Neutron Dosimeter Research Articles

Ambient dosimetry in pulsed neutron fields with LINrem detectors

The status of the LINrem project is presented, focusing on the development of innovative neutron dosimeters with enhanced energy sensitivity, time resolution, and portability. Designed to meet the technical demands of radiation protection in modern particle and nuclear facilities, these dosimeters are discussed in detail. Results from experimental campaigns showcasing their efficacy in pulsed fields generated by fusion plasmas and high-intensity pulsed lasers are presented. Additionally, prospects and future plans for the LINrem project are outlined.

Evaluation of neutron dosimetry capabilities with the MC-15 portable multiplicity counter

This work proposes a preliminary neutron dose rate estimation method for a neutron multiplicity detector through measurement- and simulation-based analyses. Uncharacterized neutron-emitting sources may be encountered in situations such as nuclear emergency response, safeguards, and treaty verification. These circumstances may present irradiation risk to personnel conducting field assay, search, and characterization measurements. It is therefore of interest to provide a field neutron dosimetry capability with the existing neutron multiplicity counting (NMC) capabilities. To date, no commercially-available neutron detection systems are capable of both accurate NMC and real-time neutron dosimetry. This work will focus on estimating dose rate using input from a single fielded NMC called the MC-15. The energy-dependent neutron detection efficiency response of the MC-15 was quantified in monoenergetic neutron simulations and evaluated in response to two neutron-emitting sources and to a polyethylene-moderated source. The results were compared to existing neutron dosimeters and established the proof of concept for further investigation of the MC-15 for dose estimation. Measurement results were also replicated in simulations; additional simulations were then conducted to expand upon the limited empirical data. The initial empirical results provided a conversion factor appropriate for use when measuring 252Cf neutrons that is independent of polyethylene shielding presence and thickness. The simulated data sets were then used to evaluate a fit equation allowing estimation of the neutron dose rate for less restricted geometric configurations and dependent only on the distance between the source and the detector. Additionally, the energy dependence of the efficiency response indicates that further empirical evaluations could provide energy-dependent conversion factors for broader neutron dosimetry capabilities with a wider range of neutron-emitting sources.

A modular, high dynamic range passive neutron dosimeter and imaging diagnostic.

The multi-decade neutron dosimeter and imaging diagnostic (MDND) is a passive diagnostic that utilizes the polyethylene (n, p) nuclear reaction to enhance the diagnostic's sensitivity for time and energy integrated neutron measurements in the range of 2.45-14.1MeV. The MDND utilizes a combination of radiochromic film, phosphor image plates, and solid-state nuclear track detectors, with the goal of providing several orders of magnitude of dynamic range in terms of measured neutron fluence. The diagnostic design was guided by simulations in the Monte Carlo N-Particle (MCNP) transport code to determine the optimum thickness of the polyethylene convertor for maximum proton fluence incident on the detection medium as a function of incident neutron energy. In addition, the simulation results of complete diagnostic assemblies, or "stacks," were used to determine the total dynamic range of an MDND in terms of measured neutron source yield, which was found to be between around 107 and 1015 emitted into 4π with the detector located 1m away from the source. Complimentary to these simulations, individual detectors within a stack were simulated and analyzed to determine response as a function of neutron energy and yield. This work presents the diagnostic design, MCNP simulation results, and analysis of expected signals for varying neutron sources.

Neutron dose evaluation inside high-energy accelerator irradiation vault using LR-115

Recently, a new radiation therapy using a high-energy accelerator irradiation vault has attracted significant attention. This therapy is very effective in destroying cancer cells because it uses much higher energy than conventional radiation therapy. Nevertheless, it also has disadvantages due to its high energy and dose, resulting in the generation of secondary radiation such as X-rays and neutrons. In particular, neutrons have a higher radiation weighting factor than photons; therefore, they are more harmful to normal tissues. However, the popular neutron dosimeter CR-39 cannot evaluate high-dose neutrons.This paper proposes a novel method for measuring high-dose neutrons generated during cancer treatment. LR-115, which has low detection efficiency but is expected to be useful for evaluating high-dose neutrons, was used to evaluate the effective dose of neutrons in a high-energy accelerator irradiation room. The results were verified using MCNP 6.2, a Monte Carlo-based particle transport code. This study provides the lower limit for evaluating neutron effective dose using LR-115. The experimental results showed that the neutron effective dose increased linearly with beam current. These findings provide basic data for evaluating neutron effective dose for radiation protection.

Design and testing of an irradiation room with low room-scattering for neutron calibration

IntroductionThis laboratory plans to establish neutron reference radiation fields with three neutron sources to calibrate neutron-measuring devices. To perform calibration at multiple dose rates, neutron ambient dose equivalent rate H˙*(10) needs to range 1 μSv/h to 10 mSv/h. The lower limit requires that the maximum available calibration distance should be at least 4.5 m. MethodsTo reduce room-scattered neutrons and extend the available calibration distance, MC simulations were conducted to determine the material and size of the irradiation room. A 3″ Bonner sphere and a LB6411 environmental neutron dosimeter were used to characterize the irradiation room. ResultsA 14.32 × 14.32 × 12.00 m3 irradiation room was built based on simulation results. Floor, roof, and walls are made of 75 cm concrete covered by a coating layer of 2 cm BPE and 3 cm PE. Experimental maximum available calibration distance reaches 4.65 m. The range of H˙*(10) for calibration covers 1 μSv/h to 10 mSv/h. Neutron and photon H˙*(10) outside the room are within 0.19 μSv/h and 0.22 μSv/h, respectively.

Condensed DNA incorporating mercaptoundecahydro-closo-dodecaborate (BSH) coated gold nanoparticles as a model system for boron neutron capture therapy (BNCT)

The experimental radiation therapy involving thermal neutron capture by boron (boron neutron capture therapy, BNCT) offers the advantage of high-LET ions with ranges on the order of a cell nucleus. Efforts to implement it have encountered serious difficulties. A practical model system would be able to address the underlying mechanisms of DNA damage and also calibrate Monte Carlo simulations. We describe the characterization of a plasmid-based model in which DNA condensed with a tetra-arginine peptide is co-aggregated with mercapto-closo-dodecaborate (BSH) coated gold nanoparticles (AuNPs). Condensed DNA is an excellent model for cellular chromatin, and the AuNPs are able to function as boron carriers and neutron dosimeters. Data from light scattering, UV–visible spectroscopy, fluorescence spectroscopy, sedimentation behavior, gel electrophoresis, and atomic force microscopy all indicate that the basic peptide acts to co-aggregate the two polyanionic species DNA and BSH-coated AuNPs. This aggregation is easily reversed by an increase in ionic strength to free the plasmid for subsequent assay. We argue that this three-component system is simpler, more convenient, and more flexible than other models requiring covalent attachments between DNA, boron, and/or gold.

Effect of gamma-irradiation on the photoluminescence of silver-doped lithium triborate glass.

Optical-luminescence properties, including absorption spectra, photoluminescence excitation spectra, and changes in photoluminescence spectra after irradiation with γ-rays in the dose range of 1-3Gy, were measured for novel L i B 3 O 5:A g dosimeter glass. A clear dependence of the intensity of the photoluminescence (PL), with a maximum of approximately 300nm, on the dose value when excited by light with λ e x c =220n m was observed. The mechanism of PL in γ-irradiated L i B 3 O 5:A g glass is proposed to be a consequence of radiation annihilation with the emission of relaxed exciton-like electronic excitations with the participation of impurity defects (A g 0) in the glass structure. Considering that L i B 3 O 5 has an effective atomic number of Z e f f =7.39, which is the closest to Z e f f =7.42 of human body tissue, L i B 3 O 5:A g glass can be very promising for γ-dosimetry in medical practice during radiation therapy of patients with cancer. The presence of lithium Li(6) and boron B(10) isotopes increases the sensitivity of L i B 3 O 5:A g glass to neutrons, which is promising for the manufacture of individual neutron dosimeters for applications in a new field of radiation therapy-neutron therapy.

Intercomparison of neutron personal dose equivalent measured by thermoluminescence dosimeters

An intercomparison of neutron personal dose equivalent measured by the Harshaw thermoluminescence neutron dosimeters (TLDs) between the National Institute of Metrology of China (NIM) and the Institute for Nuclear Science and Technology of Vietnam (INST) was performed. Three sets of TLDs (each set consisting of five TLDs) were prepared for each laboratory. Each set was then irradiated to the corresponding same nominal standard value of neutron personal dose equivalent, Hp(10)n−stdi, of 1.0, 2.0, and 3.0 mSv, respectively at these two laboratories. The irradiated TLDs were then read-out at the INST using the Harshaw 4500-type TLD reader to obtain neutron personal dose equivalents at the NIM, Hp(10)n−NIMi and at the INST, Hp(10)n−INSTi, which are corresponding to different values of Hp(10)n−stdi. The TLDs’ responses to different scattered components of neutrons in these two fields are also discussed. Comparisons between the corresponding pair values of Hp(10)n−NIMi and Hp(10)n−INSTi show good agreements within 10% with the standard uncertainty of 12.5% (k = 1). The measured values of Hp(10)n−NIMi and Hp(10)n−INSTi are satisfied the Trumpet curve criteria. This implies that the TLDs can be used for safety assessment of occupational neutron personal dose equivalents. This intercomparison result also confirms the capabilities of these two laboratories (i.e., NIM, INST) on deliveries of neutron personal dose equivalent standard values for calibrations.

The contribution of the air and room scattered neutrons in the PTB calibration facility

Neutron detectors and dosimeters used for radiation protection purposes are commonly calibrated with radioisotope-based neutron sources in dedicated calibration facilities. Due to interactions of the neutrons emitted by the sources with components of the calibration room, the neutron fluence at the measuring point is a sum of several components. As the results of calibration procedures should ideally be independent of the environment in which the measurements are performed, the effects caused by neutrons which are not directly reaching the device under test need to be considered, i.e. the influence of scattered neutrons needs to be determined and corrections have to be made. The contribution of the neutrons scattered from the air and the room to the fluence and dose must be taken into consideration, especially in the case of small and medium sized calibration rooms. The scattered neutron component can be evaluated numerically provided a detailed model of the calibration facility is available. In this work, a realistic MCNP6 model of the calibration room at the Physikalisch-Technische Bundesanstalt (PTB) has been developed in order to evaluate carefully the contribution of scattered neutrons from the air, wall, and equipment in the irradiation room for 252Cf, 241Am–Be and D2O moderated 252Cf (with and without Cd-shielding) neutron sources at different reference points and with and without shadow objects. The calculation also considers new fission spectral data for the neutron reference field and recent cross section data libraries. The results show that the contribution of air out-scattered neutrons depends on the detector position and on the neutron source spectrum. This effect is particularly significant for the moderated 252Cf source. More than a third of the total scored neutrons come from the walls and ceiling and the total contributions of the scattered neutrons to the total fluence at the reference position 170 cm away from the source are 50.6 %, 46.6 %, 55.0 % and 52.5 % for 252Cf, 241Am–Be, D2O-252Cf–Cd and D2O-252Cf sources respectively.

Analysis of EURADOS neutron intercomparison results according to new ISO standards criteria.

The European Radiation Dosimetry Group has carried out several different types of intercomparison (IC) exercises in the past that qualify as proficiency tests for different dosimetry systems and types of radiation. The first neutron dosemeter IC was held in 2012 (IC2012n) and was followed by a second in 2017/2018 (IC2017n). In sum, 31 Individual Monitoring Services (IMSs) entered 34 dosimetry systems in IC2012n, and 32 IMSs entered 33 dosimetry systems for IC2017n. Such exercises provided a rare opportunity to see how neutron dosemeters perform. For the IC2012n exercise, there were no applicable performance standards for neutron personal dosemeters. ISO/TC85/SC2 updated the ISO Standard 14146 in 2018 (ISO 14146:2018. Radiation protection-Criteria and performance limits for the periodic evaluation of dosimetry services) to include neutron dosimetry. It was thus possible to analyse the IC2017n exercise in accordance with the requirements given by this new standard. It is now of interest to reanalyse the results of IC2012n to quantify any modifications to the conclusions.

Personal dosimetry at the Paul Scherrer Institute.

The Paul Scherrer Institute (PSI) is the largest research institute for natural and engineering sciences in Switzerland. PSI develops, builds and operates complex large research facilities. Every year, >2400 scientists from Switzerland and around the world come to PSI to use the facilities and to carry out experiments. Many areas at PSI are radiation protection areas. Depending on the radiation protection area, the work carried out and the time the users spend in these areas, they have to carry a personal dosemeter. PSI runs an individual monitoring service in compliance with the Swiss legislation on radiological protection and approved by the Swiss Federal Nuclear Safety Inspectorate. The service provides about 35000 dosemeters per year for the internal and external customers consisting of whole-body dosemeters for photons and neutrons as well as extremity dosemeters. This paper gives an overview on the employed personal dosimetry techniques by the individual monitoring service of PSI, the number of distributed dosemeters for internal and external customers and statistics about the measured doses at PSI over 30years.

Revision of the 21909 standard on passive personal neutron dosemeters.

Since 2012, International Standardization Organization TC85/SC2 Working Group 19 dealing with 'Individual monitoring for external radiations' has been conducting important work revising the 21909 standard on passive personal neutron dosemeters. A consensus in the community acted that the 2005 version( 1) needed to be revised for two main reasons: (1) to achieve a document applicable to all systems irrespectively to the considered techniques and (2) to reach harmonisation between performance tests and conditions of use at the workplaces. The document was completely rewritten to be binding enough to ensure that any neutron dosimetry system fulfilling the criteria would be reliable in most of the usual work situations, in terms of dose level, energy and direction distributions of the neutron fluence. This paper explains the reasons behind the revision and the new philosophy of the series published in 2021, so that Individual Monitoring Services may better understand how to apply the requirements in practice.

Assessment of the impact on a PADC neutron dosimetry system of the new operational dose quantities proposed in ICRU report 95.

The International Commission on Radiation Units and Measurements, ICRU, have recommended, in ICRU Report 95, the adoption of new operational quantities for use in radiation protection monitoring. Their implementation will pose problems for instrument manufacturers supplying area dosemeters and for individual monitoring services providing personal dosemeters for radiation workers. The problems arise for photon, neutron and beta dosimetry, but this paper deals solely with neutron personal dosemeters based on poly-allyl diglycol carbonate (PADC) plastic. It presents information on the problems to be expected based on data from calibrations of actual personal neutron dosemeters in terms of the existing dose quantities.

DOIN: a novel electronic personal dosemeter for neutrons

Electronic personal dosemeters (EPD) are powerful tools for achieving ALARA (As Low As Reasonably Achievable) objectives in operational radiation protection. EPD for photons are well developed and their performances usually comply with relevant Standards. By contrast, a very few commercial models exist for neutrons and their energy dependence is too large for using them without pre-information on the workplace neutron spectrum. Within the INFN-based DOIN (DOsimetro Indossabile per Neutroni) project, a new EPD for neutrons was prototyped. Owing to a new patented design, a good dose response was achieved in the energy interval from thermal neutrons up to the quality of ^{241}Am-Be. The response is nearly isotropic and, if compared to commercial models, exhibits higher sensitivity.

Comparison and estimation of photoneutron dose produce between 10 MV flattened and unflattened beam in Elekta Versa HD™ medical linac.

In a high-energy medical linear accelerator (linac), if the interaction of photon energy is higher than the neutron binding energy of high atomic material, it emits a neutron field through photonuclear (γ, n) reaction. The current study, evaluates the photoneutron dose equivalent (PNDE) produced between the 10 MV flattened and unflattened beams as a function of field sizes in the Elekta Versa HD™ linac. The PNDE produced from Versa linac was recorded along the patient plane using the bubble detector personal neutron dosimeter and from the measured PNDE values, the theoretical PNDE values were simulated for various field sizes using nonlinear least-squares curve-fitting as a function of a polynomial. The percentage of deviation (PoD) and Chi-square (χ2) tests were performed between the measured and simulated PNDE values to study the reliability and validity. The results show that the mean PoD between the measured and simulated PNDE values for respective positions of a field size of FF beam was found to be -1.99% for 0.3×0.3, -4.39% for 5×5, -3.868% for 10×10, 0.590% for 15×15, 9.18% for 20×20, -4.133% for 25×25, and 0.467% for 30×30 cm2. Similarly, the mean PoD between the measured and simulated PNDE values for flattening filter-free (FFF) beam was found to be 1.36% for 0.3×0.3, -1.39% for 5×5, -5.38% for 10×10, 4.41% for 15×15, 3.84% for 20×20, 5.69% for 25×25, and -1.75% for 30×30 cm2. The maximum deviation between the measured and simulated PNDE values lies within the range ± 5%. From the study, it is observed that the FFF beam produces lesser neutron contamination than the FF beam.

ZnO-6LiF/polystyrene composite scintillator for thermal neutron radiation detection.

We report the preparation of the ZnO-6LiF composite with a polystyrene (PS) polymer as a host using the solution mixing process. 6LiF acts as a converter material that absorbs a thermal neutron and produces alpha particles, which excites ZnO micro-particles, resulting in UV-vis photons' emission. The free-standing ZnO-6LiF/PS composite film is coupled to a photomultiplier tube (PMT). 241Am-Be (1Ci) is used as the neutron radiation source for measuring the response. We compared the response of the composite scintillator consisting of (i) natural LiF and (ii) 95% 6Li enriched LiF (6LiF). The increased pulse heights are recorded for 95% 6Li enriched, i.e., 6LiF converter. It confirms the generation of alpha particles after the absorption of a neutron in 6LiF. Furthermore, ZnO and 6LiF are considered in different weight proportions, 2:1, 1:1, and 1:2, keeping the total loading 50% (w/w) of polystyrene. The ZnO:6LiF (1:1)/PS composite showed higher scintillation pulse heights than the other two composites. Repetitive measurements are performed for the ZnO-6LiF(1:1)/PS composite, showing ±5% variation in respective responses. We also investigated the impact of different counting times and source-to-detector responses for the ZnO-6LiF(1:1)/PS composite. The response increases linearly with neutron dose, exhibiting a sensitivity of ∼203 counts/μSv. Neutron measurement counts at different source-to-detector distances have a similar trend as that of neutron dose measured by using a neutron dosimeter. Thus, this work demonstrated the potential of the ZnO-6LiF/PS composite, coupled to PMT for detecting thermal neutron radiation.

DEVELOPMENT OF AN EXTENDED-RANGE BONNER SPHERE SPECTROMETER FOR CHINA INITIATIVE ACCELERATOR-DRIVEN SYSTEM.

To measure the extended-range neutron spectra and calibrate the extended-range neutron dosemeters of the China initiative Accelerator-Driven System (CiADS), an Extended-range Bonner Sphere Spectrometer (EBSS) has been developed. The EBSS was designed based on the PHITS codes, investigating various combinations of materials and diameters of the neutron moderators and the neutron multipliers for extended-range neutrons. Finally, seven polyethylene-only spheres and seven extended-range spheres were selected and subsequently built. The neutron multipliers of the extended-range spheres embedded concentric shells of lead, copper and tungsten. The response functions of the EBSS were analyzed and experimentally validated. It was subsequently tested with 252Cf neutron source and cosmic ray neutron source. The results demonstrate that the EBSS is capable of accurately measuring neutron spectra.

RESPONSE OF CURRENT AREA DOSEMETERS TO NEW OPERATIONAL QUANTITIES.

New operational quantities for external radiation exposure presented by the ICRU brought changes, which can affect the use of current dosemeters. Dosemeters calibrated in the old quantities should be tested, and the results obtained should be analyzed with respect to the new quantities. The difference between the old and new quantities depends primarily on photon energy. This work concerns instruments for area monitoring, which were exposed to reference photon and neutron spectra of mean energies from 65 to 1250keV and from 0.02 to 10MeV, respectively. As expected, it was revealed that the current photon dosemeters overestimate the new quantity ambient dose. For the measurements within the energy interval chosen, it seems to be acceptable to implement a correction factor to optimize the response. For the purposes of measurements of photons of lower energies, further research would be needed. To adapt the response of the neutron dosemeter tested, recalibration or redesign can be considered, but no fundamental changes seem to be necessary for the investigated spectra.

ИГОРЬ БОРИСОВИЧ КЕИРИМ-МАРКУС (К 100-ЛЕТИЮ СО ДНЯ РОЖДЕНИЯ)

The article contains material on the scientific activity and life path of a well-known scientist in the field of radiation safety, the first head of the laboratory of emergency and individual dosimetry, one of the developers of the first neutron dosimeter in the USSR. This publication is a historical review of materials from the collection of the Museum of the FMBTS. A.I.Burnazyan, the archive of the scientist’s family, printed publications, is of scientific interest because it holistically presents the life and scientific path of the famous scientist who gave 58 years of service to science within the walls of the Institute of Biophysics (now part of the A.I.Burnazyan FMBC) . The scientist made a great contribution to the formation of a new area of ​​science at the intersection of dosimetry, radiobiology and radiation medicine, named equidosimetry at his suggestion. The museum keeps the book of I.B.Keirim-Markus "Equidosimetry", where for the first time in the world scientific literature summarized materials on the main sections of equidosimetry, the history of its occurrence, tasks, criteria for assessing the effect of radiation, in particular in large doses, the values ​​of equidosimetry, methods and instruments for their measurement.

Radiation field characterization with emphasis on the collimator configuration at the compact neutron source RANS-II facility

ABSTRACT The RIKEN Accelerator-driven compact Neutron Source-II (RANS-II) based on the 7Li(p,n)7Be reaction with 2.49 MeV proton injection was installed in a small area of the experimental hall, where the scattered radiation is expected to strongly influence experiments. Therefore, we have begun to study the radiation characteristics of the RANS-II hall by simulation with the Particle and Heavy Ion Transport code System (PHITS) and by experiments to clearly identify the scattered component. In addition, to suppress the background, a collimator is studied with respect to its size, shape, and materials. With the use of the 30 mm borated polyethylene collimator, we found that the background becomes less than 1 of the extracted neutron beam intensity 1.0 m from the Li target. For the experiment, neutron suppression was characterized by measuring the neutron dose rate with a neutron dosimeter and a collimator. The suppression ability of the neutron beam was confirmed by comparing the results of simulations with those of experiments.