Foil Activation Method Research Articles

Experimental evaluation of neutron spectra in the irradiation fields of a typical material test research reactor

Neutron energy spectra at different regions inside and around the core of Pakistan Research Reactor-1 (PARR-1), a typical swimming pool type material test research reactor, have been evaluated. The method of multiple foil activation was used together

Prediction and measurement of neutron energy spectrum in a material test research reactor

The objective of this work is to calculate and measure the neutron energy spectrum in the Pakistan Research Reactor-1 (PARR-1), which is a typical Material Test Reactor (MTR) fueled with low enriched uranium. The method of multiple foil activation was used with the adjustment code MSITER, a modified version of the code STAY'SL, to obtain the experimentally measured spectrum. The calculated spectrum, which was used as pre-information in the adjustment procedure, was obtained through modeling the core and its surroundings in three-dimensions and the use of the transport code WIMS-D/4 and the diffusion code CITATION. A comparison between the calculated and measured spectra shows that the ratio of measured to calculated flux density in the thermal region ( E ⩽0.5 eV) is 0.998. This ratio is 1.02 for resonance (1/ E) region (0.5 eV < E ⩽ 0.5 MeV) and 0.932 for the fast region ( E > 0.5 MeV). This result is compared with other published results that report an agreement within 10 to 30% between the calculated and measured spectra.

Determination of the hydrogen content of oil samples from Nigeria using an Am–Be neutron source

A 5 Ci Am--Be neutron source-based facility, which utilises the principles of thermal neutron reflection technique in combination with foil activation method, has been used to determine the total hydrogen content of commercial oil samples from Nigeria. With an established detection limit of 0.25 H wt% for oil matrix of volume 600 ml, the total hydrogen contents of the samples were found to be in the range of 11.11--14.22 H wt%. The facility is economical and suitable for the determination of moisture in solid samples.

Search for low-energy upscattering of ultracold neutrons from a beryllium surface

We present results of a search for anomalous low-energy upscattering of ultracold neutrons from a beryllium surface. This upscattering is considered one for the possible reasons for UCN “disappearance” from very cold beryllium bottles, as observed in experiments. The indium foil activation method was used to measure a very low intensity flux of upscattered UCN. The (15–300) m/s velocity range of upscattered UCN is ruled out by these measurements at a confidence level of 90%.

Characteristics of the Kyoto University Lead Slowing-down Spectrometer (KULS) coupled to an electron linac

A lead slowing-down spectrometer coupled to a 46 MeV electron linear accelerator (linac) was installed at Research Reactor Institute, Kyoto University (KURRI). The size of this Kyoto University Lead Slowing-down Spectrometer (KULS) is 1.5 × 1.5 × 1.5 m 3, and it is covered with Cd sheets 0.5 mm thick. One of the eleven experimental holes in the KULS is covered with 10 to 15 cm thick bismuth layers to suppress high energy capture gamma-rays from lead. The characteristics of this KULS have been experimentally obtained and the results are compared with the predicted values by Monte Carlo calculations using the MCNP code. 1) The slowing-down constant K in the relation E = K t 2 between the neutron slowing-down time t and energy E is 190±2 (keV μs 2) for the bismuth hole and 156±2 (keV μs 2) for an ordinary lead hole, respectively. The K values agree with the calculated ones. 2) The measured energy resolution ΔE E at full-width-at-half-maximum (FWHM) was about 40% for both holes, while the calculated values were lower by about 10% than the measured ones in the relevant energy region. 3) The neutron energy spectrum from 0.01 eV to 20 MeV and the spatial distribution of neutrons in the KULS were measured by the foil activation method. The angular neutron spectrum perpendicular to the linac electron beam was also obtained experimentally in the energy range from a few eV to about 10 MeV by the neutron time-of-flight (TOF) method. The measured results are compared with the calculated ones in which we have used the three evaluated nuclear data JENDL-3, ENDL-85 and ENDF/B-IV for lead. Through the comparison a check on the nuclear data has been performed.

96/06041 Modelling of a vibrating reactor boundary and calculation of the induced neutron noise

Neutron spectrum analyses of spallation neutrons are conducted in the accelerator-driven system (ADS) facility at the Kyoto University Critical Assembly (KUCA). High-energy protons (100 MeV) obtained from the fixed field alternating gradient accelerator are injected onto a tungsten target, whereby the spallation neutrons are generated. For neutronic characteristics of spallation neutrons, the reaction rates and the continuous energy distribution of spallation neutrons are measured by the foil activation method and by an organic liquid scintillator, respectively. Numerical calculations are executed by MCNPX with JENDL/HE-2007 and ENDF/B-VI libraries to evaluate the reaction rates of activation foils (bismuth and indium) set at the target and the continuous energy distribution of spallation neutrons set in front of the target. For the reaction rates by the foil activation method, the C/E values between the experiments and the calculations are found around a relative difference of 10%, except for some reactions. For continuous energy distribution by the organic liquid scintillator, the spallation neutrons are observed up to 45 MeV. From these results, the neutron spectrum information on the spallation neutrons generated at the target are attained successfully in injecting 100 MeV protons onto the tungsten target.

Modification of the neutron beam spectrum for neutron radiography at Tehran Research Reactor (TRR)

Recently due to the replacement of the High Enriched Uranium (HEU) fuel with the Low Enriched Uranium (LEU) fuel and the changes in the reactor core configuration at TRR, the existing Neutron Radiography (NR) system was no longer efficient. Thus, it was decided to modify the system in order to increase the neutron flux and to improve the characteristics of the system. The neutron energy spectrum was measured by foil activation method using SAND-II code and calculated by ANISN/PC code. The general trend of the calculated and measured spectra show good similarity. By introducing different sizes of moderator and gamma absorber behind the collimator, the optimum thermal neutron flux impinge the collimator was calculated using ANISN/PC code. The inlet diameter of the collimator was changed from 1.8 to 5 cm in order to increase the neutron flux at the sample position, which should result in an increase of 8 fold in spite of a small increase in the geometrical unsharpness. The new beam characteristics at the sample position are predicted as an average thermal neutron flux of about 10 6 n cm 2 s 1 and a neutron to gamma ratio of about 10 6 n cm 2 mR −1.

Determination of the thermal neutron flux in a fast neutron beam by use of a boron‐coated ionization chamber

The thermal neutron distribution in slow and fast neutron beams is usually determined using the foil activation method. In this work a small magnesium walled ionization chamber, in which the inner surface of the wall has been coated with 10B to increase the sensitivity for thermal neutrons, is used to estimate the thermal neutron component of the beam. After calibration and determination of the directional response in a thermal neutron beam a comparison with foil activation at different depths in water was performed to investigate the reliability of the ionization measurements. The chamber was used in a computer controlled water phantom to measure the depth and lateral distribution of the thermal neutron dose. With this arrangement two-dimensional scans of the thermal neutrons could be performed quickly and with high accuracy.

Neutron field produced by a KAMAN A-711 sealed tube generator

Utilization of a sealed tube neutron generator requires the knowledge of the lateral flux density distribution and the mean energy of primary neutrons near to the target. The ion composition of the incident beam is related to the operating conditions of a generator. These parameters have been determined for a KAMAN A-711 sealed tube machine by the foil activation method using different energy and fluence monitor reactions. An analytical expression is also given for the calculation of the relative flux density for a typical geometry.

Dose Equivalent Measurement for Thermal Neutrons and Gamma Rays with a Small BGO Scintillator Connected to an Optical Fibre

A small dose equivalent counter for thermal neutrons and gamma rays for measuring the dose equivalent for a patient in boron neutron capture therapy is described. The system consisted of a small BGO scintillator partially covered with a Cd chip connected to an optical fibre. The thermal neutron response of the detector wasadjusted to correspond to the ICRP value (conversion factor from thermal neutron fluence to dose equivalent) by changing the size of the Cd chip. The dose equivalent of thermal neutrons and gamma rays was simultaneously measured in free air mixed field. The observed values were compared with the values deduced from the thermal neutron fluence determined by the gold foil activation method and the gamma ray dose by TLDs.

Radiation Damage of Silicon Junction Photodiodes by Evaporation Neutrons from Nuclear Spallation Reactions with 12 GeV Protons

Radiation damage effects of silicon PIN photodiodes were studied by using evaporation neutrons from nuclear spallation reactions of a copper target with 12 GeV protons. Photodiode samples were placed in the backward direction from the copper target and neutron flux was determined by the foil activation method. The average neutron energy was estimated to be about 6 MeV. The damage effects were observed in terms of the increase of leakage current with reverse voltage which is expected to be approximately proportional to radiation dose of neutrons. The present result of the leakage current constant is of the order of 3 to 4 times those measured by using identical photodiodes exposed to reactor neutrons with the average energy of about 2 MeV. The observed energy dependence of the leakage current constant can be qualitatively explained by the NRT model.

Radiation damage of silicon junction detectors by neutron irradiation

We report measurements of radiation damage in silicon PIN detectors resulting from irradiation by reactor neutrons. The fast component of the neutron flux was measured to be 10 13–10 14 cm −2 by a foil activation method. The damage was observed as an increase in the leakage current. The leakage current constant was evaluated to be 6.6 × 10 −17 A/cm. This value is a factor 2 smaller than those of other experiments on neutron damage, and a factor 2 larger than the damage caused by high-energy proton irradiation.

Measurements and a direct-reaction-plus-Hauser-Feshbach analysis of 89Y(p,n)89Zr, 89Y(p,2n)88, and 89Y(p,pn)88Y reactions up to 40 MeV.

We report measurements and analyses of cross sections up to 40 MeV for the reactions $^{89}(\mathit{p}$,n${)}^{89}$Zr, $^{89}(\mathit{p}$,2n${)}^{88}$Zr, and $^{89}(\mathit{p}$,pn${)}^{88}$Y obtained by the foil-activation method. The data are compared with other measurements and analyzed by the Hauser-Feshbach plus one- and two-step direct reactions. The analysis shows that Hauser-Feshbach plus one-step direct reaction is adequate to explain the data. The cross sections are also calculated with phenomenological preequilibrium models (exciton, hybrid, and geometry-dependent hybrid models), using global parameters. However, to get a good fit to the data with these phenomenological models, some adjustments of the global parameters were necessary. A discussion on the differences between the direct-reaction theory and preequilibrium models is given.

Dose measuring system for boron neutron capture therapy

Dose measuring systems for boron neutron capture therapy (BNCT) of brain tumors are presented. The systems are a real-time monitoring system, an integral measuring system and a 10B concentration measuring system. The real-time monitoring with a small PN junction silicon detector made it possible to simultaneously measure the thermal neutron flux and the gamma dose rate in a patient during neutron therapy. Another monitoring of dose equivalents of thermal neutrons and gamma rays was performed with a BGO scintillation detector connected to an optical fiber. The accurate neutron fluence and gamma dose were determined with the integral measurements of the foil activation method and thermoluminescent dosimeters (TLDs) after irradiation. Kerma doses of thermal neutrons and gamma-rays were also measured with the TLD at the same time. Preliminary measurements of 10B concentration in tissue and blood of a patient were carried out by prompt gamma-ray spectroscopy.

Cr52(p,n)52Mng,mandCr52(d,2n)52Mng,mexcitation functions

We have irradiated natural chromium with protons and deuterons to produce $^{52}\mathrm{Mn}^{\mathrm{g}}$,m, primarily from the $^{52}\mathrm{Cr}$(p,n) and $^{52}\mathrm{Cr}$(d,2n) reactions. The excitation functions up to 27 MeV were measured by the foil activation method and were compared with results from a statistical model calculation based on the Hauser-Feshbach formalism with preequilibrium emission. For $^{52}\mathrm{Cr}$(p,n${)}^{52}$${\mathrm{Mn}}^{\mathrm{g}}$,m, both the total cross section, \ensuremath{\sigma}(g+m), and the isomer ratio, \ensuremath{\sigma}(m)/\ensuremath{\sigma}(g), are reproduced to within 10% by the calculations. By contrast, for the $^{52}\mathrm{Cr}$(d,2n${)}^{52}$${\mathrm{Mn}}^{\mathrm{g}}$,m reaction, we find that calculations overestimate \ensuremath{\sigma}(g+m) by \ensuremath{\sim}50% and underestimate \ensuremath{\sigma}(m)/\ensuremath{\sigma}(g) by nearly one-half. We find that arbitrary adjustments of level density or preequilibrium parameters can yield approximate agreement for \ensuremath{\sigma}(g+m) but do not improve the agreement for \ensuremath{\sigma}(m)/\ensuremath{\sigma}(g). A part of this discrepancy may be due to the fact that the preequilibrium calculations do not include angular momentum. However, to fully understand this discrepancy, deuteron breakup effects in the entrance channel must be taken into account. A companion paper explores this idea.

Measurement of the cross sections for the 232Th(n,2n)231Th reaction in the 6.745 to 10.450 MeV energy range.

A part of the excitation function of the $^{232}\mathrm{Th}$(n,2n) reaction was determined by the foil activation method for neutron energies of 6.745, 6.938, 7.190, 7.448, 7.697, 7.944, 8.457, 8.975, 9.445, 9.934, and 10.450 MeV. Neutron flux density was measured using reactions $^{238}\mathrm{U}$(n,f), $^{56}\mathrm{Fe}$(n,p), and $^{27}\mathrm{Al}$(n,\ensuremath{\alpha}). Activity of the $^{231}\mathrm{Th}$ residual nuclei was determined by their low energy \ensuremath{\gamma} transitions detected with a high purity Ge spectrometer. Results of this experiment are compared with renormalized literature data.

測定器校正用熱中性子場の製作と特性試験

A graphite pile was constructed for routine calibration of radiation protection instruments. An isotropic thermal neutron irradiation field can be set up at the cavity in the pile. Also a collimated thermal neutron beam irradiation field can be set up at the outside facing the rectangular hole of pile. The sizes of irradiation fields are 200×276×170mm3 and 160×160mm2 respectively. Thermal neutron fluences are measured by the gold foil activation method. The over all error of fluence measurements is less than±2%. Intercomparison between ETL and JAERI on thermal neutron fluence rate measurements was done and the results have agreed within experimental errors. Distribution of fluence rate in the irradiation field are uniform with ±3%.

Measurements of neutron spectra and fluxes at spallation-neutron sources and their application to radiation effects research

We have measured the neutron spectra, fluxes, and flux distributions produced by nuclear spaliation resulting from 478 MeV proton bombardment of tantalum and depleted uranium targets surrounded by a thick lead neutron reflector. The configuration was chosen to simulate a radiation effects facility at a spallation-neutron source. The method of multiple foil activation with spectrum unfolding by the STAYSL computer code was used to measure the neutron spectra. The experimental results are compared in detail with the results of computer calculations on the same configuration of targets and reflector. The neutron production and transport codes HETC and VIM were employed in these calculations. The neutron spectra, and characteristic parameters for studies of radiation effects, are compared with those of fission-neutron sources. Based on the above measurements, the predicted performance and capabilities of the Radiation Effects Facility at the Intense Pulses Neutron Source at Argonne National Laboratory are discussed.

Measurement of local power peaking factors in heavy-water moderated plutonium lattices.

Local power peaking factors (LPF) in the heavy-water moderated plutonium lattices were measured by a new method of γ-scanning of fuel pins using the calculated power correction factor of which accuracy was evaluated with the aid of foil activation method. Accuracy in measurement of the LPF was evaluated to be within 1%. By this measurement, behaviors of the LPF have been made clear concerning the differences in fuel materials, coolant materials and arrangement in fuel enrichments. Depression of the thermal power in the fuel cluster makes LPF in the plutonium fuel lattice larger than in the uranium lattice. This tendency is more remarkable in air coolant lattice than in H2O coolant lattice. The value of LPF for the plutonium fuel cluster of different enrichments is smaller than that of a uniformly enriched fuel cluster. The reduction of LPF is smaller in H2O coolant than in air coolant lattice. The values of LPF by WIMS-D code based on the transport theory and by METHUSELAH-II code based on the diffusion theory are in agreement with the measured ones, within 1.5 and 2.4% respectively.

International Comparison of Flux Density Measurements for Monoenergetic Fast Neutrons

An international flux density comparison of monoenergetic fast neutrons has been organized by the Bureau International des Poids et Mesures. Nine laboratories carried out measurements which extended from December 1973 to February 1978. The five neutron energies 250 keV, 565 keV, 2.2 MeV, 2.5 MeV and 14.8 MeV were selected. A polyethylene sphere with a small BF3 counter at the center served as a transfer instrument at all energies except for 14.8 MeV. A 3He proportional counter was used at the two lower energies as the second transfer instrument. At 14.8 MeV a fission chamber (238U) and the iron foil activation method based on the 56Fe (n,p) 56Mn reaction were used. The experimental conditions, the neutron flux density determinations and the transfer instrument measurements are described. The results of the intercomparison are summarized in tables and graphs.