Investigation of 252Cf sources with a Ge gamma-ray detector.

This study investigated the feasibility of combining high-resolution gamma-ray spectroscopy with the simulation capabilities of the Nucleonica Nuclear Science Portal with the aim to determine the properties of Cf sources. In this contribution, we present the results for a 20-month-old and a 49-year-old Cf source. In particular, the question arises whether the neutron emission rate can be determined using gamma-ray spectroscopy.

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%.

ESTIMATION OF THE ANISOTROPY EMISSION OF LPN/CIEMAT NEUTRON SOURCES: EFFECT OF HEAVY CAPSULE HOLDERS.

In order to estimate the anisotropy emission of 241Am-Be and 252Cf neutron sources from the Spanish Neutron Standards Laboratory (LPN/CIEMAT) detailed models of sources capsules and capsule holders were designed with the MCNPX code. Simulations of the sources inside the capsules without the capsule holders were done to validate the MC model by comparison with experimental results provided by other authors. After that, the capsule holders were incorporated to the simulation. In general, a good agreement has been found between measurements and our calculations. Results show the additional encapsulations have significant influence on anisotropy factors, energy spectra and dose rates.

Study on the Relationship between Seed Absorbed Dose and Seed Composition of 252Cf Neutron Source Irradiated Bean Seed.

This study aims to further identify the biological effects of neutron-irradiated plants and provides insights into the mutation breeding of such plants. In this study, the neutron irradiation device designed by our institute was used to analyze the relationship between the seed components in different legume crops and their neutron absorption dose rate, fission gamma absorption dose rate, and induced gamma absorption dose rate. The results show that the effect sizes of the components on the neutron absorbed dose rate are as follows: ash > fat > moisture > carbohydrate > protein. The effect sizes of the components on the absorbed dose rate of fission gamma are as follows: ash > moisture > fat > carbohydrate > protein. There is a positive correlation between fission gamma absorbed dose rate and the weight of ash, water and fat, while a negative correlation with carbohydrate and protein. However, the linear relationship between each component and the absorbed dose rate of induced gamma is not significant, this needs to be identified by further researches. Based on the results of the present study, we conclude that the neutron absorbed dose can be calculated without taking into account the fat composition of bean crop seeds (except for soybean seeds) in the process of mutation breeding induced by radiation. In special cases where the accuracy requirement of the dose rate is not high, it is possible to use protein instead of legume crop seeds for neutron absorption dose calculations.

THE DETERMINATION OF NEUTRON FLUENCE TO ABSORBED DOSE CONVERSION COEFFICIENTS AND RELATIVE BIOLOGICAL EFFECT BASED ON MICRODOSIMETRY MEASUREMENTS.

A tissue-equivalent proportional counter (TEPC) is a reference detector to measure microdosimetric quantities. A conventional spherical TEPC and a novel TEPC based on a ceramic thick gas electron multiplier (THGEM) foil were developed to carry out microdosimetric measurements of lineal energy spectra in monoenergetic and 252Cf/241Am-Be neutron radiation fields, and the absorbed dose values had been derived. In order to go further in radiobiology and therapy, the fluence to absorbed dose conversion coefficients in neutron fields were also determined. According to the dose distribution in lineal energy, the neutron relative biological effect (RBE) values were also calculated using an empirical procedure applying biological weighting functions.

EXPERIMENTAL AND COMPUTATIONAL EVALUATION OF EFFECTIVE CENTRE FROM A LONG COUNTER AT NEUTRON METROLOGY LABORATORY IN BRAZIL.

A long counter detector was manufactured by the Institute of Advanced Studies (IEAV) and was characterised in the neutron low scattering room at Brazilian National Ionising Radiation Metrology Laboratory (LNMRI/IRD) to deploy a secondary Standard for neutron fluence. The effective centre was measured experimentally with 252Cf+D2O, 252Cf, 241AmBe and 238PuBe neutron sources, having average energies from 0.55 to 4.16 MeV. The experimental arrangement and detector construction were carefully reproduced in Monte Carlo simulations, and the computational results were found to be in good agreement with those from experiment.

ALGORITHM BASED ON ARTIFICIAL BEE COLONY FOR UNFOLDING OF NEUTRON SPECTRA OBTAINED WITH BONNER SPHERES.

Occupational neutron fields usually have energies from the thermal range to some MeV and the characterization of the spectra is essential for estimation of the radioprotection quantities. Thus, the spectrum must be unfolded based on a limited number of measurements. This study implemented an algorithm based on the bee colonies behavior, named Artificial Bee Colony (ABC), where the intelligent behavior of the bees in search of food is reproduced to perform the unfolding of neutron spectra. The experimental measurements used Bonner spheres and 6LiI (Eu) detector, with irradiations using a thermal neutron flux and three reference fields: 241Am-Be, 252Cf and 252Cf (D2O). The ABC obtained good estimation of the expected spectrum even without previous information and its results were closer to expected spectra than those obtained by the SPUNIT algorithm.

REFERENCE NEUTRON FIELDS IN CALIBRATION LABORATORY; SIMPLE DOSIMETRIC PARAMETERS AND THEIR CHANGES IN TIME.

The calibration laboratory revises periodically the dosimetric properties of its calibration fields. For reference neutron fields based on radionuclide sources, it is due to presence of isotopic impurities in the source ingredients. Consequently, in long term, the admixtures and their decay products become significant. Neglecting the impact of such effect leads to inconsistencies between the neutron emission rate observed during the measurements and the emission rate derived from the decay curve of the main isotope. In the Radiation Protection Measurements Laboratory at the National Centre for Nuclear Research the neutron fields of the bare sources of 252Cf, 241AmBe and 239PuBe have been examined for nearly 30 years, regularly, in fixed geometry. Additionally, at the particular point of a calibration bench, determination of neutron fluence rate, ambient dose equivalent rate and its scattered component, total neutron and gamma dose rate, gamma to total dose ratio, radiation quality factor and total tissue kerma are occasionally determined. In this article, we would like to present recently achieved results and their comparison with the measurement data formerly presented. The growth of the neutron emission of 239PuBe source, as well as discrepancies between the decay curves of relevant isotopes and the emission rate of 252Cf and 241AmBe will be presented.

NEUTRON FIELD CHARACTERISTICS AT RADIATION METROLOGY LABORATORY OF STUK.

Radiation metrology laboratory at Radiation and Nuclear Safety Authority (STUK) is Finnish national laboratory for ionizing radiation. In order to establish personal dose equivalent available for neutron calibration service, a project was started to evaluate the neutron reference fields in the present facility. The irradiation room conditions were characterized in order to establish reference conditions for personal dose equivalent. To verify the validity of the calculations, ambient dose equivalent rates and room return were measured and calculated for 241AmBe and 252Cf sources. First results of Monte Carlo calculations and measurements are presented in this article.

DETERMINATION OF THE RESPONSE TO THE ATMOSPHERIC COSMIC RADIATION OF A NEUTRON DOSIMETER ASSISTED BY MONTE CARLO SIMULATION.

A TLD-based dosimeter of polyethylene-lead-polyethylene, was developed and characterized with Monte Carlo simulations, using the MCNPX code. This passive system for the determination of the ambient dose equivalent (H*(10)) for neutrons over a wide energy range can be used for the dosimetry of neutrons from atmospheric cosmic radiation, on the ground, and onboard aircraft. A method assisted by Monte Carlo simulations that improves the calibration of fast neutron dosimeters based on moderation and thermalization of the incident fast flux and the measurement of the thermal flux by a sensor, which respond mainly to thermal neutrons, is presented in this work. The H*(10) energy response of this dosimeter was obtained from simulations for monoenergetic neutrons from 10-10 to 104 MeV. The validation of the modeling was done with irradiations for ISO standard neutron fields of 241Am-Be, 252Cf and 252Cf(D2O) at Instituto de Radioproteção e Dosimetria (IRD, Brazil) and at CERN-EU high-energy reference field (CERF).

ORGANIC SCINTILLATOR FOR REAL-TIME NEUTRON DOSIMETRY.

We developed a radiation detector based on an organic scintillator for spectrometry and dosimetry of out-of-field secondary neutrons from clinical proton beams. The detector consists of an EJ-299-34 crystalline organic scintillator, coupled by fiber optic cable to a silicon photomultiplier (SiPM). Proof of concept measurements were taken with 137Cs and 252Cf, and corresponding simulations were performed in MCNPX-PoliMi. Despite its small size, the detector is able to discriminate between neutron and gamma-rays via pulse shape discrimination. We simulated the response function of the detector to monoenergetic neutrons in the 100 keV-0 MeV range using MCNPX-PoliMi. The measured unfolded 252Cf neutron spectrum is in good agreement with the theoretical Watt fission spectrum. We determined the ambient dose equivalent by folding the spectrum with the fluence-to-ambient dose conversion coefficient, with a 1.4% deviation from theory. Some preliminary proton beam experiments were preformed at the Bronowice Cyclotron Center patient treatment facility using a clinically relevant proton pencil beam for brain tumor and craino-spinal treatment directed at a child phantom.

EVALUATION OF NEUTRON SCATTERING CORRECTION USING THE SEMI-EMPIRICAL METHOD AND THE SHADOW-CONE METHOD FOR THE NEUTRON FIELD OF THE KOREA ATOMIC ENERGY RESEARCH INSTITUTE.

When neutron survey metres are calibrated in neutron fields, the results for room- and air-scattered neutrons vary according to the distance from the source and the size, shape and construction of the neutron calibration room. ISO 8529-2 recommends four approaches for correcting these effects: the shadow-cone method, semi-empirical method, generalised fit method and reduced-fitting method. In this study, neutron scattering effects are evaluated and compared using the shadow-cone and semi-empirical methods for the neutron field of the Korea Atomic Energy Research Institute (KAERI). The neutron field is constructed using a 252Cf neutron source positioned in the centre of the neutron calibration room. To compare the neutron scattering effects using the two correction methods, measurements and simulations are performed using respectively KAERI's Bonner sphere spectrometer (BBS) and Monte Carlo N-Particle code at twenty different positions. Neutron spectra are measured by a europium-activated lithium iodide [6LiI(Eu)] scintillator in combination with the BBS. The calibration factors obtained using each methods show good agreement within 1.1%.

DEVELOPMENT, CHARACTERIZATION AND TESTING OF A SIMPLIFIED BONNER SPHERE SYSTEM FOR RAPID NEUTRON FIELD CHARACTERIZATION.

Personal neutron dosimetry requires workplace specific correction factors from detailed neutron field characterization. A Bonner sphere system (BSS) with a 3He proportional counter and a 3 inch an a 9 inch sphere was developed at SCK·CEN for rapid neutron field characterization. The BSS energy and angular responses were simulated with MCNPX and validated by irradiations with a thermal neutron beam at the Belgian Reactor 1 (BR1) and 252Cf and Am-Be sources at the SCK·CEN Laboratory for Nuclear Calibration (LNK). The BSS unfolding performance was tested using the FRUIT unfolding code for measurements of the scattered neutron field at the LNK. The unfolded spectra were in good agreement with MCNPX simulations.

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.

DETERMINATION OF NEUTRON EFFECTIVE DOSES IN WHOLE BODY POINT SOURCE EXPOSURES.

The objective of this work is to obtain fluence to effective dose conversion coefficients for neutron point sources, using the GEANT4 toolkit. These calculations aim to investigate the aspects of neutron transport in the human body through Monte Carlo simulation using the International Commission on Radiological Protection (ICRP) voxel phantoms, described in its publication 110. A benchmarking of the code was made for the case of monoenergetic plane parallel neutron beam in the antero-posterior (AP) irradiation geometry and organ absorbed dose conversion coefficients were compared with those found in the ICRP publication 116. The results showed good agreement with ICRP results in the studied energy range. Conversion coefficients were presented for specific conditions with 241Am-Be and 252Cf point neutron sources 1 m away from the phantom in the AP geometry.

INVESTIGATION OF THE EXTENDED RANGE REM-COUNTER SMARTREM-LINUS IN REFERENCE AND WORKPLACE FIELDS EXPECTED AROUND HIGH-ENERGY ACCELERATORS.

Radiation survey instrumentation is adequate for the use around high-energy accelerators if capable to measure the dose arising from neutrons with energies ranging from thermal up to a few gigaelectronvolts. The SmartREM-LINUS is a commercial extended range rem-counter, consisting of a central (3)He-proportional counter surrounded by a spherical moderator made of borated polyethylene with an internal shield made of lead. The dose rate indicated by the SmartREM-LINUS was investigated for two different irradiation conditions. The linearity and the angular dependence of the indicated dose rate were investigated using reference neutron fields produced by (241)AmBe and (252)Cf. Additional measurements were performed in two different workplace fields with a component of neutrons with energies >20 MeV, namely the CERN-EU high-energy reference field and near the beam dump of the SwissFEL injector test facility. The measured dose rates were compared to a commercial rem-counter (WENDI2) and the results of Monte Carlo simulations.

DEVELOPMENT OF A RAPID PROCEDURE TO ANALYSE Pu, Am AND 90Sr IN EMERGENCY URINE BIOASSAY IN CIEMAT BIOELIMINATION LABORATORY: METHOD VALIDATION BY EMERGENCY BIOASSAY INTERCOMPARISON EXERCISES.

After a radiological or nuclear incident, it is necessary to give a prompt response and to know the number of persons exposed to internal contamination, to evaluate the contamination levels in each person and even and to identify the radionuclides involved. In vitro laboratories routine monitoring measurements employed to quantify (90)Sr and actinides in urine require radiochemical separation and long counting time, which implies a minimum of 1 or 2 weeks to obtain the results, respectively. In this work, rapid radiochemical separation method applied directly to urine samples is presented. It is based on minimal sample preparation, without co-precipitation phase, using extraction resin columns and vacuum box technology. Pu isotopes and (241)Am are isolated, electrodeposited and measured by alpha spectrometry, whereas (90)Sr is measured by liquid scintillation counting. Finally, results of the participation in European Radiation Dosimetry Group intercomparison on Emergency Bioassay exercise and Bundesamt für Strahlenschutz exercise validate the accuracy of this procedure.

MCNP SIMULATION OF THE HP(10) ENERGY RESPONSE OF A BRAZILIAN TLD ALBEDO NEUTRON INDIVIDUAL DOSEMETER, FROM THERMAL TO 20 MeV.

The Brazilian Instituto de Radioproteção e Dosimetria (IRD) runs a neutron individual monitoring system with a home-made TLD albedo dosemeter. It has already been characterised and calibrated in some reference fields. However, the complete energy response of this dosemeter is not known, and the calibration factors for all monitored workplace neutron fields are difficult to be obtained experimentally. Therefore, to overcome such difficulties, Monte Carlo simulations have been used. This paper describes the simulation of the HP(10) neutron response of the IRD TLD albedo dosemeter using the MCNPX transport code, for energies from thermal to 20 MeV. The validation of the MCNPX modelling is done comparing the simulated results with the experimental measurements for ISO standard neutron fields of (241)Am-Be, (252)Cf, (241)Am-B and (252)Cf(D2O) and also for (241)Am-Be source moderated with paraffin and silicone. Bare (252)Cf are used for normalisation.

Measurement of the 250Cf component in a 252Cf neutron source at KRISS.

Neutron emission rate measurements have been carried out at the Korea Research Institute of Standards and Science using a manganese sulphate bath system for (252)Cf and (241)Am-Be sources since 2004. The relative measurement method was chosen in 2012, and the neutron emission rates agreed with those by the absolute measurement method within uncertainties. The neutron emission rate of an old (252)Cf source has been measured three times: in 2004, 2009 and 2012. The (250)Cf component was fitted to a double-exponential function of (252)Cf+(250)Cf, and the ratio of the (250)Cf component to the (252)Cf component was estimated to be 7.8 % in 2004 and 46.8 % in 2012. Underestimation of the neutron emission rates of old (252)Cf sources can be corrected if the neutron emission rate of the (250)Cf component is taken into account.

Addressing nuclides not in the CAP88-PC Version-3 library.

Versions of the computer program, CAP88, are widely used to calculate the radiological doses from radionuclides emitted into the air. CAP88-PC Version-3 includes an extensive library of radionuclides, but there are many more that are not included. Surrogates are often used to substitute for nuclides not in the library, though the results are usually overestimates. This paper addresses nuclides that are not in the library and describes methods to obtain more accurate results.