Radionuclide Neutron Sources Research Articles

Estimation of Valuable Metal Content in Coal Combustion Products of the Russian Far East by INAA with a Radionuclide Neutron Source

Estimation of Valuable Metal Content in Coal Combustion Products of the Russian Far East by INAA with a Radionuclide Neutron Source

Development and testing of a fission-like neutron field for dosimetry and instrument calibrations based on deuterium-tritium generator

For nearly five decades, 252Cf sources have been used for testing and calibration over a wide range of neutron detection devices used for nuclear safety and radiation protection. However, the vastly increased cost of 252Cf sources, its short half-life, and concerns about future shortage have prompted nuclear facilities and radiological calibration laboratories that rely on high-intensity neutron sources to seek alternatives. There appear to be no other radionuclide neutron sources that would match “like-to-like” 252Cf performance characteristics, although “fit for purpose” substitutions could be possible. Alternatives of producing a fission-like neutron spectrum based on commercially available deuterium-tritium (D-T) generators, and without any fissionable materials or beryllium, have been explored. This paper describes efforts of designing the spectrum shaping assembly, its construction and installation. The produced neutron field was evaluated both from the perspective of the neutron fluence spectrum and the resulting dosimetric outcome. Finally, it was tested for practical application by examining the response of various health physics instruments and personal monitoring devices in comparison to the response of those devices when exposed to 252Cf. The response of survey-type instruments tested within the prototype, surrogate fission, reference field are acceptably close to their response in the targeted 252Cf field.

Determination of uranium-238 in solid materials of various compositions by instrumental neutron activation analysis with a radionuclide neutron source based on Cf-252

Determination of uranium-238 in solid materials of various compositions by instrumental neutron activation analysis with a radionuclide neutron source based on Cf-252

Activity determination of 56Mn using extended TDCR-Čerenkov method

The accuracy of activity determination for activated nuclide 56Mn is the key to the manganese bath method applying to the characterization of radionuclide neutron source. As an alternative to the 4π(Č)-γ method, TDCR-Čerenkov method could also be applied to the measurement of 56Mn in the manganese bath device, if the existing calculation model is extended. There are two difficulties when the existing TDCR-Čerenkov method is applied to the activity determination of 56Mn. One is that the efficiency computation of gamma transitions, and the other is the interference contributed by Čerenkov photons emitted in the photomultiplier windows induced by Compton scattering. In this study, the above two difficulties are solved by extending the calculation model. For efficiency computation, the decay scheme of 56Mn is taken into account in the calculation of efficiency. Among them, the efficiency of gamma transition is calculated from the simulated secondary electronic spectra. In addition, Čerenkov photons emitted in photomultiplier windows are corrected by additional light proof experiment and improved calculation model. The results derived from this extended method are in good agreement with other standardization technique.

The National Research Council of Canada manganese bath system.

The manganese salt bath is considered a primary standard for determining the absolute emission rate of radionuclide neutron sources. The National Research Council of Canada has recently revived its manganese salt bath and a full description of the system is given here. The physical characteristics of the bath, as well as the methods for determining the efficiency of the bath system and the induced activity in the bath, are described. An in-depth analysis of the fraction of neutrons captured in the manganese and the correction factor for neutron losses is also provided. Finally, the results of emission rate measurements of four different sources, complete with an uncertainty budget, are given. The emission rates of three americium-beryllium neutron sources and one californium-252 neutron source were found to agree with the known values, within a standard uncertainty of 1.7%.

Characterization of Thermoluminescent Dosimeters for Neutron Dosimetry at High Altitudes.

Neutrons constitute a significant component of the secondary cosmic rays and are one of the most important contributors to natural cosmic ray radiation background dose. The study of the cosmic ray neutrons’ contribution to the dose equivalent received by humans is an interesting and challenging task for the scientific community. In addition, international regulations demand assessing the biological risk due to radiation exposure for both workers and the general population. Because the dose rate due to cosmic radiation increases significantly with altitude, the objective of this work was to characterize the thermoluminescent dosimeter (TLDs) from the perspective of exposing them at high altitudes for longtime neutron dose monitoring. The pair of TLD-700 and TLD-600 is amply used to obtain the information on gamma and neutron dose in mixed neutron-gamma fields due to the present difference in 6Li isotope concentration. A thermoluminescence dosimeter system based on pair of TLD-600/700 was characterized to enable it for neutron dosimetry in the thermal energy range. The system was calibrated in terms of neutron ambient dose equivalent in an experimental setup using a 241Am-B radionuclide neutron source coated by a moderator material, polyethylene, creating a thermalized neutron field. Afterward, the pair of TLD-600/700 was exposed at the CERN-EU High-Energy Reference Field (CERF) facility in Geneva, which delivers a neutron field with a spectrum similar to that of secondary cosmic rays. The dosimetric system provided a dose value comparable with the calculated one demonstrating a good performance for neutron dosimetry.

Determination of a Pu-Be source neutron spectrum at Czech Metrology Institute

Contrary to the radionuclide neutron sources of 252Cf and 241Am-Be type, which are most widely used for the dosimetric instrumentation calibrations, the spectral source strength and spectrum averaged fluence to dose equivalent conversion coefficients of Pu-Be source are not recommended by an international standard. This work describes the determination of those parameters for the Pu-Be source used at Czech Metrology Institute (CMI) by means of a Bonner spectrometer in combination with neutron emission rate measurement in a manganese bath.

Photoneutron sources spectra calculation

The issues of increasing the accuracy of determining the average energy and the type of energy distribution (shape of the spectrum) of neutrons from Sb-Be- and Ra-Be-sources are considered. The indicated photoneutron sources are used to calibrate the energy scales of neutron spectrometers and to determine the energy dependence of the sensitivity of radiometers and neutron dosimeters. A method is proposed for calculating the energy spectra of radionuclide neutron sources of the Sb-Be (γ, n) and Ra-Be (γ, n) types, and the calculation results are presented. The dependence of the spectrum of the source on its design is investigated. The correctness of the results of calculating the energy spectra of radionuclide sources has been confirmed experimentally. The data obtained will be used to determine the energy dependence of the sensitivity of neutron spectrometers, radiometers and dosimeters.

Determination of uranium in solutions by the neutron activation analysis method with 252Cf radionuclide neutron source

A determination option for 238U in solutions by the neutron activation analysis was investigated. A radionuclide treatment based on 252Cf was used as a neutron source. The limit for the determination of uranium using thermal neutrons of 0.1 mgL?1 was determined within the research framework. With the help of the preliminary concentration of the sample via the evaporation method, it is possible to further reduce the detection limit by 2-5 times. It is defined that the optimal time for measuring the induced activity of the sample when working with a short-life isotope of 239U, is 30 minutes. The influence of the effects of scattering and self-absorption of gamma radiation of the uranium line 74.6 keV on the results of the neutron activation analysis was determined. The studies conducted made it possible to determine uranium by the neutron activation analysis method in solutions of various compositions, in a wide range of concentrations.

An approach to determination of dosimetric characteristics of radionuclide neutron sources with specific constants and effective quality factors

It is desirable to examine the characteristics of radiation and dosimetric of radionuclide neutron sources. In this work, the ambient dose equivalent (H*(10)) was calculated based on the conversion coefficient from different reports with the simulation of various types of radionuclide neutron sources by Monte Carlo code MCNPX2.6. The ambient dose equivalent was obtained for neutron and gamma emitted from these sources, separately. Then, each dose was parameterized as a function of emission rate and distance from the source, to estimate the dose for any strength and each distance. Furthermore, using simulation of the ICRU spherical phantom irradiated by radionuclide neutron sources, the neutron and photon absorbed doses were calculated at a depth of 10 mm in this phantom, and then they were compared with the ambient dose equivalent for neutrons and photons. From this comparison, the effective quality factor of these sources was estimated.

Study of neutron energy and directional distribution at the Beloyarsk NPP selected workplaces

Energy and directional distribution of neutrons at the Beloyarsk NPP workplaces is a subject of this study. Measurements of H∗(10) rate and neutron energy distribution were taken at 8 workplaces, which can be divided into three categories: work with spent or fresh nuclear fuel, work with radionuclide neutron sources, work at the rooms adjusted to reactors. The Hp(10) measurements were performed only at 6 out of 8 locations, due to the fact that long term placing of an effective neutron moderator in fresh nuclear fuel storage facility is forbidden. As a result of the research energy and direction distribution of the neutron fields at 8 locations of the Beloyarsk NPP workplaces was obtained. To estimate the accuracy of the H∗(10) rate and Hp (10) measurements the reference values of dose equivalents were calculated using energy and directional distribution. To take into account the difference between the reference values and the measured results site-specific correction factors were calculated.

Assessing the suitability of three proxy sources for the development of detectors of special nuclear materials

Numerous techniques and equipment have been developed to provide a capability for the detection of special nuclear materials (SNM), but due to the necessary security measures surrounding these materials alternate, or proxy, neutron sources are often utilised in their stead. In this paper we report the neutron and gamma pulse shape discrimination response of plastic scintillator to mixed neutron/gamma beams produced from two radionuclide neutron sources, and also from an SNM source of weapons-grade plutonium. We discuss the suitability of using radionuclide sources, with appropriate shielding configurations as proxy sources for SNM.A 3σnth-γ discrimination level has been achieved for an SNM source at a low-level energy threshold of ∼220 keVee when a shielding configuration of 5 cm of lead was implemented. Varying amounts of lead and high-density polyethylene (HDPE) shielding were also investigated with the 3σ limit being reached by ∼240 keVee.This work shows that an AmBe neutron source serves as an appropriate SNM proxy achieving a comparable value for figure of merit above ∼1 MeVee. For energies below 1 MeVee down to ∼100 keVee a closer approximation of the expected FoM for SNM can be attained when using 252Cf as a proxy source or by utilising an ‘enhanced’ AmBe source with the addition of a further low energy γ ray source.

Specifying the flux and dose-equivalent buildup factors for infinite slabs irradiated by radionuclide neutron sources

In this work, flux and neutron dose-equivalent buildup parameters are calculated for six radionuclide point-like neutron sources having broad energy spectrum which irradiate infinite slab-like common shielding materials of beryllium, concrete, iron, graphite, water, and lead, employing the MCNPX code. The description of the buildup factor is made in a straightforward way which is analogous to that of gamma. The parameters are obtained for thicknesses of shield from 0.5 to 10 mean free paths (mfp). The achieved dose-equivalent buildup parameters are parameterized by polynomial expressions. Using this parameterization, one can determine these factors for the desired thickness of shield material and each neutron source.

Production of a carrier-free standard 56Mn source for the NRC manganese salt bath

The National Research Council (NRC) of Canada's primary method for emission rate for radionuclide neutron sources utilizes a manganese salt bath which was last calibrated in the 1960s. At that time, an NRC RaBe neutron source was used to irradiate a solution of calcium permanganate to take advantage of the Szilard-Chalmers effect in producing the bulk 56Mn material for standardization and calibration of the bath. When attempting to repeat this exercise, a small amount (~100 kBq) was produced. This amount was sufficient for the standardization process but did not yield enough material to calibrate the bath to a sufficient level of precision.Improvements upon the previous separation scheme adopted at NRC for the separation of the 56Mn from the bulk irradiated material included the rinsing of the 56Mn dioxide precipitate using a mixture of sulfuric acid and hydrogen peroxide. While these improvements made in the separation chemistry improved the yield of 56Mn extraction from 60% to above 95% the maximum amount of activity was still quite low. Hence in March of 2018, the SLOWPOKE-2 Facility at the Royal Military College in Kingston, ON, was used to irradiate three vials of KMnO4 in solution. An estimated 2 GBq was produced and sent to NRC, from which the extraction procedure recovered essentially all of the available 56Mn.The 56Mn was standardized using the 4πβ−γ anti-coincidence counting system and confirmed using the CIEMAT/NIST primary method. The resulting bulk material was certified with an uncertainty of 0.8% (k = 2). Minor quantities of 65Zn, 69mZn and 42K were unexpectedly observed but were in minute quantities so as not to affect the results of the standardization or calibration. The standardized 56Mn artifact was used to calibrate the Secondary Standard Ionizing Radiation Chamber System (SSIRCS) for a more rapid deployment of the calibrant in the future.

Radionuclide neutron source trajectories in the closed nuclear fuel cycle

Abstract The highest efficiency in the usage of nuclear energy resources can be implemented in fast breeder reactors of generation IV. It is achieved thanks to the ability of consuming minor actinides (MAs) in energy production. One of the options to use this benefit is full recycling of MAs to close the nuclear fuel cycle. Monte Carlo burn up (MCB), an integrated burn-up calculation code, deals with the complexity of the burn-up process which is applied to the European Lead-cooled Fast Reactor (ELFR). MCB uses continuous energy representation of cross section and spatial effects of full core reactor model; however, it automatically calculates nuclide production in all possible reactions or decay channels. Multi-recycling of MAs can cause an intensified build-up of curium, berkelium and californium. Some of their isotopes are strong neutron emitters from spontaneous fission, which hinders fuel recycling. The implementation of a novel methodology for trajectory period folding allows us to trace the life cycle of crucial MAs from the beginning of the reactor life towards the state of adiabatic equilibrium. The result of the analysis performed is presented, showing the sources of strong contribution to the neutron production rate. The parametric sensitivity analysis method for selected nuclide reactions is applied, revealing sensitivity of transmutation chains for the production of neutron emitter isotopes.

Study of different solutes for determination of neutron source strength based on the water bath

Time required for activation to saturation and background measurement is considered a limitation of strength determination of radionuclide neutron sources using manganese bath system (MBS). The objective of this research was to evaluate the other solutes based on water bath for presentation of the suitable replacement with MBS. With the aid Monte Carlo simulation, for three neutron sources, having different neutron spectra, immersed in six aqueous solutions, i.e., Na2SO4, VOSO4, MnSO4, Rh2(SO4)3, In2(SO4)3, I2O5, the correction factors in all nuclei of solutions for neutron losses with different process were obtained. The calculations results indicate that the Rh2(SO4)3 and VOSO4 are best options for replacing with MnSO4.

PHOTON SPECTRA IN NPL STANDARD RADIONUCLIDE NEUTRON FIELDS.

A HPGe detector has been used to measure the photon spectra from the majority of radionuclide neutron sources in use at NPL (252Cf, 241Am-Be, 241Am-Li, 241Am-B). The HPGe was characterised then modelled to produce a response matrix. The measured pulse height spectra were then unfolded to produce photon fluence spectra. Changes in the photon spectrum with time from a 252Cf source are evident. Spectra from a 2-year-old and 42-year-old 252Cf source are presented showing the change from a continuum to peaks from long-lived isotopes of Cf. Other radionuclide neutron source spectra are also presented and discussed. The new spectra were used to improve the photon to neutron dose equivalent ratios from some earlier work at NPL with GM tubes and EPDs.

The thermal neutron facility HOTNES: theoretical design

HOTNES (HOmogeneous Thermal NEutron Source) is a thermal neutron irradiation facility with extended and very uniform irradiation area. A 241Am-B radionuclide neutron source with nominal strenght 3.5×106 s−1 is located on bottom of a large cylindrical cavity (30cm diameter, 70cm in height) delimited by polyethylene walls. The upper part of this volume (30cm diameter, 40cm in height) is used to irradiate samples. A polyethylene cylinder, acting as shadowing object, prevents fast neutrons to directly reach the irradiation volume. Indeed neutrons can only reach the irradiation volume after multiple scattering with the cavity walls. The facility was designed trough extensive calculations with MCNPX. Irradiation planes are disks with 30cm diameter, centred on the cavity axis, and parallel to the cavity bottom. The value of thermal fluence in a given irradiation plane is as uniform as 1–2%. The value of thermal fluence rate simply depends on the height from the cavity bottom. Values of thermal fluence rate in the range 700–1000cm−2s−1 are available, depending on the irradiation plane chosen. The fraction of thermal neutrons is in the order of 90%, also depending on the irradiation plane. The angular distribution of thermal neutrons is roughly isotropic. Taking advantage of the HOTNES design, even large devices can be uniformly irradiated.This work presents HOTNES's design and describes the neutron field in the irradiation volume in terms of spatial, energy and direction distributions.

Estimating the effective system dead time parameter for correlated neutron counting

Neutron time correlation analysis is one of the main technical nuclear safeguards techniques used to verify declarations of, or to independently assay, special nuclear materials. Quantitative information is generally extracted from the neutron-event pulse train, collected from moderated assemblies of 3He proportional counters, in the form of correlated count rates that are derived from event-triggered coincidence gates. These count rates, most commonly referred to as singles, doubles and triples rates etc., when extracted using shift-register autocorrelation logic, are related to the reduced factorial moments of the time correlated clusters of neutrons emerging from the measurement items. Correcting these various rates for dead time losses has received considerable attention recently. The dead time losses for the higher moments in particular, and especially for large mass (high rate and highly multiplying) items, can be significant. Consequently, even in thoughtfully designed systems, accurate dead time treatments are needed if biased mass determinations are to be avoided. In support of this effort, in this paper we discuss a new approach to experimentally estimate the effective system dead time of neutron coincidence counting systems. It involves counting a random neutron source (e.g. AmLi is a good approximation to a source without correlated emission) and relating the second and higher moments of the neutron number distribution recorded in random triggered interrogation coincidence gates to the effective value of dead time parameter. We develop the theoretical basis of the method and apply it to the Oak Ridge Large Volume Active Well Coincidence Counter using sealed AmLi radionuclide neutron sources and standard multiplicity shift register electronics. The method is simple to apply compared to the predominant present approach which involves using a set of 252Cf sources of wide emission rate, it gives excellent precision in a conveniently short time, and it yields consistent results as a function of the order of the moment used to extract the dead time parameter. This latter observation is reassuring in that it suggests the assumptions underpinning the theoretical analysis are fit for practical application purposes. However, we found that the effective dead time parameter obtained is not constant, as might be expected for a parameter that in the dead time model is characteristic of the detector system, but rather, varies systematically with gate width.

INFLUENCE OF SCATTERED NEUTRON RADIATION ON METROLOGICAL CHARACTERISTICS OF АТ140 NEUTRON CALIBRATION FACILITY

Today facilities with collimated radiation field are widely used as reference in metrological support of devices for neutron radiation measurement. Neutron fields formed by radionuclide neutron sources. The aim of this research was to study characteristics of experimentally realized neutron fields geometries on АТ140 Neutron Calibration Facility using Monte Carlo method. For calibration, we put a device into neutron field with known flux density or ambient equivalent dose rate. We can form neutron beam from radionuclide fast-neutron source in different geometries. In containercollimator of АТ140 Neutron Calibration Facility we can install special inserts to gather fast-neutron geometry or thermal-neutron geometry. We need to consider neutron scattering from air and room’s walls. We can conduct measurements of neutron field characteristics in several points and get the other using Monte Carlo method. Thermal neutron collimator forms a beam from radionuclide source with a significant amount of neutrons with thermal energies. From found relationship between full neutron flux and distance to neutron source we see that inverse square law is violated. Scattered radiation contribution into total flux increases when we are moving away from neutron source and significantly influences neutron fields characteristics. While source is exposed in shadow-cone geometry neutron specter has pronounced thermal component from wall scattering. In this work, we examined main geometry types used to acquire reference neutron radiation using radionuclide sources. We developed Monte Carlo model for 238 Pu-Be neutron source and АТ140 Neutron Calibration Facility’s container-collimator. We have shown the most significant neutron energy distribution factor to be scattered radiation from room’s walls. It leads to significant changes of neutron radiation specter at a distance from the source. When planning location, and installing the facility we should consider radiation quality requirements.