

Science.gov (United States)

Grossman, Mark W.; George, William A.; Marcucci, Rudolph V.

1988-01-01

An apparatus for enriching the isotopic Hg content of mercury is provided. The apparatus includes a reactor, a low pressure electric discharge lamp containing a fill including mercury and an inert gas. A filter is arranged concentrically around the lamp. In a preferred embodiment, constant mercury pressure is maintained in the filter by means of a water-cooled tube that depends from it, the tube having a drop of mercury disposed in it. The reactor is arranged around the filter, whereby radiation from said lamp passes through the filter and into said reactor. The lamp, the filter and the reactor are formed of a material which is transparent to ultraviolet light.

Project Plan Remote Target Fabrication Refurbishment Project

International Nuclear Information System (INIS)

Bell, Gary L.; Taylor, Robin D.

2009-01-01

In early FY2009, the DOE Office of Science - Nuclear Physics Program reinstated a program for continued production of 252 Cf and other transcurium isotopes at the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). The FY2009 major elements of the workscope are as follows: (1) Recovery and processing of seven transuranium element targets undergoing irradiation at the High Flux Isotope Reactor (HFIR) at ORNL; (2) Development of a plan to manufacture new targets for irradiation beginning in early- to mid-FY10 to supply irradiated targets for processing Campaign 75 (TRU75); and (3) Refurbishment of the target manufacturing equipment to allow new target manufacture in early FY10 The 252 Cf product from processing Campaign 74 (recently processed and currently shipping to customers) is expected to supply the domestic demands for a period of approximately two years. Therefore it is essential that new targets be introduced for irradiation by the second quarter of FY10 (HFIR cycle 427) to maintain supply of 252 Cf; the average irradiation period is ∼10 HFIR cycles, requiring about 1.5 calendar years. The strategy for continued production of 252 Cf depends upon repairing and refurbishing the existing pellet and target fabrication equipment for one additional target production campaign. This equipment dates from the mid-1960s to the late 1980s, and during the last target fabrication campaign in 2005- 2006, a number of component failures and operations difficulties were encountered. It is expected that following the target fabrication and acceptance testing of the targets that will supply material for processing Campaign 75 a comprehensive upgrade and replacement of the remote hot-cell equipment will be required prior to subsequent campaigns. Such a major refit could start in early FY 2011 and would take about 2 years to complete. Scope and cost estimates for the repairs described herein were developed, and authorization for the work

Clumped isotope effects during OH and Cl oxidation of methane

DEFF Research Database (Denmark)

Whitehill, Andrew R.; Joelsson, Lars Magnus T.; Schmidt, Johan Albrecht

2017-01-01

A series of experiments were carried out to determine the clumped (13CH3D) methane kinetic isotope effects during oxidation of methane by OH and Cl radicals, the major sink reactions for atmospheric methane. Experiments were performed in a 100 L quartz photochemical reactor, in which OH was produ......A series of experiments were carried out to determine the clumped (13CH3D) methane kinetic isotope effects during oxidation of methane by OH and Cl radicals, the major sink reactions for atmospheric methane. Experiments were performed in a 100 L quartz photochemical reactor, in which OH...... effects for singly substituted species were consistent with previous experimental studies. For doubly substituted methane, 13CH3D, the observed kinetic isotope effects closely follow the product of the kinetic isotope effects for the 13C and deuterium substituted species (i.e., 13,2KIE = 13KIE × 2KIE...... reactions. In a closed system, however, this effect is overtaken by the large D/H isotope effect, which causes the residual methane to become anti-clumped relative to the initial methane. Based on these results, we demonstrate that oxidation of methane by OH, the predominant oxidant for tropospheric methane...

Antibodies and isotopes, a chemical approach to tumour targeting

International Nuclear Information System (INIS)

Vaughan, A.T.M.; Yankuba, S.C.S.; Anderson, P.

1986-01-01

In this study, scandium-47 and yttrium-90 have been used as representatives of potential cytotoxic labels. Both isotopes have a high yield of energetic beta particles and half-lives of the same order as indium-111. In addition they are both members of Group III and so may be used as a base for chemical comparisons in the future with radiotoxic isotopes from other chemical groups

Interim waste storage for the Integral Fast Reactor

International Nuclear Information System (INIS)

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes that are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig

Performance of the multiple target He/PbI sub 2 aerosol jet system for mass separation of neutron-deficient actinide isotopes

CERN Document Server

Ichikawa, S; Asai, M; Haba, H; Sakama, M; Kojima, Y; Shibata, M; Nagame, Y; Oura, Y; Kawade, K

2002-01-01

A multiple target He/PbI sub 2 aerosol jet system coupled with a thermal ion source was installed in the isotope separator on line (JAERI-ISOL) at the JAERI tandem accelerator facility. The neutron-deficient americium and curium isotopes produced in the sup 2 sup 3 sup 3 sup , sup 2 sup 3 sup 5 U( sup 6 Li, xn) and sup 2 sup 3 sup 7 Np( sup 6 Li, xn) reactions were successfully mass-separated and the overall efficiency including the ionization of Am atoms was evaluated to be 0.3-0.4%. The identification of a new isotope sup 2 sup 3 sup 7 Cm with the present system is reported.

Cross section of α-induced reactions on iridium isotopes obtained from thick target yield measurement for the astrophysical γ process

Directory of Open Access Journals (Sweden)

T. Szücs

2018-01-01

Full Text Available The stellar reaction rates of radiative α-capture reactions on heavy isotopes are of crucial importance for the γ process network calculations. These rates are usually derived from statistical model calculations, which need to be validated, but the experimental database is very scarce. This paper presents the results of α-induced reaction cross section measurements on iridium isotopes carried out at first close to the astrophysically relevant energy region. Thick target yields of 191Ir(α,γ195Au, 191Ir(α,n194Au, 193Ir(α,n196mAu, 193Ir(α,n196Au reactions have been measured with the activation technique between Eα=13.4 MeV and 17 MeV. For the first time the thick target yield was determined with X-ray counting. This led to a previously unprecedented sensitivity. From the measured thick target yields, reaction cross sections are derived and compared with statistical model calculations. The recently suggested energy-dependent modification of the α+nucleus optical potential gives a good description of the experimental data.

Recycling, inventory and permeation of hydrogen isotopes and helium in the first wall of a thermonuclear fusion reactor

International Nuclear Information System (INIS)

Gervasini, G.; Reiter, F.

1989-01-01

The work was divided into three parts. The first part, which is theoretical, examines the behaviour of hydrogen in metals. After an introduction on the presence of hydrogen isotopes in fusion reactors, the main phenomena connected with hydrogen-metal interaction are summarised: solubility, diffusivity and trapping in material defects. The metal temperature is highlighted as the main parameter in the description of the phenomena. The second part of the work, also theoretical, concerns the interaction between helium and metals. We have tried as much as possible to show analogies and differences in the comparisons of the behaviour of hydrogen. The main types of damage caused by helium in metallic structures, which are the most important consequence of helium-metal interaction, were summarised. The characteristics of helium were treated in greater depth than those of hydrogen, because the latter are very well known. Also, there is a vast literature on the hydrogen-metal interaction. In the third and last part of the work a model was identified which allows the simulation of the evolution of a system formed from a metal in which hydrogen and helium isotopes have been introduced. A system of algebraic-differential equations was used to study the temporal evolution of the concentrations, the recycling, the inventory and the permeation of tritium and helium considering that these atoms diffuse in the metallic lattice and remain trapped in the vacancies created inside the metal by the bombardment of the neutrons from the fusion reactions. For the numerical simulation a series of data intended to represent the situation inside a thermonuclear reactor as precisely as possible were used for the numerical simulation. Analysis of the system was preceded by the analytical resolution of the steady state equations so that they could be compared with the simulation results

Numerical investigations of the fuel cycle for a 10 GW(TH)-OTTO-pebble-bed reactor with regard to high conversion ratio under special consideration of U-236 disconnexion through isotope-separation

International Nuclear Information System (INIS)

Werner, H.

1976-12-01

A conversion ratio of near 1.0 can be achieved in a pebble-bed reactor using the OTTO (once through then out) loading scheme, having an economic burn-up of the fuel, an economic power density and a moderation ratio, which is considered realistically for the future. The flexibility of the reactor concept and of the fuel element design allows to recycle the fuel during full-power operation. In the present report first the criteria are shown, which are necessary to reach a high conversion ratio. Further it is presented that the conversion ratio increases considerably by closing the fuel cycle in consequence of the building-up of U-233. In this way the fuel inventory and the fuel consumption can considerably be diminished. It is demonstrated that the building-up and the accumulation of U-236 effects an important deterioration of the neutron economy. By taking the reprocessed uranium through an isotope separation (for example: ultra-gas-centrifugation) and by separation of U-236 from the other uranium isotopes it is possible to reduce the fuel consumption considerably. The expenditure and the cost which are necessary for the isotope separation are presented. (orig.) [de

The text of the instruments connected with the Agency's assistance to Argentina in establishing a research and isotope production reactor project

International Nuclear Information System (INIS)

1995-01-01

The Agreement between the Republic of Argentina, the Federative Republic of Brazil, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials and the International Atomic Energy Agency for the Application of Safeguards came into force on 4 March 1994. As a result of the coming into force of the aforesaid Agreement for Argentina, the application of safeguards under the Project Agreement of 2 December 1964 between Argentina and the IAEA in connection with the Agency's assistance to Argentina in establishing a research and isotope production reactor project has been suspended

Replacement research reactor for Australia

International Nuclear Information System (INIS)

Miller, Ross

1998-01-01

In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

ITER TASK T26/28 (1995): Solubility, diffusion and absorption of hydrogen isotopes in potential fusion reactor ceramics

International Nuclear Information System (INIS)

Thompson, D.A.; Macauley-Newcombe, R.G.

1996-04-01

Ceramic insulators are integral parts of numerous components essential for the heating control and diagnostic measurement of fusion plasmas. For safe and reliable reactor operations it is necessary to be able to predict the resultant tritium inventories and permeation fluxes through these materials. Some materials being considered are Al 2 O 3 (both as single crystal sapphire and polycrystalline alumina) and BeO. This report contains results of ion-implantation, thermal absorption (diffusion loading) and ion-beam analysis experiments performed in 1994 and 1995 for ITER task T26/28. The combination of implantation and thermal absorption capabilities enable us to load samples with hydrogen isotopes under differing conditions. 13 figs., 1 tab., 11 refs

Development of a high-density gas-jet target for nuclear astrophysics and reaction studies with rare isotope beams. Final Report

Energy Technology Data Exchange (ETDEWEB)

Uwe, Greife [Colorado School of Mines, Golden, CO (United States)

2014-08-12

The purpose of this project was to develop a high-density gas jet target that will enable a new program of transfer reaction studies with rare isotope beams and targets of hydrogen and helium that is not currently possible and will have an important impact on our understanding of stellar explosions and of the evolution of nuclear shell structure away from stability. This is the final closeout report for the project.

Development of a high-density gas-jet target for nuclear astrophysics and reaction studies with rare isotope beams. Final Report

International Nuclear Information System (INIS)

Uwe, Greife

2014-01-01

The purpose of this project was to develop a high-density gas jet target that will enable a new program of transfer reaction studies with rare isotope beams and targets of hydrogen and helium that is not currently possible and will have an important impact on our understanding of stellar explosions and of the evolution of nuclear shell structure away from stability. This is the final closeout report for the project.

A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

International Nuclear Information System (INIS)

Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

2010-01-01

Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

Investigation of the properties of the nuclei using on the new generation reactor technology systems

International Nuclear Information System (INIS)

Tel, E.; Sahin, H. M.; Yalcin, S.; Altinok, T.; Kaplan, A.; Aydin, A.

2007-01-01

The application fields of the fast neutron are Accelerator-Driven subcritical Systems (ADS) for fission energy production and hybrid reactor systems. The technical design hybrid reactor and ADS systems potentialities require the knowledge of a wide range of better data and much effort. Thorium (Th) and Uranium (U) are nuclear fuels in these reactor systems. Lead (Pb), Bismuth (Bi) and Tungsten (W) are the target nuclei in the ADS reactor systems. The Hartree-Fock (H-F) method with an effective interaction with Skyrme forces is widely used for studying the properties of nuclei such as binding energy, Root Mean Square (RMS) charge radii, mass radii, neutron density, proton density, electromagnetic multipole moments, etc. In this study, by using H-F method with interaction Skyrme RMS charge radii, RMS mass radii, neutron density and proton density were calculated for the 2 32Th, 2 38U, 2 07Pb, 2 09Bi and 1 84W isotopes used on the new generation reactor systems. The calculation results of charge radii have been compared with experimental data and obtained other results have been discussed for hybrid and ADS reactor systems

Separation process for boron isotopes

Energy Technology Data Exchange (ETDEWEB)

Rockwood, S D

1975-06-12

The method according to the invention is characterized by the steps of preparing a gaseous mixture of BCl/sub 3/ containing the isotopes of boron and oxygen as the extractor, irradiating that mixture in the tube of the separator device by means of P- or R-lines of a CO/sub 2/ laser for exciting the molecules containing a given isotope of boron, simultaneously irradiating the mixture with UV for photodissociating the excited BCl/sub 3/ molecules and separating BCl/sub 3/ from the reaction products of photodissociation and from oxygen. Such method is suitable for preparing boron used in nuclear reactors.

Determination of Unknown Neutron Cross Sections for the Production of Medical Isotopes

Energy Technology Data Exchange (ETDEWEB)

Stephen E. Binney

2004-04-09

Calculational assessment and experimental verification of certain neutron cross sections that are related to widely needed new medical isotopes. Experiments were performed at the Oregon State University TRIGA Reactor and the High Flux Irradiation Reactor at Oak Ridge National Laboratory.

Evaluation and Compilation of Neutron Activation Cross Sections for Medical Isotope Production

International Nuclear Information System (INIS)

Binney, Stephen E.

2004-01-01

Calculational assessment and experimental verification of certain neutron cross sections that are related to widely needed new medical isotopes. Experiments were performed at the Oregon State University TRIGA Reactor and the High Flux Irradiation Reactor at Oak Ridge National Laboratory

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

International Nuclear Information System (INIS)

Rosenthal, Murray Wilford

2009-01-01

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Fuel and target programs for the transmutation at Phenix and other reactors

International Nuclear Information System (INIS)

Gaillard-Groleas, G.

2002-01-01

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

Transmutation studies in France, R and D programme on fuels and targets

International Nuclear Information System (INIS)

Boidron, M.; Chauvin, N.; Garnier, J.C.; PIllon, S.; Vambenepe, G.

2001-01-01

For the management of high level and long-lived radioactive waste, a large and continuous research and development effort is carried out in France, to provide a wide range of scientific and technical alternatives along three lines, partitioning and transmutation, disposal in deep geological formations and long term interim surface or subsurface storage. For the line one, and in close link with the partitioning studies, research is carried out to evaluate the transmutation potential of long-lived waste in appropriate reactors configurations (scenarios) relying on current technologies as well as innovative reactors. Performed to evaluate the theoretical feasibility of the Pu consumption and waste transmutation from the point of view of the reactor cores physics to reach the equilibrium of the material fluxes (i.e. consumption = production) and of the isotopic compositions of the fuels, these studies insure the 'scientific' part of the transmutation feasibility. For the technological part of the feasibility of waste transmutation in reactors, a large programme on fuel development is underway. This includes solutions based on the advanced concepts for plutonium fuels in PWR and the development of specific fuels and targets for transmutation in fast reactors in the critical or sub-critical state. For the waste transmutation in fast reactors, an important programme has been launched to develop specific fuels and targets with experiments at various stages of preparation in different experimental reactors including Phenix. Composite fuels as well as particle fuels are considered. This programme is presented and recent results concerning the preparation of the experiments, the characterisation of the compounds properties, the thermal and mechanical modelling and the behaviour of U free fuels are given. (author)

Study of medical isotope production facility stack emissions and noble gas isotopic signature using automatic gamma-spectra analysis platform

Science.gov (United States)

Zhang, Weihua; Hoffmann, Emmy; Ungar, Kurt; Dolinar, George; Miley, Harry; Mekarski, Pawel; Schrom, Brian; Hoffman, Ian; Lawrie, Ryan; Loosz, Tom

2013-04-01

ratios showed distinct differences between the closed CANDU primary coolant system and radiopharmaceutical production releases. According to the concept proposed by Kalinowski and Pistner (2006), the relationship between different isotopic activity ratios based on three or four radioxenon isotopes was plotted in a log-log diagram for source characterisation (civil vs. nuclear test). The multiple isotopic activity ratios were distributed in three distinct areas: HC atmospheric monitoring ratios extended to far left; the CANDU primary coolant system ratios lay in the middle; and 99Mo stack monitoring ratios for ANSTO and CRL were located on the right. The closed CANDU primary coolant has the lowest logarithmic mean ratio that represents the nuclear power reactor operation. The HC atmospheric monitoring exhibited a broad range of ratios spreading over several orders of magnitude. In contrast, the ANSTO and CRL stack emissions showed the smallest range of ratios but the results indicate at least two processes involved in the 99Mo productions. Overall, most measurements were found to be shifted towards the reactor domain. The hypothesis is that this is due to an accumulation of the isotope 131mXe in the stack or atmospheric background as it has the longest half-life and extra 131mXe emissions from the decay of 131I. The contribution of older 131mXe to a fresh release shifts the ratio of 133mXe/131mXe to the left. It was also very interesting to note that there were some situations where isotopic ratios from 99Mo production emissions fell within the nuclear test domain. This is due to operational variability, such as shorter target irradiation times. Martin B. Kalinowski and Christoph Pistner, (2006), Isotopic signature of atmospheric xenon released from light water reactors, Journal of Environmental Radioactivity, 88, 215-235.

The geo-reactor. A link between nuclear fission and geothermal energy?

International Nuclear Information System (INIS)

Degueldre, Claude; Fiorina, Carlo

2013-01-01

Recent high-precision isotope analysis data suggests the potential occurrence of a geo-reactor. Specific gas isotopes that could have been generated by binary and ternary fissions were identified in volcano emanations or as soluble/associated species in crystalline rocks and semi-quantitatively evaluated as isotopic ratio or estimated amounts. Presently if it is evident that according to the actinide inventory on the Earth, local potential criticality of the geo-system may have been reached, several questions remain such as why, where and when did a geo-reactor be operational? Even if the hypothesis of a geo-reactor operation in the proto-Earth period should be acceptable, it could be difficult to anticipate that a geo-reactor is still operating today. This could be tested in the future by assessing and reconstructing the system by antineutrino detection and tomography through the Earth. The present paper focuses on the geo-reactor conditions including history, spatial extension and regimes. The discussion based on recent calculations involves investigations on the limits in term of fissile inventory, size and power, based on stratification through the gravitational field and the various features through the inner mantel, the boundary with the core, the external part and the inner-core. the reconstruction allows to formulating that from the history point of view there are possibilities that the geo-reactor reached criticality in a proto-Earth period as a thorium/uranium reactor triggered by an under-layer with heavier actinides. The geo-reactor should be a key component of geothermal energy sources. (author)

Selective Gaseous Extraction: Research, Development and Training for Isotope Production, Final Technical Report

Energy Technology Data Exchange (ETDEWEB)

Bertch, Timothy C, [General Atomics

2014-03-31

General Atomics and the University of Missouri Research Reactor (MURR) completed research and development of selective gaseous extraction of fission products from irradiated fuel, which included training and education of MURR students. The process used porous fuel and after irradiation flowed product gases through the fuel to selectively removed desired fission products with the primary goal of demonstrating the removal of rhodium 105. High removal rates for the ruthenium/rhodium (Ru/Rh), tellurium/iodine (Te/I) and molybdenum/technetium (Mo/Tc) series were demonstrated. The success of this research provides for the reuse of the target for further production, significantly reducing the production of actinide wastes relative to processes that dissolve the target. This effort was conducted under DOE funding (DE-SC0007772). General Atomics objective of the project was to conduct R&D on alternative methods to produce a number of radioactive isotopes currently needed for medical and industry applications to include rhodium-105 and other useful isotopes. Selective gaseous extraction was shown to be effective at removing radioisotopes of the ruthenium/rhodium, tellurium/iodine and molybdenum/technetium decay chains while having trace to no quantities of other fission products or actinides. This adds a new, credible method to the area of certain commercial isotope production beyond current techniques, while providing significant potential reduction of process wastes. Waste reduction, along with reduced processing time/cost provides for superior economic feasibility which may allow domestic production under full cost recovery practices. This provides the potential for improved access to domestically produced isotopes for medical diagnostics and treatment at reduced cost, providing for the public good.

Radioisotope Production Plan and Strategy of Kijang Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Kye Hong; Lee, Jun Sig [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

This reactor will be located at Kijang, Busan, Korea and be dedicated to produce mainly medical radioisotopes. Tc-99m is very important isotope for diagnosis and more than 80% of radiation diagnostic procedures in nuclear medicine depend on this isotope. There were, however, several times of insecure production of Mo-99 due to the shutdown of major production reactors worldwide. OECD/NEA is leading member countries to resolve the shortage of this isotope and trying to secure the international market of Mo-99. The radioisotope plan and strategy of Kijang Research Reactor (KJRR) should be carefully established to fit not only the domestic but also international demand on Mo-99. The implementation strategy of 6 principles of HLG-MR should be established that is appropriate to national environments. Ministry of Science, ICT and Future Planning and Ministry of Health and welfare should cooperate well to organize the national radioisotope supply structure, to set up the reasonable and competitive pricing of radioisotopes, and to cope with the international supply strategy.

The 1975 DAtF-KTG reactor conference in Nuernberg

International Nuclear Information System (INIS)

Henssen, H.; Rossbach, W.

1975-01-01

A comprehensive review on the meeting is given which reports on the most important of the 204 papers read in the four session groups: 1) reactor design and experiments, 2) fuel elements, fuel cycle, and isotope technique, 3) planning, construction and operation of nuclear reactor facilities and their components, and 4) reactor types and problems of cost-efficiency. (UA/AK) [de

Isotopic composition and radiological properties of uranium in selected fuel cycles

International Nuclear Information System (INIS)

Fleischman, R.M.; Liikala, R.C.

1975-04-01

Three major topic areas are discussed: First, the properties of the uranium isotopes are defined relative to their respective roles in the nuclear fuel cycle. Secondly, the most predominant fuel cycles expected in the U. S. are described. These are the Light Water Reactor (LWR), High Temperature Gas Cooled Reactor (HTGR), and Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycles. The isotopic compositions of uranium and plutonium fuels expected for these fuel cycles are given in some detail. Finally the various waste streams from these fuel cycles are discussed in terms of their relative toxicity. Emphasis is given to the high level waste streams from reprocessing of spent fuel. Wastes from the various fuel cycles are compared based on projected growth patterns for nuclear power and its various components. (U.S.)

Proceedings of the Conference on research reactors application in Yugoslavia

International Nuclear Information System (INIS)

1978-05-01

The Conference on research reactors operation was organised on the occasion of 20 anniversary of the RB zero power reactor start-up. The presentations showed that research reactors in Yugoslavia, RB, RA and TRIGA had an important role in development of nuclear sciences and technology in Yugoslavia. The reactors were applied in non-destructive testing of materials and fuel elements, development of reactor noise techniques, safety analyses, reactor control methods, neutron activation analysis, neutron radiography, dosimetry, isotope production, etc [sr

Modelling of infrared multiphoton absorption and dissociation for design of reactors for isotope separation by lasers

International Nuclear Information System (INIS)

Takeuchi, Kazuo; Nakane, Ryohei; Inoue, Cihiro

1981-01-01

A series of experiments were performed on infrared laser beam absorption (multiphoton absorption) and subsequent dissociation (multiphoton dissociation) of CF 3 Cl to propose models for the design of reactors for isotope separation by lasers. A parallel beam geometry was utilized in batch irradiation experiments to make direct compilation of lumped-parameter data possible. Multiphoton absorption is found to be expressed by a power-law extension of the law of Lambert and by an addition of a new term for buffer gas effect to the law of Beer. For reaction analysis, a method to evaluate the effect of incomplete mixing on apparent reaction rates is first presented. Secondly, multiphoton dissociation of Cf 3 Cl is found to occur in pseudo-first order fashion and the specific reaction rates for different beam fluence are shown to be correlated to the absorbed energy. (author)

Calculations for HFIR [High Flux Isotope Reactor] fuel plate non- bonding and fuel segregation uncertainty factors

International Nuclear Information System (INIS)

Kirkpatrick, J.R.

1990-10-01

The effects of non-bonds and of fuel segregation on the package factors of the heat flux in the High Flux Isotope Reactor (HFIR) are examined. The effects of the two defects are examined both separately and together. It is concluded that the peaking factors that are used in the present HFIR thermal analysis code are conservative and thus no changes in the peaking factors are necessary to continue to ensure that HFIR is safe. A study was made of the effect of the non-bond spot diameter on the peaking factor. The conclusion is that the spot can have diameter more than three times the maximum value allowed by the specifications before the peaking factor is greater than the maximum value specified in the present HFIR thermal analysis code. 6 refs., 7 figs., 8 tabs

A Canadian isotope success story

International Nuclear Information System (INIS)

Malkoske, G.

1997-01-01

This paper provides some historical background on the commercial production of radioisotopes in Canada, and the evolution of the present vendor, MDS Nordion. The chief isotopes are molybdenum 99, iodine 131, and cobalt 60. Cobalt 60 for medical sterilization and irradiation is considered to be a significant growing market. Food irradiation is believed to be a big marketing opportunity, although attempts to popularize it have so far met with limited success. Candu reactors supply the bulk of the world's 60 Co supply. Eighty percent of the world's 99 Mo supply for medical imaging comes from Canada, and is at present produced in NRU Reactor, which is to be replaced by two Maple reactors coming into production in 1999 and 2000

Technology for down-blending weapons grade uranium into commercial reactor-usable uranium

International Nuclear Information System (INIS)

Arbital, J.G.; Snider, J.D.

1996-01-01

The US Department of Energy (DOE) is evaluating options for rendering surplus inventories of highly enriched uranium (HEU) incapable of being used in nuclear weapons. Weapons-capable HEU was earlier produced by enriching the uranium isotope 235 U from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by permanently diluting the concentration of the 235 U isotope, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope re-enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended, low-enriched uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel. The DOE has evaluated three candidate processes for down blending surplus HEU. These candidate processes are: (1) uranium hexafluoride blending; (2) molten uranium metal blending; and (3) uranyl nitrate solution blending. This paper describes each of these candidate processes. It also compares the relative advantages and disadvantages of each process with respect to: (1) the various forms and compounds of HEU comprising the surplus inventory, (2) the use of down-blended product as commercial reactor fuel, or (3) its disposal as waste

Breeding and plutonium characterization analysis on actinides closed water-cooled thorium reactor

International Nuclear Information System (INIS)

Permana, Sidik; Sekimoto, Hiroshi; Takaki, Naoyuki

2009-01-01

Higher difficulties (barrier) or more complex design of nuclear weapon, material fabrication and handling and isotopic enrichment can be achieved by a higher isotopic barrier. The isotopic material barrier includes critical mass, heat-generation rate, spontaneous neutron generation and radiation. Those isotopic barriers in case of plutonium isotope is strongly depend on the even mass number of plutonium isotope such as 238 Pu, 240 Pu and 242 Pu and for 233 U of thorium cycle depends on 232 U. In this present study, fuel sustainability as fuel breeding capability and plutonium characterization as main focus of proliferation resistance analysis have been analyzed. Minor actinide (MA) is used as doping material to be loaded into the reactors with thorium fuel. Basic design parameters are based on actinide closed-cycle reactor cooled by heavy water. The evaluation use equilibrium burnup analysis coupled with cell calculation of SRAC and nuclear data library is JENDL.32. Parametrical survey has been done to analyze the effect of MA doping rate, different moderation ratio for several equilibrium burnup cases. Plutonium characterization which based on plutonium isotope composition is strongly depending on MA doping concentration and different moderation conditions. Breeding condition can be achieved and high proliferation resistance level can be obtained by the present reactor systems. Higher isotopic plutonium composition of Pu-238 (more than 40%) can be obtained compared with other plutonium isotopes. In addition, higher moderation ratio gives the isotope composition of 238 Pu increases, however, it obtains lower composition when MA doping is increased and it slightly lower composition for higher burnup. Meanwhile, higher 240 Pu composition can be achieved by higher MA doping rate as well as for obtaining higher breeding capability. (author)

Practical course on reactor instrumentation

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-06-01

This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

Cross sections for the production of Li and Be isotopes in carbon targets irradiated by 300 GeV protons

International Nuclear Information System (INIS)

Raisbeck, G.M.; Lestringuez, J.; Yiou, F.

1975-01-01

Cross sections for the production of Li and Be isotopes in carbon targets irradiated by 300 GeV protons were measured by mass spectrometry. The results are compared with lower energy measurements and discussed in terms of the variation of the cosmic ray L/M ratio in this energy region [fr

Overview of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Nguyen Thai Sinh; Luong Ba Vien

2016-01-01

The present reactor called Dalat Nuclear Research Reactor (DNRR) has been reconstructed from the former TRIGA Mark II reactor which was designed by General Atomic (GA, San Diego, California, USA), started building in early 1960s, put into operation in 1963 and operated until 1968 at nominal power of 250 kW. In 1975, all fuel elements of the reactor were unloaded and shipped back to the USA. The DNRR is a 500-kW pool-type research reactor using light water as both moderator and coolant. The reactor is used as a neutron source for the purposes of: (1) radioactive isotope production; (2) neutron activation analysis; and (3) research and training

Nuclear material safeguards surveillance and accountancy by isotope correlation techniques

International Nuclear Information System (INIS)

Persiani, P.J.; Goleb, J.A.; Kroc, T.K.

1981-11-01

The purpose of this study is to investigate the applicability of isotope correlation techniques (ICT) to the Light Water Reactor (LWR) and the Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycles for nuclear material accountancy and safeguards surveillance. The isotopic measurement of the inventory input to the reprocessing phase of the fuel cycle is the primary direct determination that an anomaly may exist in the fuel management of nuclear material. The nuclear materials accountancy gap which exists between the fabrication plant output and the input to the reprocessing plant can be minimized by using ICT at the dissolver stage of the reprocessing plant. The ICT allows a level of verification of the fabricator's fuel content specifications, the irradiation history, the fuel and blanket assemblies management and scheduling within the reactor, and the subsequent spent fuel assembly flows to the reprocessing plant. The investigation indicates that there exist relationships between isotopic concentration which have predictable, functional behavior over a range of burnup. Several cross-correlations serve to establish the initial core assembly-averaged composition. The selection of the more effective functionals will depend not only on the level of reliability of ICT for verification, but also on the capability, accuracy and difficulty of developing measurement methods. The propagation of measurement errors on the correlation functions and respective sensitivities to isotopic compositional changes have been examined and found to be consistent with current measurement methods

Isotope exchange reactions in hydrogen mixtures

International Nuclear Information System (INIS)

Czaplinski, W.; Gula, A.; Kravtsov, A.; Mikhailov, A.; Popov, N.

1990-12-01

The rates of isotopic exchange for the excited states of muonic hydrogen are calculated as functions of collision energy. Ground state population q 1s for different collision energies, target densities and isotope concentrations is obtained. It is shown that for principal quantum numbers n > 5 the isotopic exchange still considerably influences the value of q 1s . (author)

Design and use of the ORNL HFIR [High Flux Isotope Reactor] pneumatic tube irradiation systems

International Nuclear Information System (INIS)

Dyer, F.F.; Emery, J.F.; Robinson, L.; Teasley, N.A.

1987-01-01

A second pneumatic tube that was recently installed in the High Flux Isotope Reactor for neutron activation analysis is described. Although not yet tested, the system is expected to have a thermal neutron flux of about 1.5 x 10 14 cm -2 s -1 . A delayed neutron counter is an integral part of the pneumatic tube, and all of the hardware is present to enable automated use of the counter. The system is operated with a Gould programmable controller that is programmed with an IBM personal computer. Automation of any mode of operation, including the delayed neutron counter, will only require a nominal amount of software development. Except for the lack of a hot cell, the irradiation facility has all of the advantageous features of an older pneumatic tube that has been in operation for 17 years. The design of the system and some applications and methods of operation are described

Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

International Nuclear Information System (INIS)

Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

1999-01-01

This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept

Cross sections for the production of Li and Be isotopes in carbon targets irradiated by 300 GeV protons

International Nuclear Information System (INIS)

Raisbeck, G.M.; Lestringuez, J.; Yiou, F.

1975-01-01

Cross sections for the production of Li and Be isotopes in carbon targets irradiated by 300 GeV protons have been measured by mass spectrometry. The results are compared with lower energy measurements and discussed in terms of the variation of the cosmic ray L/M ratio in the energy region [fr

Contained fissionly vaporized imploded fission explosive breeder reactor

International Nuclear Information System (INIS)

Marwick, E.F.

1978-01-01

Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

The first university research reactor in India

International Nuclear Information System (INIS)

Murthy, G.S.

1999-01-01

At low power research reactor is being set up in Andhra University to cater to the needs of researchers and isotope users by the Department of Atomic Energy in collaboration with Andhra University. This reactor is expected to be commissioned by 2001-02. Departments like Chemistry, Earth Sciences, Physics, Life Sciences, Pharmacy, Medicine and Engineering would be the beneficiaries of the availability of this reactor. In this paper, details of the envisaged research programme and training activities are discussed. (author)

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

Verpoortem, J. R.; Van Dyck, S.

2014-01-01

At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

The resonance absorption controlled reactor

Energy Technology Data Exchange (ETDEWEB)

Caro, R

1977-07-01

In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D{sub 2}O/H{sub 2}O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs.

The resonance absorption controlled reactor

International Nuclear Information System (INIS)

Caro, R.

1977-01-01

In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D 2 O/H 2 O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs

Hydrogen isotope double differential production cross sections induced by 62.7 MeV neutrons on a lead target

International Nuclear Information System (INIS)

Kerveno, M.; Haddad, F.; Eudes, Ph.; Kirchner, T.; Lebrun, C.; Slypen, I.; Meulders, J.P.; Le Brun, C.; Lecolley, F.R.; Lecolley, J.F.; Louvel, M.; Lefebvres, F.; Hilaire, S.; Koning, A.J.

2002-01-01

Double differential hydrogen isotope production cross sections have been extracted in 62.7 MeV neutron induced reactions on a lead target. The angular distribution was measured at eight angles from 20 deg. to 160 deg. allowing the extraction of angle-differential, energy differential, and total production cross sections. A first set of comparisons with several theoretical calculations is also presented

Particle and radiation simulations for the proposed rare isotope accelerator facility

Energy Technology Data Exchange (ETDEWEB)

Remec, Igor [Oak Ridge National Laboratory, Oak Ridge, P. O. Box 2008, TN 37831-6172 (United States)]. E-mail: remeci@ornl.gov; Gabriel, Tony A. [Oak Ridge National Laboratory, Oak Ridge, P. O. Box 2008, TN 37831-6172 (United States); Wendel, Mark W. [Oak Ridge National Laboratory, Oak Ridge, P. O. Box 2008, TN 37831-6172 (United States); Conner, David L. [Oak Ridge National Laboratory, Oak Ridge, P. O. Box 2008, TN 37831-6172 (United States); Burgess, Thomas W. [Oak Ridge National Laboratory, Oak Ridge, P. O. Box 2008, TN 37831-6172 (United States); Ronningen, Reginald M. [National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824 (United States); Blideanu, Valentin [National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824 (United States); Bollen, Georg [National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824 (United States); Boles, Jason L. [Lawrence Livermore National Laboratory, P. O. Box 808, L-446, Livermore, CA 94550 (United States); Reyes, Susana [Lawrence Livermore National Laboratory, P. O. Box 808, L-446, Livermore, CA 94550 (United States); Ahle, Larry E. [Lawrence Livermore National Laboratory, P. O. Box 808, L-446, Livermore, CA 94550 (United States); Stein, Werner [Lawrence Livermore National Laboratory, P. O. Box 808, L-446, Livermore, CA 94550 (United States)

2006-06-23

The Rare Isotope Accelerator (RIA) facility, planned to be built in the USA, will be capable of delivering diverse beams, from protons to uranium ions, with energies from 1 GeV to at least 400 MeV per nucleon to rare isotope-producing targets. High beam power-400 kW-will allow RIA to become the most powerful rare isotope beam facility in the world; however, it also creates challenges for the design of the isotope-production targets. This paper focuses on the isotope-separator-on-line (ISOL) target work, particularly the radiation transport aspects of the two-step fission target design. Simulations were performed with the PHITS, MCNPX, and MARS15 computer codes. A two-step ISOL target considered here consists of a mercury or tungsten primary target in which primary beam interactions release neutrons, which in turn induce fissions-and produce rare isotopes-in the secondary target filled with fissionable material. Three primary beams were considered: 1-GeV protons, 622-MeV/u deuterons, and 777-MeV/u {sup 3}He ions. The proton and deuterium beams were found to be about equivalent in terms of induced fission rates and heating rates in the target, while the {sup 3}He beam, without optimizing the target geometry, was less favorable, producing about 15% fewer fissions and about 50% higher heating rates than the proton beam at the same beam power.

Second international conference on isotopes. Conference proceedings

International Nuclear Information System (INIS)

Hardy, C.J.

1997-10-01

The Second International Conference on Isotopes (2ICI) was hosted by the Australian Nuclear Association in Sydney, NSW, Australia. The Theme of the Second Conference: Isotopes for Industry, Health and a Better Environment recognizes that isotopes have been used in these fields successfully for many years and offer prospects for increasing use in the future. The worldwide interest in the use of research reactors and accelerators and in applications of stable and radioactive isotopes, isotopic techniques and radiation in industry, agriculture, medicine, environmental studies and research in general, was considered. Other radiation issues including radiation protection and safety were also addressed. International and national overviews and subject reviews invited from leading experts were included to introduce the program of technical sessions. The invited papers were supported by contributions accepted from participants for oral and poster presentation. A Technical Exhibition was held in association with the Conference. This volume contains the foreword, technical program, the author index and of the papers (1-60) presented at the conference

Second international conference on isotopes. Conference proceedings

Energy Technology Data Exchange (ETDEWEB)

Hardy, C J [ed.

1997-10-01

The Second International Conference on Isotopes (2ICI) was hosted by the Australian Nuclear Association in Sydney, NSW, Australia. The Theme of the Second Conference: Isotopes for Industry, Health and a Better Environment recognizes that isotopes have been used in these fields successfully for many years and offer prospects for increasing use in the future. The worldwide interest in the use of research reactors and accelerators and in applications of stable and radioactive isotopes, isotopic techniques and radiation in industry, agriculture, medicine, environmental studies and research in general, was considered. Other radiation issues including radiation protection and safety were also addressed. International and national overviews and subject reviews invited from leading experts were included to introduce the program of technical sessions. The invited papers were supported by contributions accepted from participants for oral and poster presentation. A Technical Exhibition was held in association with the Conference. This volume contains the full text or extended abstracts of papers number 61- to number 114

Second international conference on isotopes. Conference proceedings

Energy Technology Data Exchange (ETDEWEB)

Hardy, C J [ed.

1997-10-01

The Second International Conference on Isotopes (2ICI) was hosted by the Australian Nuclear Association in Sydney, NSW, Australia. The Theme of the Second Conference: Isotopes for Industry, Health and a Better Environment recognizes that isotopes have been used in these fields successfully for many years and offer prospects for increasing use in the future. The worldwide interest in the use of research reactors and accelerators and in applications of stable and radioactive isotopes, isotopic techniques and radiation in industry, agriculture, medicine, environmental studies and research in general, was considered. Other radiation issues including radiation protection and safety were also addressed. International and national overviews and subject reviews invited from leading experts were included to introduce the program of technical sessions. The invited papers were supported by contributions accepted from participants for oral and poster presentation. A Technical Exhibition was held in association with the Conference. This volume contains the foreword, technical program, the author index and of the papers (1-60) presented at the conference.

Multiple-isotope separation in gas centrifuge

International Nuclear Information System (INIS)

Wood, Houston G.; Mason, T.C.; Soubbaramayer

1996-01-01

In previous works, the Onsager's pancake equation was used to provide solution to the internal counter-current flow field, which is necessary to calculate solutions to the isotope transport equation. The diffusion coefficient was assumed to be constant which is a good approximation for gases with large molecular weights, and small differences in the molecular weights of the various isotopes. A new optimization strategy was presented for determining the operating parameters of a gas centrifuge to be used for multiple-component isotope separation. Scoop drag, linear wall temperature gradient, the feed rate ant the cut have been chosen as operating parameters for the optimization. The investigation was restricted to a single centrifuge, and the problem of cascading for multiple-isotope separation was not addressed. The model describing the flow and separation phenomena in centrifuge includes a set of equations describing the fluid dynamics of the counter-current flow and the diffusion equations written for each isotope of the mixture. In this paper, an optimization algorithm is described and applied to an example for the re enrichment of spent reactor uranium

High spin K isomeric target of 177mLu

International Nuclear Information System (INIS)

Roig, O.; Belier, G.; Daugas, J.-M.; Delbourgo, P.; Maunoury, L.; Meot, V.; Morichon, E.; Sauvestre, J.-E.; Aupiais, J.; Boulin, Y.; Fioni, G.; Letourneau, A.; Marie, F.; Ridikas, D.

2004-01-01

The techniques used to produce a 177m Lu (J π =23/2 - ,T 1/2 =160.4 days) target are described in this paper. Firstly, an isotopic separation of an enriched lutetium sample was used to reach a purity of 176 Lu close to 99.993%. Afterwards, the high neutron flux of the Grenoble Institut Laue-Langevin reactor was used to produce the 177m Lu isomer by the 176 Lu(n,γ) reaction. Finally, a chemical separation was performed to extract 10 13 nuclei of 177m Lu. Thanks to this experiment, we have been able to estimate the destruction cross-section of the 177m Lu

Post-Irradiation Examination of Array Targets - Part I

Energy Technology Data Exchange (ETDEWEB)

Icenhour, A.S.

2004-01-23

During FY 2001, two arrays, each containing seven neptunium-loaded targets, were irradiated at the Advanced Test Reactor in Idaho to examine the influence of multi-target self-shielding on {sup 236}Pu content and to evaluate fission product release data. One array consisted of seven targets that contained 10 vol% NpO{sub 2} pellets, while the other array consisted of seven targets that contained 20 vol % NpO{sub 2} pellets. The arrays were located in the same irradiation facility but were axially separated to minimize the influence of one array on the other. Each target also contained a dosimeter package, which consisted of a small NpO{sub 2} wire that was inside a vanadium container. After completion of irradiation and shipment back to the Oak Ridge National Laboratory, nine of the targets (four from the 10 vol% array and five from the 20 vol% array) were punctured for pressure measurement and measurement of {sup 85}Kr. These nine targets and the associated dosimeters were then chemically processed to measure the residual neptunium, total plutonium production, {sup 238}Pu production, and {sup 236}Pu concentration at discharge. The amount and isotopic composition of fission products were also measured. This report provides the results of the processing and analysis of the nine targets.

Solid State Reactor Final Report

Energy Technology Data Exchange (ETDEWEB)

Mays, G.T.

2004-03-10

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

Efforts to save 244Pu in Mark 18A targets for use in international safeguards measurements

International Nuclear Information System (INIS)

Goldberg, Steven A.; Cappis, John; Clarke, Stephanie; Whitesel, Robert

2001-01-01

Full text: The Office of Arms Control and Nonproliferation and the Office of Security and Emergency Operations are working collaboratively to evaluate the disposition of a large quantity of the 244 Pu isotope contained in 65 Mark ISA targets at the Savannah River Site (SRS). 244 Pu is used as a standard reference material for plutonium analytical measurements required for both domestic and international safeguards. 244 Pu is particularly valuable for high accuracy measurements of plutonium in small samples containing trace quantities of plutonium (environmental analysis) and for measurements of material through-put in bulk processing facilities handling large volumes of plutonium and plutonium-bearing materials. In October 2000, an assessment team was tasked by the U.S. Department of Energy (DOE) Under Secretary to evaluate pathways and costs for the chemical separation and isotopic enrichment of the 244 Pu identified in the targets. Even though the target materials have recently been designated as a National Resource, they are scheduled for waste disposal unless funds can be identified and assigned to the project. Background information on the Mark ISA targets and a review of the assessment process are presented below to inform other organizations and governments of current efforts to examine potential disposition options and to solicit international cooperation for the extraction of the 244 Pu. Background - The United States possesses the bulk of the world's supply of the rare isotope 244 Pu. This isotope was produced by extremely long neutron irradiation of 242 Pu in a high-flux reactor during experiments used primarily to create isotopes of medical interest. In its separated enriched form, 244 Pu is regarded as the most accurate and desirable spike for safeguards, forensics, and environmental analysis of plutonium, allowing the simultaneous measurement of a sample for isotopic abundances and elemental concentration. Such measurements are a critical component of

Development of a general learning algorithm with applications in nuclear reactor systems

Energy Technology Data Exchange (ETDEWEB)

Brittain, C.R.; Otaduy, P.J.; Perez, R.B.

1989-12-01

The objective of this study was development of a generalized learning algorithm that can learn to predict a particular feature of a process by observation of a set of representative input examples. The algorithm uses pattern matching and statistical analysis techniques to find a functional relationship between descriptive attributes of the input examples and the feature to be predicted. The algorithm was tested by applying it to a set of examples consisting of performance descriptions for 277 fuel cycles of Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR). The program learned to predict the critical rod position for the HFIR from core configuration data prior to reactor startup. The functional relationship bases its predictions on initial core reactivity, the number of certain targets placed in the center of the reactor, and the total exposure of the control plates. Twelve characteristic fuel cycle clusters were identified. Nine fuel cycles were diagnosed as having noisy data, and one could not be predicted by the functional relationship. 13 refs., 6 figs.

Development of a general learning algorithm with applications in nuclear reactor systems

International Nuclear Information System (INIS)

Brittain, C.R.; Otaduy, P.J.; Perez, R.B.

1989-12-01

The objective of this study was development of a generalized learning algorithm that can learn to predict a particular feature of a process by observation of a set of representative input examples. The algorithm uses pattern matching and statistical analysis techniques to find a functional relationship between descriptive attributes of the input examples and the feature to be predicted. The algorithm was tested by applying it to a set of examples consisting of performance descriptions for 277 fuel cycles of Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR). The program learned to predict the critical rod position for the HFIR from core configuration data prior to reactor startup. The functional relationship bases its predictions on initial core reactivity, the number of certain targets placed in the center of the reactor, and the total exposure of the control plates. Twelve characteristic fuel cycle clusters were identified. Nine fuel cycles were diagnosed as having noisy data, and one could not be predicted by the functional relationship. 13 refs., 6 figs

Stable isotope enrichment: Current and future potential

International Nuclear Information System (INIS)

Tracy, J.G.; Aaron, W.S.

1992-01-01

Oak Ridge National Laboratory (ORNL) operates the Isotope Enrichment Facility for the purpose of providing enriched stable isotopes, selected radioactive isotopes (including the actinides), and isotope-related materials and services for use in various research applications. ORNL is responsible for isotope enrichment and the distribution of approximately 225 nongaseous stable isotopes from 50 multi-isotopic elements. Many enriched isotope products are of prime importance in the fabrication of nuclear targets and the subsequent production of special radionuclides. State-of-the-art techniques to achieve special isotopic, chemical, and physical requirements are performed at ORNL This report describes the status and capabilities of the Isotope Enrichment Facility and the Isotope Research Materials Laboratory as well as emphasizing potential advancements in enrichment capabilities

Stable isotope enrichment - current and future potential

International Nuclear Information System (INIS)

Tracy, J.G.; Aaron, W.S.

1993-01-01

Oak Ridge National Laboratory (ORNL) operates the Isotope Enrichment Facility for the purpose of providing enriched stable isotopes, selected radioactive isotopes (including the actinides), and isotope-related materials and services for use in various research applications. ORNL is responsible for isotope enrichment and the distribution of approximately 225 nongaseous stable isotopes from 50 multi-isotopic elements. Many enriched isotope products are of prime importance in the fabrication of nuclear targets and the subsequent production of special radionuclides. State-of-the-art techniques to achieve special isotopic, chemical, and physical requirements are performed at ORNL. This report describes the status and capabilities of the Isotope Enrichment Facility and the Isotope Research Materials Laboratory as well as emphasizing potential advancements in enrichment capabilities. (orig.)

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-09-01

The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

Optimization of on-line hydrogen stable isotope ratio measurements of halogen- and sulfur-bearing organic compounds using elemental analyzer-chromium/high-temperature conversion isotope ratio mass spectrometry (EA-Cr/HTC-IRMS).

Science.gov (United States)

Gehre, Matthias; Renpenning, Julian; Geilmann, Heike; Qi, Haiping; Coplen, Tyler B; Kümmel, Steffen; Ivdra, Natalija; Brand, Willi A; Schimmelmann, Arndt

2017-03-30

Accurate hydrogen isotopic analysis of halogen- and sulfur-bearing organics has not been possible with traditional high-temperature conversion (HTC) because the formation of hydrogen-bearing reaction products other than molecular hydrogen (H 2 ) is responsible for non-quantitative H 2 yields and possible hydrogen isotopic fractionation. Our previously introduced, new chromium-based EA-Cr/HTC-IRMS (Elemental Analyzer-Chromium/High-Temperature Conversion Isotope Ratio Mass Spectrometry) technique focused primarily on nitrogen-bearing compounds. Several technical and analytical issues concerning halogen- and sulfur-bearing samples, however, remained unresolved and required further refinement of the reactor systems. The EA-Cr/HTC reactor was substantially modified for the conversion of halogen- and sulfur-bearing samples. The performance of the novel conversion setup for solid and liquid samples was monitored and optimized using a simultaneously operating dual-detection system of IRMS and ion trap MS. The method with several variants in the reactor, including the addition of manganese metal chips, was evaluated in three laboratories using EA-Cr/HTC-IRMS (on-line method) and compared with traditional uranium-reduction-based conversion combined with manual dual-inlet IRMS analysis (off-line method) in one laboratory. The modified EA-Cr/HTC reactor setup showed an overall H 2 -recovery of more than 96% for all halogen- and sulfur-bearing organic compounds. All results were successfully normalized via two-point calibration with VSMOW-SLAP reference waters. Precise and accurate hydrogen isotopic analysis was achieved for a variety of organics containing F-, Cl-, Br-, I-, and S-bearing heteroelements. The robust nature of the on-line EA-Cr/HTC technique was demonstrated by a series of 196 consecutive measurements with a single reactor filling. The optimized EA-Cr/HTC reactor design can be implemented in existing analytical equipment using commercially available material and

Development of empirical relation for isotope of uranium in enriched uranium matrix

International Nuclear Information System (INIS)

Srivastava, S.K.; Vidyasagar, D.; Jha, S.K.; Tripathi, R.M.

2018-01-01

Uranium enriched in 235 U is required in commercial light water reactors to produce a controlled nuclear reaction. Enrichment allows the 235 U isotopes to be increased from 0.71% to a range between 2% to 5% depending upon requirement. The enriched uranium in the form of sintered UO 2 pellet is used for any commercially operating boiling light water reactors. The enriched uranium fuel bundle surface swipes sample is being analysed to assess the tramp uranium as a quality control parameter. It is known that the 234 U isotope also enriched along with 235 U isotope in conventional gaseous diffusion enrichment process. The information about enrichment percentage of 234 U helps to characterize isotopic properties of enriched uranium. A few reports provide the empirical equation and graphs for finding out the specific activity, activity percentage, activity ratio of 234 U isotopes for enriched uranium. Most of them have not provided the reference for the data used and their source. An attempt has been made to model the relationship between 234 U and 235 U as a function of uranium enrichment at low level

Oklo natural reactor

International Nuclear Information System (INIS)

Fujii, Isao

1985-01-01

In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

Radioactive fallout from the Chernobyl nuclear reactor accident

International Nuclear Information System (INIS)

Beiriger, J.M.; Failor, R.A.; Marsh, K.V.; Shaw, G.E.

1987-08-01

This report describes the detection of fallout in the United States from the Chernobyl nuclear reactor accident. As part of its environmental surveillance program, Lawrence Livermore National Laboratory maintained detectors for gamma-emitting radionuclides. Following the reactor accident, additional air filters were set out. Several uncommon isotopes were detected at the time the plume passed into the US

Homogeneous aqueous solution nuclear reactors for the production of Mo-99 and other short lived radioisotopes

International Nuclear Information System (INIS)

2008-09-01

Technetium-99m ( 99m Tc), the daughter of Molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. It accounts for over twenty-five million medical procedures each year worldwide, comprising about 80% of all radiopharmaceutical procedures. 99 Mo is mostly prepared by the fission of uranium-235 targets in a nuclear reactor with a fission yield of about 6.1%. Currently over 95% of the fission product 99 Mo is obtained using highly enriched uranium (HEU) targets. Smaller scale producers use low enriched uranium (LEU) targets. Small quantities of 99 Mo are also produced by neutron activation through the use of the (n, γ) reaction. The concept of a compact homogeneous aqueous reactor fuelled by a uranium salt solution with off-line separation of radioisotopes of interest ( 99 Mo, 131 I) from aliquots of irradiated fuel solution has been cited in a few presentations in the series of International Conference on Isotopes (ICI) held in Vancouver (2000), Cape Town (2003) and Brussels (2005) and recently some corporate interest has also been noticeable. Calculations and some experimental research have shown that the use of aqueous homogeneous reactors (AHRs) could be an efficient technology for fission radioisotope production, having some prospective advantages compared with traditional technology based on the use of solid uranium targets irradiated in research reactors. This review of AHR status and prospects by a team of experts engaged in the field of homogeneous reactors and radioisotope producers yields an objective evaluation of the technological challenges and other relevant implications. The meeting to develop this report facilitated the exchange of information on the 'state of the art' of the technology related to homogeneous aqueous solution nuclear reactors, especially in connection with the production of radioisotopes. This publication presents a summary of discussions of a consultants meeting which is followed by the technical

Oklo reactors: natural analogs to nuclear waste repositories

International Nuclear Information System (INIS)

Curtis, D.B.; Benjamin, T.M.; Gancarz, A.J.

1981-01-01

The 2-billion-year-old fossil reactors at Oklo are ancient natural nuclear waste sites. Isotope dilution mass spectrometric analyses of the fission products in the reactor core uraninite and the peripheral pelitic sandstone provide data for calculating the reactor operating parameters, the quantities of fissiogenic isotopes produced, the fraction of these isotopes retained in the cores, and the location in the peripheral rocks of the fissiogenic fraction lost from the cores. For a duration of criticality of 3 x 10 5 yrs, the thermal plus resonance neutron fluence ranged between 10 20 and 10 21 neutrons/cm 2 . The fraction of technetium (60 to 85%), ruthenium (75 to 90%), and neodymium (85 to 100%) retained is negatively correlated with fluence. The lost fission products are contained within a few tens of meters of their source, the reactor cores. The systematics of the decay of 99 Tc (t/sub 1/2/ = 2.13 x 10 5 yr) to 99 Ru limits the period of fissiogenic element migration to approximately 1 million yr at a time 2 billion yr ago. Thermodynamic calculations of the temperature-dependent solubilities indicate that the loss of fissiogenic elements is diffusion controlled, whereas retention in the surrounding rocks is a result of temperature-dependent deposition from an aqueous solution. These results concerning the geochemistry of technetium, ruthenium, and neodymium at a natural waste site support the concept of geologic burial of man-made radioactive wastes

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.

Inertial fusion reactor designs

International Nuclear Information System (INIS)

Meier, W.

1987-01-01

In this paper, a variety of reactor concepts are proposed. One of the prime concerns is dealing with the x-rays and debris that are emitted by the target. Internal neutron shielding can reduce radiation damage and activation, leading to longer life systems, reduced activation and fewer safety concerns. There is really no consensus on what the best reactor concept is at this point. There has been virtually no chamber technology development to date. This is the flip side of the coin of the separability of the target physics and the reactor design. Since reactor technology has not been required to do target experiments, it's not being developed. Economic analysis of conceptual designs indicates that ICF can be economically competitive with magnetic fusion, fission and fossil plants

Benchmarking burnup reconstruction methods for dynamically operated research reactors

Energy Technology Data Exchange (ETDEWEB)

Sternat, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Charlton, William S. [Univ. of Nebraska, Lincoln, NE (United States). National Strategic Research Institute; Nichols, Theodore F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-01

The burnup of an HEU fueled dynamically operated research reactor, the Oak Ridge Research Reactor, was experimentally reconstructed using two different analytic methodologies and a suite of signature isotopes to evaluate techniques for estimating burnup for research reactor fuel. The methods studied include using individual signature isotopes and the complete mass spectrometry spectrum to recover the sample’s burnup. The individual, or sets of, isotopes include ¹⁴⁸Nd, ¹³⁷Cs+¹³⁷Ba, ¹³⁹La, and ¹⁴⁵Nd+¹⁴⁶Nd. The storage documentation from the analyzed fuel material provided two different measures of burnup: burnup percentage and the total power generated from the assembly in MWd. When normalized to conventional units, these two references differed by 7.8% (395.42GWd/MTHM and 426.27GWd/MTHM) in the resulting burnup for the spent fuel element used in the benchmark. Among all methods being evaluated, the results were within 11.3% of either reference burnup. The results were mixed in closeness to both reference burnups; however, consistent results were achieved from all three experimental samples.

Separation and preparation of "6"2Ni isotope

International Nuclear Information System (INIS)

Ren Xiuyan; Mi Yajing; Zeng Ziqiang; Li Gongliang; Tu Rui

2014-01-01

Micro nuclear battery is the perfect power of space craft equipment. "6"3Ni is the core operation material of the "6"3Ni battery. It can produce radioisotope "6"3Ni while high abundance "6"2Ni is irradiated in the reactor. In order to meet the requirements of the abundance and the purity, research of the separation for "6"2Ni isotope was developed. The magnetic field and beam transmission status were simulated. The improvement designs of the ion source and the collector pocket were carried out. The process flow of high abundance "6"2Ni using electromagnetic separation method was established. The experiment of "6"2Ni isotope was developed by using electromagnetism isotope separator. The results show that the enrichment of "6"2Ni isotope is more than 90%. (authors)

New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Ma, X.B., E-mail: maxb@ncepu.edu.cn; Qiu, R.M.; Chen, Y.X.

2017-02-15

Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between {sup 235}U and {sup 239}Pu, the covariance coefficient changes from 0.15 to −0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller. - Highlights: • The covariance coefficients between isotopes vs reactor burnup may change its sign because of two opposite effects. • The relation between fission fraction uncertainty and atomic density are first studied. • A new MC-based method of evaluating the covariance coefficients between isotopes was proposed.

Cost targets for at-reactor spent fuel rod consolidation

International Nuclear Information System (INIS)

Macnabb, W.V.

1985-01-01

The high-level nuclear waste management system in the US currently envisions the disposal of spent fuel rods that have been removed from their assemblies and reconfigured into closely packed arrays. The process of fuel rod removal and packaging, referred to as rod consolidation, can occur either at reactors or at an integrated packaging facility, monitored retrievable storage (MRS). Rod consolidation at reactors results in cost savings down stream of reactors by reducing needs for additional storage, reducing the number of shipments, and reducing (eliminating, in the extreme) the amount of fuel handling and consolidation at the MRS. These savings accrue to the nuclear waste fund. Although private industry is expected to pay for at-reactor activities, including rod consolidation, it is of interest to estimate cost savings to the waste system if all fuel were consolidated at reactors. If there are savings, the US Department of Energy (DOE) may find it advantageous to pay for at-reactor rod consolidation from the nuclear waste fund. This paper assesses and compares the costs of rod consolidation at reactors and at the MRS in order to determine at what levels the former could be cost competitive with the latter

Safety re-assessment of AECL test and research reactors

International Nuclear Information System (INIS)

Winfield, D.J.

1990-01-01

Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

Selection of support structure materials for irradiation experiments in the HFIR [High Flux Isotope Reactor] at temperatures up to 500 degrees C

International Nuclear Information System (INIS)

Farrell, K.; Longest, A.W.

1990-01-01

The key factor in the design of capsules for irradiation of test specimens in the High Flux Isotope Reactor at preselected temperatures up to 500 degree C utilizing nuclear heating is a narrow gas-filled gap which surrounds the specimens and controls the transfer of heat from the specimens through the wall of a containment tube to the reactor cooling water. Maintenance of this gap to close tolerances is dependent on the characteristics of the materials used to support the specimens and isolate them from the water. These support structure materials must have low nuclear heating rates, high thermal conductivities, and good dimensional stabilities under irradiation. These conditions are satisfied by certain aluminum alloys. One of these alloys, a powder metallurgy product containing a fine dispersion of aluminum oxide, is no longer manufactured. A new alloys of this type, with the trade name DISPAL, is determined to be a suitable substitute. 23 refs., 13 figs., 3 tabs

Particle and radiation simulations for the proposed rare isotope accelerator facility

Science.gov (United States)

Remec, Igor; Gabriel, Tony A.; Wendel, Mark W.; Conner, David L.; Burgess, Thomas W.; Ronningen, Reginald M.; Blideanu, Valentin; Bollen, Georg; Boles, Jason L.; Reyes, Susana; Ahle, Larry E.; Stein, Werner

2006-06-01

The Rare Isotope Accelerator (RIA) facility, planned to be built in the USA, will be capable of delivering diverse beams, from protons to uranium ions, with energies from 1 GeV to at least 400 MeV per nucleon to rare isotope-producing targets. High beam power—400 kW—will allow RIA to become the most powerful rare isotope beam facility in the world; however, it also creates challenges for the design of the isotope-production targets. This paper focuses on the isotope-separator-on-line (ISOL) target work, particularly the radiation transport aspects of the two-step fission target design. Simulations were performed with the PHITS, MCNPX, and MARS15 computer codes. A two-step ISOL target considered here consists of a mercury or tungsten primary target in which primary beam interactions release neutrons, which in turn induce fissions—and produce rare isotopes—in the secondary target filled with fissionable material. Three primary beams were considered: 1-GeV protons, 622-MeV/u deuterons, and 777-MeV/u 3He ions. The proton and deuterium beams were found to be about equivalent in terms of induced fission rates and heating rates in the target, while the 3He beam, without optimizing the target geometry, was less favorable, producing about 15% fewer fissions and about 50% higher heating rates than the proton beam at the same beam power.

A new infiltration method for coating highly permeable matrices with compound materials for high-power isotope-separator-on-line production target applications

International Nuclear Information System (INIS)

Kawai, Y.; Bilheux, Jean-Christophe; Stracener, Daniel W; Alton, Gerald D

2005-01-01

A new infiltration coating method has been conceived for uniform and controlled thickness deposition of target materials onto highly permeable, complex-structure matrices to form short-diffusion-length isotope-separator-on-line (ISOL) production targets for radioactive ion beam research applications. In this report, the infiltration technique is described in detail and the universal character of the technique illustrated in the form of SEMs of several metal-carbide, metal-oxide and metal-sulfide targets for potential use at present or future radioactive ion beam research facilities

Future prospects of reactor Ra at VINCA institute

International Nuclear Information System (INIS)

Davidovic, M.; Babic-Stojic, B.; Dobrijevic, R.

1997-01-01

Reactor RA at Nuclear Research Institute Vinca belongs to a group of the medium thermal neutron flux reactors, according to classification at end of nineties. At the beginning reactor RA has been used as a powerful source of neutrons and gamma-quanta for various experiments (interaction of neutrons and gamma-quanta with materials) and for production of artificial radioactive materials for commercial use. Very successful utilization of this neutron spectrum has been in its use for structural studies of crystal materials and liquid metals, for magnetic structure studies of various magnetic materials, as well as, dynamic properties of ferro magnetics, ferroelectrics, etc. This kind of spectrometers still exist at reactor RA and with an improved detection system could be used again if reactor starts functioning. Besides this, a part of activity was devoted to construction of neutron guide tubes for thermal neutrons and this could also be accomplished relatively easy in the future. A part of the activities of the reactor should in the future be devoted to the training of students in the field of solid state physics and nuclear physics. Particular attention will be paid to the use of established technologies in production of radioactive isotopes and a new class of isotopes for custom use will be developed as well as highly commercial and prospective products (silicon doping, radiography, etc.). (author)

Development of High Flux Isotope Reactor (HFIR) subcriticality monitoring methods

International Nuclear Information System (INIS)

Rothrock, R.B.

1991-01-01

Use of subcritical source multiplication measurements during refueling has been investigated as a possible replacement for out-of-reactor subcriticality measurements formerly made on fresh HFIR fuel elements at the ORNL Critical Experiment Facility. These measurements have been used in the past for preparation of estimated critical rod positions, and as a partial verification, prior to reactor startup, that the requirements for operational shutdown margin would be met. Results of subcritical count rate data collection during recent HFIR refuelings and supporting calculations are described illustrating the intended measurement method and its expected uncertainty. These results are compared to historical uses of the out-of-reactor core measurements and their accuracy requirements, and a planned in-reactor test is described which will establish the sensitivity of the method and calibrate it for future routine use during HFIR refueling. 2 refs., 1 fig., 2 tabs

The South African isotope facility project

Science.gov (United States)

Bark, R. A.; Barnard, A. H.; Conradie, J. L.; de Villiers, J. G.; van Schalkwyk, P. A.

2018-05-01

The South African Isotope Facility (SAIF) is a project in which iThemba LABS plans to build a radioactive-ion beam (RIB) facility. The project is divided into the Accelerator Centre of Exotic Isotopes (ACE Isotopes) and the Accelerator Centre for Exotic Beams (ACE Beams). For ACE Isotopes, a high-current, 70 MeV cyclotron will be acquired to take radionuclide production off the existing Separated Sector Cyclotron (SSC). A freed up SSC will then be available for an increased tempo of nuclear physics research and to serve as a driver accelerator for the ACE Beams project, in which protons will be used for the direct fission of Uranium, producing beams of fission fragments. The ACE Beams project has begun with "LeRIB" - a Low Energy RIB facility, now under construction. In a collaboration with INFN Legnaro, the target/ion-source "front-end" will be a copy of the front-end developed for the SPES project. A variety of targets may be inserted into the SPES front-end; a uranium-carbide target has been designed to produce up to 2 × 1013 fission/s using a 70 MeV proton beam of 150 µA intensity.

Enrichment of {sup 15}N and {sup 10}B isotopes by chemical exchange process

Energy Technology Data Exchange (ETDEWEB)

D` Souza, A B; Sonwalkar, A S; Subrahmanyam, B V; Valladares, B A [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

Many processes are available for separation of stable isotopes like distillation, chemical exchange, thermal diffusion, gaseous diffusion, centrifuge etc. Chemical exchange process is eminently suitable for separation of isotopes of light elements. Work done on separation and enrichment of two of the stable isotopes viz. {sup 15}N and {sup 10}B in Chemical Engineering Division is presented. {sup 15}N is widely used as a tracer in agricultural research and {sup 10}B is used in nuclear industry as control rod material, soluble reactor poison, neutron detector etc. The work on {sup 15}N isotope resulted in a pilot plant, which was the only source of this material in the country for many years and later it was translated into a production plant as M/s. RCF Ltd. The work done on the ion-exchange process for enrichment of {sup 10}B isotope which is basically a chemical exchange process, is now being updated into a pilot plant to produce enriched {sup 10}B to be used as soluble reactor poison. (author). 5 refs., 2 figs., 3 tabs.

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Rosenthal, Murray Wilford [ORNL

2009-08-01

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

RA research reactor - potentials and prospective

International Nuclear Information System (INIS)

Sotic, O.

1984-01-01

Since December 1959, the RA reactor was operated successfully, except for a few shorter periods needed for maintenance and a four longer shutdown periods caused by decrease in the heavy water quality. Accordingly, reconstruction of some reactor systems was started at the beginning of this decad, as well as increase of its experimental potential which would enable its efficient reliable operation in the future period. Reconstruction is concerned with emergency core cooling system, special ventilation system, and modernization of the reactor instrumentation. Improvement of the experimental potential is related to modifications of the neutron scattering instruments. Development of methods for isotope production is described as well. Design of the reactor experimental loop with external cooling system will be of significant importance in improvement of reactor potential in the future

Hydrogen isotope separation by cryogenic distillation method

International Nuclear Information System (INIS)

Hayakawa, Nobuo; Mitsui, Jin

1987-01-01

Hydrogen isotope separation in fusion fuel cycle and tritium recovery from heavy water reactor are very important, and therefore the early establishment of these separation techniques are desired. The cryogenic distillation method in particular is promising for the separation of hydrogen isotope and the recovery of high concentrated tritium. The studies of hydrogen isotope separation by cryogenic distillation method have been carried out by using the experimental apparatus made for the first time in Japan. The separation of three components (H 2 -HD-D 2 ) under total reflux conditions was got by using the packing tower of 500 mm height. It was confirmed that the Height Equivalent Theoretical Plate (HETP) was 20 - 30 mm for the vapor's line velocity of 20 - 80 mm/s. (author)

Selective photoionization of gadolinium isotopes with a polarized laser

International Nuclear Information System (INIS)

Le Guyadec, E.

1990-06-01

The aim of this study is the use of gadolinium 157 as burnable poison in nuclear reactors. Spectroscopic isotopic displacements between Gd 156 and Gd 157 are low and the separation method studied is based on differentiated behavior, concerning polarized light, of even and odd gadolinium isotopes coming from their difference of nuclear spin. On this principle is based the simplest photoionization scheme. Selective ionization of odd isotopes is realized from the fundamental state with three resonating photons colinearly polarized. The experimental study confirms the possibility of efficient photoionization. The measured selectivity between Gd 157 and even isotope is over 48 in defined conditions because it can be destroyed by a magnetic field or if photons are not well polarized. Calculations and observations are in good agreement. Odd gadolinium isotope separation is feasible and effects preventing separation are evidenced [fr

High Flux Isotope Reactor cold neutron source reference design concept

International Nuclear Information System (INIS)

Selby, D.L.; Lucas, A.T.; Hyman, C.R.

1998-05-01

In February 1995, Oak Ridge National Laboratory's (ORNL's) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH 2 ) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH 2 cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept

High flux isotope reactor cold source preconceptual design study report

International Nuclear Information System (INIS)

Selby, D.L.; Bucholz, J.A.; Burnette, S.E.

1995-12-01

In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH 2 moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project

An analysis of water reactor burnup data with the METHUSELAH II code

International Nuclear Information System (INIS)

Floyd, M.; Hicks, D.

1964-10-01

The METHUSELAH II code has been used to predict long term reactivity and isotopic changes in the YANKEE, Dresden and NRX reactors. In general it is shown that there is a satisfactory measure of agreement and the first core lives of YANKEE and Dresden appear well predicted. However there are discrepancies in the isotopic composition of the plutonium formed which appear to be correlated with the degree of hardness of the reactor spectrum. It is demonstrated that plausible changes in nuclear data could reduce the discrepancies. (author)

Markets for reactor-produced non-fission radioisotopes

International Nuclear Information System (INIS)

Bennett, R.G.

1995-01-01

Current market segments for reactor produced radioisotopes are developed and reported from a review of current literature. Specific radioisotopes studied in is report are the primarily selected from those with major medical or industrial markets, or those expected to have strongly emerging markets. Relative market sizes are indicated. Special emphasis is given to those radioisotopes that are best matched to production in high flux reactors such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. A general bibliography of medical and industrial radioisotope applications, trends, and historical notes is included

New possibilities of the isotope distribution examination in irradiated absorbing materials using secondary ion mass spectrometry method

International Nuclear Information System (INIS)

Goncharenko, Y. D.; Evseev, L.A.; Risovany, V.D.

2005-01-01

The SIMS technique (with using a linear analysis and 2D surface imaging) has been to measure the radial distribution of the boron isotope ratio in the boron carbide pellets irradiated in the fast reactor. It was revealed that a radial distribution of isotope ratio in the boron carbide pellets is significantly different after irradiation in fast and thermal reactors. It was showed the advisability of using ion images for such examinations. (Author)

Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

International Nuclear Information System (INIS)

Washington, J.; King, J.; Shayer, Z.

2017-01-01

Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO_2, Pu_3Si_2, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO_2) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO_2), Pu_0_._3_1ZrH_1_._6Th_1_._0_8, and PuZrO_2MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B_4C, CdO, Dy_2O_3, Er_2O_3, Eu_2O_3, Gd_2O_3, HfO_2, In_2O_3, Lu_2O_3, Sm_2O_3, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO_2MgO (8 wt% Pu) target fuel with a coating of Lu_2O_3 resulted in the highest rate of plutonium transmutation with the greatest reduction in curium

Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Washington, J., E-mail: jwashing@gmail.com [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Shayer, Z., E-mail: zshayer@mines.edu [Department of Physics, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States)

2017-03-15

Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO{sub 2}, Pu{sub 3}Si{sub 2}, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO{sub 2}) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO{sub 2}), Pu{sub 0.31}ZrH{sub 1.6}Th{sub 1.08}, and PuZrO{sub 2}MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B{sub 4}C, CdO, Dy{sub 2}O{sub 3}, Er{sub 2}O{sub 3}, Eu{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, HfO{sub 2}, In{sub 2}O{sub 3}, Lu{sub 2}O{sub 3}, Sm{sub 2}O{sub 3}, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO{sub 2}MgO (8 wt% Pu) target

Reactor-based management of used nuclear fuel: assessment of major options.

Science.gov (United States)

Finck, Phillip J; Wigeland, Roald A; Hill, Robert N

2011-01-01

This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms). Copyright © 2010 Health Physics Society

Analysis and modeling of flow-blockage-induced steam explosion events in the high-flux isotope reactor

International Nuclear Information System (INIS)

Taleyarkhan, R.P.; Georgevich, V.; Nestor, C.W.; Gat, U.; Lepard, B.L.; Cook, D.H.; Freels, J.; Chang, S.J.; Luttrell, C.; Gwaltney, R.C.

1994-01-01

This article provides a perspective overview of the analysis and modeling work done to evaluate the threat from steam explosion loads in the High-Flux Isotope Reactor (HFIR) during flow blockage events. The overall work scope included modeling and analysis of core-melt initiation, melt propagation, bounding and best-estimate steam explosion energetics, vessel failure from fracture, bolts failure from exceedance of elastic limits, and, finally, missile evolution and transport. Aluminum ignition was neglected. Evaluations indicated that a thermally driven steam explosion with more than 65 MJ of energy insertion in the core region over several milliseconds would be needed to cause a sufficiently energetic missile with a capacity to cause early confinement failure. This amounts to about 65% of the HFIR core mass melting and participating in a steam explosion. Conservative melt propagation analyses have indicated that at most only 24% of the HFIR core mass could melt during flow blockage events under full-power conditions. 19 refs., 11 figs

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.

Dissolution Flowsheet for High Flux Isotope Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U₃O₈) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H₂. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H₂) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H₂ and other permanent gases in the dissolution offgas, allowing the development of H₂ generation rate profiles. The H₂ generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

PRISM reactor. An option for plutonium disposition?

Energy Technology Data Exchange (ETDEWEB)

Fehlinger, Sebastian; Friess, Friederike; Kuett, Moritz [IANUS, Technische Universitaet Darmstadt (Germany)

2015-07-01

The Power Reactor Innovative Small Module (PRISM) is sodium cooled fast reactor model. The energy output depends on the core configuration, however with an energy output of approximately 300 MWe, the PRISM reactor belongs to the class of small modular reactors. Beside using the reactor as a breeder reactor or for the transmutation of nuclear waste, it might also be used as a burner reactor for separated plutonium. This includes for example U.S.-American excess weapon-grade plutonium as well as separated reactor-grade plutonium. Recently, there has been an ongoing discussion in GB to use the PRISM reactor to dispose their excess civilian plutonium. Depending on the task, the core configuration varies slightly. We will present different layouts and the matching MCNP models, these models can then be used to conduct depletion calculations. From these results, analysis of the change in the plutonium isotopics in the spent fuel, the amount of fissioned plutonium, and the possible annual plutonium throughputs is possible.

Application of ICP-MS and AMS for determination of Pu- and U-isotope ratios for source identification

Energy Technology Data Exchange (ETDEWEB)

Skipperud, L. (Norwegian Univ. of Life Sciences, Isotope Lab.. Dept. of Plant and Environmental Sciences, AAs (Norway))

2010-03-15

Full text: Anthropogenic plutonium has been introduced into the environment over the past 50 years as the result of the detonation of nuclear weapons and operational releases from the nuclear industry. In the Arctic environment, the main source of plutonium is from atmospheric weapons testing, which have resulted in a relatively uniform, underlying global distribution of plutonium. Plutonium isotope ratios are known to vary with reactor type, nuclear fuel-burn up time, neutron flux, and energy, and for fallout from nuclear detonations, weapon type and yield. Weapons-grade plutonium is characterized by a low content of the 240Pu isotope, with 240Pu/239Pu isotope ratio less than 0.05. In contrast, both global weapons fallout and spent nuclear fuel from civil reactors have higher 240Pu/239Pu isotope ratios (civil nuclear power reactors have 240Pu/239Pu atom ratios of between about 0.2-1). Thus, different sources often exhibit characteristic plutonium isotope ratios and these ratios can be used to identify the origin of contamination, calculate inventories, or follow the migration of contaminated sediments and waters. The measurement of the plutonium-isotope ratios in these studies offers both a means of identifying the origin of radionuclide contamination and the influence of the various nuclear installations on inputs to the Arctic, as well as a potential method for following the movement of water and sediment loads in the rivers. The present paper presents results from determination of plutonium concentrations and isotope ratios in sediment samples collected during various expeditions to the Kara Sea, the Ob and Yenisey estuaries and their river systems and also Pu isotope ratios in the near area of Mayak PA. Weapons-grade plutonium is characterized by a low content of the Pu-240 isotope, with Pu-240/Pu-239 isotope ratio less than 0.05. In contrast, both global weapons fallout and spent nuclear fuel from civil reactors have higher Pu-240/Pu-239 isotope ratios, and

The Y-12 National Security Complex Foreign Research Reactor Uranium Supply Production

Energy Technology Data Exchange (ETDEWEB)

Nelson, T. [Nuclear Technology and Nonproliferation Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States); Keller, A.P. [Disposition and Supply Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States)

2011-07-01

The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the National Nuclear Security Administration (NNSA) HEU Disposition, the Reduced Enrichment Research and Test Reactors (RERTR), and the United States (U.S.) FRR Spent Nuclear Fuel (SNF) Acceptance Programs. The FRR Supply Program supports the important U.S. government nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to Low-Enriched Uranium (LEU) fuel under the RERTR Program. The NNSA Y-12 Site Office maintains the prime contracts with foreign government agencies for the supply of LEU for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. In addition to uranium metal feedstock for fuel fabrication, Y-12 can produce LEU in different forms to support new fuel development or target fabrication for medical isotope production. With production improvements and efficient delivery preparations, Y-12 continues to successfully support the global research reactor community. (author)

C.E.C. - cod for calculus of the evolution fuel for thermal reactors

International Nuclear Information System (INIS)

Biciolla, L.; Marcu, G.; Mociornita, G.

1975-01-01

The study of ''burnup'' into thermal reactor involves two main aspects: the economic one and another regarding the reactor operation, its stability and control. In the CEC-code written in FORTRAN IV language was analysed the change of the isotopic composition of nuclear fuel from thermal reactor during its operation

High Flux Isotope Reactor quarterly report, July--September 1975

International Nuclear Information System (INIS)

McCord, R.V.; Corbett, B.L.

1975-01-01

The replacement of the permanent beryllium reflector was completed this quarter. The reactor was shut down for 87 days for this maintenance operation. Erosion of the sealing surface at the stainless steel adaptor flange on the HB-1 beam tube facility was confirmed. A soft metallic O-ring was used to effect a seal when this facility was reassembled. A comprehensive inspection of the normally inaccessible parts of the reactor pressure vessel was made. No abnormalities were found

Separation of carrier-free rhodium isotopes from ruthenium cyclotron targets by the extraction of nitrosylruthenium from hydrochloric acid solution

International Nuclear Information System (INIS)

Haasbroek, F.J.; Strelow, F.W.E.; Van der Walt, T.N.

1981-01-01

A method is presented for the separation of rhodium isotopes from ruthenium cyclotron targets. After bombardment with deuterons and dissolution of the target material, the ruthenium is converted into a nitrosyl complex by treatment with hydroxylammonium chloride. Aluminium and other elements which have been introduced in the dissolution step, are separated by cation exchange. Ruthenium is then separated by extraction with a mixture of tri-n-butyl phosphate and hexane (4:1), leaving the rhodium in the aqueous phase. No ruthenium is found in the rhodium fraction and the recovery of rhodium is better than 90 per cent [af

Scientific upgrades at the high flux isotope reactor at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Selby, D.L.; Garrett, D.L.; Lucas, A.T.; Reeves, M.E.

2001-01-01

The United States Department of Energy is sponsoring a number of projects that will provide scientific upgrades to the neutron science facilities associated with the high flux isotope reactor (HFIR) located at Oak Ridge National Laboratory. Funding for the first upgrade project was initiated in 1996 and all presently identified upgrade projects are expected to be completed by the end of 2003. The upgrade projects include: 1) larger beam tubes, 2) a new monochromator drum for the HB-1 beam line, 3) a new HB-2 beam line system that includes one thermal guide and a new monochromator drum, 4) new instruments for the HB-2 beamline, 5) a new monochromator drum for the HB-3 beam line, 6) a supercritical hydrogen cold source system to be retrofitted into the HB-4 beam tube, 7) a 3.5 kW refrigeration system at 20 K to support the cold source and a new building to house it, 8) a new HB-4 beam line system composed of four cold neutron guides with various mirror coatings and associated shielding, 9) a number of new instruments for the cold beams including two new SANS instruments, and 10) construction of support buildings. This paper provides a short summary of these projects including their present status and schedule. (orig.)

Design and safety considerations for the 10 MW(t) multipurpose TRIGA reactor in Thailand

International Nuclear Information System (INIS)

Razvi, J.; Bolin, J.M.; Saurwein, J.J.; Whittemore, W.L.; Proongmuang, S.

1999-01-01

General Atomics (GA) is constructing the Ongkharak Nuclear Research Center (ONRC) near Bangkok, Thailand for the Office of Atomic Energy for Peace. The ONRC complex includes the following: A multipurpose 10 MW(t) research reactor; An Isotope Production Facility; Centralized Radioactive Waste Processing and Storage Facilities. The Center is being built 60-km northeast of Bangkok, with a 10 MW(t) TRIGA type research reactor as the centerpiece. Facilities are included for neutron transmutation doping of silicon, neutron capture therapy neutron beam research and for production of a variety of radioisotopes. The facility will also be utilized for applied research and technology development as well as training in reactor operations, conduct of experiments and in reactor physics. The multipurpose, pool-type reactor will be fueled with high-density (45 wt%), low-enriched (19.7 wt%) uranium-erbium-zirconium-hydride (UErZrH) fuel rods, cooled and moderated by light water, and reflected by beryllium and heavy water. The general arrangement of the reactor and auxiliary pool structure allows irradiated targets to be transferred entirely under water from their irradiation locations to the hot cell, then pneumatically transferred to the adjacent Isotope Production Facility for processing. The core configuration includes 4 x 4 array standard TRIGA fuel clusters, modified clusters to serve as fast-neutron irradiation facilities, control rods and an in-core Ir-192 production facility. The active core is reflected on two sides by beryllium and on the other two sides by D 2 O. Additional irradiation facilities are also located in the beryllium reflector blocks and the D 2 O reflector blanket. The fuel provides the fundamental safety feature of the ONRC reactor, and as a result of all the well established accident-mitigating characteristics of the UErZrH fuel itself (large prompt negative temperature coefficient of reactivity, fission product retention and chemical stability), a

Optimization by simulation of the coupling between a sub-critical reactor and its spallation source. Towards a pilot reactor

International Nuclear Information System (INIS)

Kerdraon, D.

2001-10-01

Accelerator Driven Systems (ADS), based on a proton accelerator and a sub-critical core coupled with a spallation target, offer advantages in order to reduce the nuclear waste radiotoxicity before repository closure. Many studies carried out on the ADS should lead to the definition of an experimental plan which would federate the different works in progress. This thesis deals with the neutronic Monte Carlo simulations with the MCNPX code to optimize such a system in view of a pilot reactor building. First, we have recalled the main neutronic properties of an hybrid reactor. The concept of gas-cooled eXperimental Accelerator Driven System (XADS) chosen for our investigations comes from the preliminary studies done by the Framatome company. In order to transmute minor actinides, we have considered the time evolution of the main fuels which could be reasonably used for the demonstration phases. The neutronic parameters of the reactor, concerning minor actinide transmutation, are reported. Also, we have calculated the characteristic times and the transmutation rates in the case of 99 Tc and 129 I isotopes. We have identified some neutronic differences between an experimental and a power ADS according to the infinite multiplication coefficient, the shape factor and the level of flux to extend the demonstrator concept. We have proposed geometric solutions to keep the radial shape factor of a power ADS acceptable. In the last part, beyond the experimental XADS scope, we have examined the possible transition towards an uranium/thorium cycle based on Molten Salt Reactors using a power ADS in order to generate the required 233 U proportion. (author)

SCK-CEN increases production of medical isotopes by half

International Nuclear Information System (INIS)

Ponsard, B.; Leysen, P.; Janssens, J.

2010-01-01

It is impossible to imagine the medical world today without radioisotopes, and due to rapid technological progress in nuclear medicine their use is still on the rise. An important role of research reactors is the production of molybdenum-99. Around the world this is done primarily by five nuclear research reactors, one of which is the BR2 reactor of SCK-CEN. As a result of checks and maintenance work on three other reactors, for a few years there has been a serious crisis in the availability of this medical isotope. In order to guarantee the worldwide supply of radioisotopes, SCK-CEN expanded its production of molybdenum-99 by 50 percent in 2010.

Progress in the Use of Isotopes: The Atomic Triad - Reactors, Radioisotopes and Radiation

Science.gov (United States)

Libby, W. F.

1958-08-04

Recent years have seen a substantial growth in the use of isotopes in medicine, agriculture, and industry: up to the minute information on the production and use of isotopes in the U.S. is presented. The application of radioisotopes to industrial processes and manufacturing operations has expanded more rapidly than any one except its most ardent advocates expected. New uses and new users are numerous. The adoption by industry of low level counting techniques which make possible the use of carbon-14 and tritium in the control of industrial processes and in certain exploratory and research problems is perhaps most promising of current developments. The latest information on savings to industry will be presented. The medical application of isotopes has continued to develop at a rapid pace. The current trend appears to be in the direction of improvements in technique and the substitution of more effective isotopes for those presently in use. Potential and actual benefits accruing from the use of isotopes in agriculture are reviewed. The various methods of production of radioisotopes are discussed. Not only the present methods but also interesting new possibilities are covered. Although isotopes are but one of the many peaceful uses of the atom, it is the first to pay its way. (auth)

Neutron resonance parameters of CM isotopes

International Nuclear Information System (INIS)

Belanova, T.S.; Kolesov, A.G.; Poruchikov, V.A.

1977-01-01

The total neutron cross sections of isotopes 244, 245, 246, 248 Curium have been measured on reactor CM-2 using the time-of-flight method. Single-level Breit-Wigner resonance parameters: energy E 0 , neutron width 2g GITAn, total width GITA, total neutron cross section in resonance sigma 0 have been obtained by the shape and area methods

Production, separation and target preparation of 171Tm an 147Pm for neutron cross section measurements

CERN Document Server

Heinitz, S; Schumann, D; Dressler, R; Kivel, N; Guerrero, C; Köster, U; Tessler, M; Paul, M; Halfon, S

2015-01-01

The knowledge of the neutron capture cross sections of s-process branching point isotopes represents a basic requirement for the understanding of star evolution. Since such branching point isotopes are by definition radioactive, the measurement of their cross sections from thermal to stellar energies becomes a challenging task. Considerable amounts of material have to be produced, representing a significant radioactive hazard. We report here on the production and separation of 3.5 mg 171Tm from 240 mg 170Er2O3 and 72 µg 147Pm from 100 mg 146Nd2O3 irradiated at the ILL high flux reactor. Thin targets were prepared with high chemical and radioisotopic purity suitable for neutron capture measurements at n_TOF CERN and the SARAF-LiLiT facility.

Manufacturing W fibre-reinforced Cu composite pipes for application as heat sink in divertor targets of future nuclear fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Mueller, Alexander v.; You, Jeong-Ha [Max-Planck-Institut fuer Plasmaphysik, 85748 Garching (Germany); Ewert, Dagmar [Institut fuer Textil- und Verfahrenstechnik Denkendorf, 73770 Denkendorf (Germany); Siefken, Udo [Louis Renner GmbH, 85221 Dachau (Germany)

2016-07-01

An important plasma-facing component (PFC) in future nuclear fusion reactors is the so-called divertor which allows power exhaust and removal of impurities from the main plasma. The most highly loaded parts of a divertor are the target plates which have to withstand intense particle bombardment. This intense particle bombardment leads to high heat fluxes onto the target plates which in turn lead to severe thermomechanical loads. With regard to future nuclear fusion reactors, an improvement of the performance of divertor targets is desirable in order to ensure reliable long term operation of such PFCs. The performance of a divertor target is most closely linked to the properties of the materials that are used for its design. W fibre-reinforced Cu (Wf/Cu) composites are regarded as promising heat sink materials in this respect. These materials do not only feature adequate thermophysical and mechanical properties, they do also offer metallurgical flexibility as their microstructure and hence their macroscopic properties can be tailored. The contribution will point out how Wf/Cu composites can be used to realise an advanced design of a divertor target and how these materials can be fabricated by means of liquid Cu infiltration.

Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Bodey, Isaac T [ORNL

2014-05-01

Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

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.

Computer simulation of multi-elemental fusion reactor materials

International Nuclear Information System (INIS)

Voertler, K.

2011-01-01

Thermonuclear fusion is a sustainable energy solution, in which energy is produced using similar processes as in the sun. In this technology hydrogen isotopes are fused to gain energy and consequently to produce electricity. In a fusion reactor hydrogen isotopes are confined by magnetic fields as ionized gas, the plasma. Since the core plasma is millions of degrees hot, there are special needs for the plasma-facing materials. Moreover, in the plasma the fusion of hydrogen isotopes leads to the production of high energetic neutrons which sets demanding abilities for the structural materials of the reactor. This thesis investigates the irradiation response of materials to be used in future fusion reactors. Interactions of the plasma with the reactor wall leads to the removal of surface atoms, migration of them, and formation of co-deposited layers such as tungsten carbide. Sputtering of tungsten carbide and deuterium trapping in tungsten carbide was investigated in this thesis. As the second topic the primary interaction of the neutrons in the structural material steel was examined. As model materials for steel iron chromium and iron nickel were used. This study was performed theoretically by the means of computer simulations on the atomic level. In contrast to previous studies in the field, in which simulations were limited to pure elements, in this work more complex materials were used, i.e. they were multi-elemental including two or more atom species. The results of this thesis are in the microscale. One of the results is a catalogue of atom species, which were removed from tungsten carbide by the plasma. Another result is e.g. the atomic distributions of defects in iron chromium caused by the energetic neutrons. These microscopic results are used in data bases for multiscale modelling of fusion reactor materials, which has the aim to explain the macroscopic degradation in the materials. This thesis is therefore a relevant contribution to investigate the

The global threat reduction initiative and conversion of isotope production to LEU targets

International Nuclear Information System (INIS)

Kuperman, A. J.

2005-01-01

The U.S. Global Threat Reduction Initiative (GTRI) has given a decisive impetus to the RERTR program's longstanding goal of converting worldwide production of medical radioisotopes from reliance on bomb-grade, highly enriched uranium (HEU) to low-enriched uranium (LEU) unsuitable for weapons. Although the four major; isotope producers continue to resist calls for conversion, they face mounting pressure from a variety of fronts including: (1) GTRI; (2) a related, multilateral U.S. initiative to forge agreement on conversion among the states that are home to the major producers; (3) an IAEA effort to provide technical assistance that will facilitate large-scale production of medical isotopes using LEU by producers who seek to do so; (4) planned production in the United States of substantial quantities of medical isotopes using LEU; and (5) pending U.S. legislation that would prohibit the export of HEU for production of isotopes as soon as alternative, LEU-produced isotopes are available. Accordingly, it now appears inevitable that worldwide isotope production will be converted from reliance on HEU to LEU. The only remaining question is which producers will be the first to reliably deliver sizeable quantities of LEU-produced isotopes and thereby capture global market share from the others. (author)

Research reactor spent fuel management in Argentina

International Nuclear Information System (INIS)

Audero, M.A.; Bevilacqua, A.M.; Mehlich, A.M.; Novara, O.

2002-01-01

The research reactor spent fuel (RRSF) management strategy will be presented as well as the interim storage experience. Currently, low-enriched uranium RRSF is in wet interim storage either at reactor site or away from reactor site in a centralized storage facility. High-enriched uranium RRSF from the centralized storage facility has been sent to the USA in the framework of the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The strategy for the management of the RRSF could implement the encapsulation for interim dry storage. As an alternative to encapsulation for dry storage some conditioning processes are being studied which include decladding, isotopic dilution, oxidation and immobilization. The immobilized material will be suitable for final disposal. (author)

Production of exotic, short lived carbon isotopes in ISOL-type facilities

CERN Document Server

Franberg, Hanna; Köster, Ulli; Ammann, Markus

2008-01-01

The beam intensities of short-lived carbon isotopes at Isotope Separation On-Line (ISOL) facilities have been limited in the past for technical reasons. The production of radioactive ion beams of carbon isotopes is currently of high interest for fundamental nuclear physics research. To produce radioactive ions a target station consisting of a target in a container connected to an ion source via a transfer line is commonly used. The target is heated to vaporize the product for transport. Carbon in elementary form is a very reactive element and react strongly with hot metal surfaces. Due to the strong chemisorption interaction, in the target and ion source unit, the atoms undergo significant retention on their way from the target to the ion source. Due to this the short lived isotopes decays and are lost leading to low ion yields. A first approach to tackle these limitations consists of incorporating the carbon atoms into less reactive molecules and to use materials for the target housing and the transfer line ...

Building a dynamic code to simulate new reactor concepts

International Nuclear Information System (INIS)

Catsaros, N.; Gaveau, B.; Jaekel, M.-T.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.