Cadmium Filters Research Articles

The VENUS imaging beamline at the Spallation Neutron Source: layout, expected performance and status of the construction project

Located at the Spallation Neutron Source, VENUS is an instrument optimized for wavelength-dependent neutron imaging techniques, namely Bragg edge and resonance imaging, across a broad range of neutron energies (from meV to hundreds of eV). The VENUS construction project started in Fall 2018, establishing a conceptual design of the major components and timeline for purchasing, testing, and installation. Completion of the VENUS instrument is anticipated late 2024. This instrument comprises components such as a set of boron carbide apertures designed for both thermal/cold and epithermal neutrons, a suite of choppers (T0 and single disks), beam collimators, and a cadmium filter. The time-of-flight detector at VENUS will be the micro-channel-plate Timepix detector and will be positioned 25 m away from the source. This manuscript describes the instrument general layout, explains the choice of the moderator, and specifies the VENUS key capabilities. It also provides the status of the construction project with installed components. Discussions about software development are also presented.

Irradiation performance of nonfertile (Pu-MA-Zr) fast reactor metal fuels

This work was part of a program begun in 2001 to develop advanced nuclear fuels, originally as carriers for plutonium and minor actinides (neptunium, curium, and americium) taken from spent commercial light-water reactors (LWR) so that the plutonium and minor actinides could be ‘burned’ or transmuted in an accelerator or a fast nuclear reactor. A central part of these experiment programs has been the development of advanced fast reactor fuels, because a fast reactor was considered the most efficient vehicle to transmute the actinide waste products, and metallic fuels is a central focus of these tests. An experiment design was developed in which a thermal test reactor, the Advanced Test Reactor (ATR), was used to test small fuel pin prototypes, by creating areas in the core shielded by cadmium filters to produce a largely epithermal and fast neutron spectrum environment in which the pins could be irradiated. The results of non-fertile metallic fuel (no uranium) tests are presented here.Pu-Am-Np-Zr fuels were irradiated to fission densities up to 33 × 1020 fission/cm3 and Pu-239 depletions of up to 39%. The depletions were created by roughly 2/3 by fission and 1/3 by transmutation neutron capture. Up to five fuel ‘rodlets’ were irradiated in three sealed capsules stacked axially in the core, and the peak cladding temperatures ranged from 300°C to 500°C, depending on axial location as those near the core centerline are operating hotter and to higher fission densities. Several post-irradiation examinations (precision gamma scanning and fission gas release) were similar to other historical metal fuel experiments in fast reactors. However, optical metallography indicated that two of the rodlets had breached. The exact reasons are unclear. Due to the design of this irradiation experiment a rodlet breach could have increased the temperature in others in the same capsule by contaminating the thermal gap helium with heavier and less conductive fission product gases. Some of those rodlets showed high amounts of fuel/cladding chemical interaction (FCCI).

A Passive Neutron Dosemeter for High Energy Accelerators

A passive neutron dosemeter (PND) for high-energy accelerators is described. The slide of the DVGN-01 personal dosemeter, which is placed at the center of a polyethylene moderator, is used as a thermal-neutron detector. The moderator includes a lead insert, which increases the sensitivity of the dosemeter to high-energy neutrons, and a cadmium filter. Optimization calculations of the dosemeter design parameters have been performed with respect to the energy dependence of the response in the range from the thermal energy to 1 GeV. The results of the calculations are in good agreement with the experimental data obtained in low-energy neutron reference fields. Comparative measurements using the PND and a low-energy passive dosemeter have been performed in the neutron fields outside the shield of the 70-GeV U-70 proton accelerator at the Institute for High Energy Physics.

Fission rates of nuclides in the neutron field formed on the surface of a massive lead target irradiated with protons at energy of 5.0 Gev

Fission rates of the nuclides Th-232, U-235 (with and without cadmium filter), U-234, U-236, Np-237 and U-238 were measured on the side surface of a massive lead target irradiated at the Syncrophasotron accelerator (JINR, Dubna) with protons at kinetic energy of 5.0 GeV. These measurements were carried out using the technique of solid-state nuclear track detectors. Comparison of the measured values was made with the results obtained by means of Monte-Carlo simulation using the computer code DCM/CEM.

FISSILE MATERIALS DETECTION VIA NEUTRON DIFFERENTIAL DIE-AWAY TECHNIQUE

This work is devoted to the differential die-away technique that is widely used for active detection of fissile materials via pulsed neutron generators. The technique allows direct detection of milligram quantities of uranium-235 and plutonium-239 in objects with volumes up to several cubic meters. Our group has demonstrated this technique, creating a special installation based on the commercially produced ING-07T pulsed neutron generator. The installation includes eight proportional 3 He -counters mounted inside a polyethylene moderator with a cadmium filter, as well as a polyethylene chamber into which a 70-liter container is loaded for inspection. Preliminary testing showed that the minimum detectable mass of unshielded uranium-235 is ∼3 mg, using a 5.108 n/s neutron yield and 8 min measurement time. When the container is filled with neutron absorbing materials, e.g., iron, the minimum detectable mass increases to ∼30 mg. Use of borated screens further increases the minimum mass that can be detected. The tested installation and/or its modifications can be used for control and detection of fissile materials in various applications from luggage inspection to control containers with nuclear fuel cycle radioactive wastes.

Determination of trace concentrations of thorium in uranium oxide matrix by epithermal instrumental neutron activation analysis

An epithermal instrumental neutron activation analysis (EINAA) method using cadmium filter was standardized to determine trace concentrations of thorium in four samples of uranium oxide (U3O8) samples. Samples and thorium standards, wrapped with cadmium foil, were irradiated at a reactor neutron flux of about 1012 cm−2 s−1. Radioactive assay was carried out using a Compton suppressed anticoincidence gamma ray spectrometer consisting of HPGe-BGO detectors coupled to MCA. Concentrations of thorium in these samples were found to be in the range of 15–72 mg kg−1. EINAA results were validated by determining thorium concentrations in uranium matrix by standard addition method. EINAA results were compared with those obtained by two wet chemical methods namely ion chromatography (IC) and inductively coupled plasma atomic emission spectrometry (ICP-AES). The results obtained by the three methods were found to be in good agreement, indicating further validity of the proposed EINAA method.

The use of enriched 6Li and 7Li Lif:Mg,Cu,P glass-rod thermoluminescent dosemeters for linearaccelerator out-of-field radiation dose measurements

(6)LiF:Mg,Cu,P and (7)LiF:Mg,Cu,P glass-rod thermoluminescent dosemeters (TLDs) were used for measurements of out-of-field photon and neutron doses produced by Varian iX linear accelerator. Both TLDs were calibrated using 18-MV X-ray beam to investigate their dose-response sensitivity and linearity. CR-39 etch-track detectors (Luxel+, Landauer) were employed to provide neutron dose data to calibrate (6)LiF:Mg,Cu,P TLDs at various distances from the isocentre. With cadmium filters employed, slow neutrons (<0.5 eV) were distinguished from fast neutrons. The average in-air photon dose equivalents per monitor unit (MU) ranged from 1.5±0.4 to 215.5±94.6 μSv at 100 and 15 cm from the isocentre, respectively. Based on the cross-calibration factors obtained with CR-39 etch-track detectors, the average in-air fast neutron dose equivalents per MU range from 10.6±3.8 to 59.1±49.9 μSv at 100 and 15 cm from the isocentre, respectively. Contribution of thermal neutrons to total neutron dose equivalent was small: 3.1±7.2 μSv per MU at 15 cm from the isocentre.

A neutronic method to determine low hydrogen concentrations in metals

We propose a method for the non-destructive determination of low hydrogen content in metals. The method is based on measurements of neutron inelastic scattering combined with cadmium filters. Determination is simple and the method would allow to construct a mobile device, to perform the analysis “in situ”. We give a brief description of the usual methods to determine low hydrogen contents in solids, paying special attention to those methods supported by neutron techniques. We describe the proposed method, calculations to achieve a better sensitivity, and experimental results.

Measurement of the fast neutron flux in the MNSR inner irradiation site

The WIMSD4 code was used to calculate the fast neutron flux spectrum and the fast neutron fission cross-sections for 238U, using six energy groups ranging from 0.5 to 10 MeV. These results, with the measured radioactivities of the 140Ba, 131I, 103Ru, 95Zr and 97Zr fission products emerging from the fission of the 238U foil covered with a cadmium filter, were used to measure the fast neutron flux in the Syrian Miniature Neutron Source Reactor inner irradiation site.

Investigation of the Neutron Fields on the Surface of a Massive Lead Target Irradiated with 1.5 and 5 GeV Protons

The spatial and energy characteristics of the neutron fields produced on the surface of a massive lead target irradiated with 1.5 and 5 GeV protons were investigated. In the course of the experiment, the distribution of the fission rates of the threshold nuclides 232Th and 209Bi on the lateral surface of the target was obtained in two directions (transverse and longitudinal). The fission rates of 235U (in and without a cadmium filter), 234U, 236U, 237Np, and 238U and the corresponding spectral indices and cadmium ratio for 235U were also measured on the target surface. For the two values of the proton energy, the experimental values were compared with one another and with Monte Carlo simulation results.

Neutron flux micro-distribution and self-shielding measurement in the Syrian Miniature Neutron Source Reactor

Measurement of neutron flux micro distribution within absorbing samples have been made using Dy and Au foils activated in cadmium filter for both thermal and epithermal regions, respectively. Self-shielding effect for highly absorbing material has been investigated for the purpose of activation analysis in the Syrian Miniature Neutron Source Reactor (MNSR). Self-shielding factors for both thermal and resonance regions have been evaluated. Measured and theoretical values of self-shielding factors have been compared. Correction expression of the neutron activation analysis measurement for self-shielding has been proposed.

Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed.

Characteristics of the KUR Heavy Water Neutron Irradiation Facility as a neutron irradiation field with variable energy spectra

The Heavy Water Neutron Irradiation Facility (HWNIF) of the Kyoto University Research Reactor (KUR) was updated in March 1996, mainly for the improvement in neutron capture therapy (NCT). A striking feature of the updated facility is that the energy spectrum of the neutron beam can be controlled from almost pure thermal to epi-thermal, within 5 min by remote control under a continuous reactor operation. This feature is advantageous not only to medical science such as NCT, but also to the other research fields such as physics, engineering, biology, etc. The performance of the updated facility as a neutron irradiation field with variable energy spectra, was characterized. Thermal neutron flux, cadmium ratio, gamma-ray dose rate, etc., at the normal irradiation position for various irradiation modes were determined, mainly on the basis of the measurement using gold activation foils and thermo-luminescent dosimeters (TLDs). The emphasis was on the performance of the new neutron energy spectrum shifter and cadmium thermal neutron filter, that control the mixing ratio of thermal and epi-thermal neutrons, through the change in the heavy water thickness of the spectrum shifter and the aperture size of the cadmium filter. The evaluation of neutron energy spectra at the normal irradiation position was also performed for three representative irradiation modes, in which the neutron intensities are largest of all the irradiation modes. In addition, the irradiation characteristics of two irradiation devices, namely the Irradiation Rail Device and the Remote Patient Carrier, which were updated concurrently with the facility update, were evaluated.

Determination of chlorine, bromine and iodine in rock samples by radiochemical neutron activation analysis

Chlorine, bromine and iodine (hereafter, halogens) were detemined for rock samples by radiochemical neutron activation analysis. The powdered samples and reference standards prepared from chemical reagents were simultaneously irradiated for 10 to 30 minutes with or without a cadmium filter in a TRIGA-II reactor at the Institute for Atomic Energy, Rikkyo University. The samples were subjected to radiochemical procedures of halogens immediately after the irradiation. Iodine was firstly precipitated as PdI2, and chlorine and bromine were successively precipitated as Ag-halides at the same time. In this study, geological standard rocks, sedimentary rocks and meteorites were analyzed for trace halogens. In some Antarctic meteorites, iodine contents were observed to be anomalously high. Chlorine contents also are somewhat high. The overabundance of iodine and chlorine must be caused by terrestrial contamination on the Antarctica.

Track method for personal dosimetry for a nuclear-powered fleet

This work presents the development results of a high-sensitivity personal neutron track dosimeter for nuclear-powered ice-breakers. The dosimeter consists of a uranium-235 target and PET track detector in cadmium filters. Dose per track value of the dosimeter is 0.012 mrem fission −1 with scattering of 15%, affording a minimum measurable dose of 0.1 mrem. The main characteristics of this dosimeter are compared with the TLD-albedo ALNOR dosimeter.

Determination of multielement in Sargasso reference material by instrumental neutron activation analysis.

Concentrations of 38 elements in an environmental reference material Sargasso (NIES No. 9) prepared by the National Institute for Environmental Studies of Japan (NIES) were determined by instrumental neutron activation analysis (NIAA). The Sargasso samples (ca. 150500 mg) were irradiated by thermal neutron without cadmium filter and epithermal neutron with cadmium filter at Musashi Institute of Technology Research Reactor (MITRR). The activated samples were measured by three methods; conventional gamma-ray spectrometry using a coaxial Ge detector, anticoincidence counting spectrometry and coincidence counting spectrometry. Concentrations of 31 elements (Na, Mg, Al, S, Cl, K, Ca, Sc, V, Cr, Mn, Fe, Co, Zn, As, Br, Rb, Sr, Ag, Sb, Te, I, Cs, Ba, Ce, Sm, Eu, Tb, Yb, Au, U) were determined by the conventional combination of thermal neutron irradiation without cadminum filter and the gamma-ray spectrometry method. Concentrations of other 7 element (Ni, Se, Zr, Mo, La, Hf, Th) were determined by other combinations of irradiation and spectrometry method.

Instrumental Neutron Activation Analysis of Vehicle Exhaust Particulates as Environmental Reference Material

Trace elements in a vehicle exhaust particulates environmental reference material (NIES No. 8) prepared by the National Institute for Environmental Studies were determined by an instrumental neutron activation analysis method (INAA). The vehicle exhaust particulates samples (ca. 150-550mg) were irradiated by neutrons for a short time (30s) and a long time (5h) at Musashi Institute of Technology Research Reactor (MITRR). Samples were irradiated by two methods: namely, thermal neutron irradiation without cadmium filter and epithermal neutron irradiation with cadmium filter. Activated samples were measured by four methods: γ-ray spectrometry, anticoincidence counting, coincidence counting and low energy photon spectrometry using a planar Ge detector LEPS). Concentrations of 52 elements were determined by these irradiation and counting methods.

Epithermal neutron activation analysis using the monostandard method

A method is described for epithermal neutron activation analysis of 17 elements in granite rock samples using a single standard. Gold has been used as a single comparator due to its relatively high resonance integral value (I0=400 barn). In addition, it is preferable to Co in order to obtain a large epithermal activation in a short irradiation. The method of calculation is simple and rapid and can be done using a small calculator. Epithermal activation is able to overcome the difficulty arising from changing irradiation position as well as increasing the number of determinable elements by eliminating the interference from undesired isotopes which have relatively high thermal cross section values (σ0), when reactor neutron flux is used. The coupling of epithermal activation with the monostandard method has the advantage of using a small Cd-cover which overcomes most of the difficulties arising in the relative method with large volume cadmium filters.

Neutron Capture and Fission in254gEs

Integral neutron capture and neutron fission cross sections have been measured for the 276-day /sup 254g/ Es. Thermal cross sections and resonance integrals were evaluated using a cadmium filter technique. Capture cross sections were determined from alpha-particle spectrum measurements following neutron irradiations with cobalt flux monitors. Fission cross sections were measured using fission track detection techniques with STTU monitors. The fission cross-section values compared favorably with an absorption cross-section determination from a burnout experiment of SVTEs-SVUEs. The integral neutron capture and fission cross sections determined for /sup 254g/ Es are: sigma /sub c/ /sup th/ = 28.3 + or - 2.5 and I /sub c/ = 18.2 + or - 1.5 b, and sigma /sub F/ /sup th/ = 1970 + or - 200 and I /sub F/ = 1200 + or - 250 b.

Epithermal neutron activation analysis using a boron carbide irradiation filter

The use of boron carbide as a thermal neutron filter in epithermal neutron activation (ENAA) analysis has been investigated. As compared to the use of a cadmium filter, boron provides a greater reduction of activities from elements relatively abundant in terrestrial rocks and fossil fuels, such as Na, La, Sc and Fe. These elements have excitation functions which follow the 1/v law in the 1 to 10 eV lower epithermal region. This enhances the sensitivity of ENAA for elements such as U, Th, Ba and etc. which have strong resonances in the higher epithermal region above 10 eV. In addition, a boron carbide filter has the advantages over cadmium of acquiring a relatively low level of induced activity which poses minimal radiation safety problems, when used for ENAA.