Neutron Radiation Fields Research Articles

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.

Reconstruction of the neutron radiation field on nuclear facilities near the shield using Bayesian inference

This work aims to reconstruct the neutron radiation field on nuclear facilities by Bayesian inference. Firstly, the net function interpolation method is used to reconstruct the neutron radiation field on the basis of the observed data. Taking into account the effects of radiation shielding, the neutron radiation field near the shields should be reconstructed again. According to the measured data, Bayesian inference succeeds in reconstructing a more accurate neutron radiation field. The effectiveness of this method can be verified by two cases. In the first case, there are two calibration areas. Compared with the neutron radiation field simulated by Monte Carlo method, the average relative errors of the net function method are 11.37% and 23.65%, respectively. After the dose rate distribution of the calibration areas is calibrated by Bayesian inference, the average relative errors become 3.70% and 6.67%. There are six calibration areas in the other case, and Bayesian inference can also improve the accuracy.

Mathematic Model of Neutron Dose Rates in the EAST Hall

Radiation particles and rays will be produced during each nuclear fusion plasma discharge, especially neutrons which will be largely emitted out with high intensity and energy under the neutral beam injection (NBI) discharge mode. To well obtain the neutron dose rates during the nuclear fusion plasma discharge, 3He-based neutron detector is adopted in this paper, and to fully obtain the neutron dose rates in the Experimental Advanced Superconducting Tokamak (EAST) hall, three neutron detectors have been set at three selected monitoring sites with a long distance to the monitoring personal computer (PC), and the monitoring system, based on the RS-485 differential communication mode, was designed and setup. Along with each nuclear fusion plasma discharge, the neutron radiation field should appear as a pulsed radiation field. Based on the neutron dose rates, saved by the monitoring system, the neutron radiation dose rates were studied during one NBI-based nuclear fusion plasma discharge. Finally, the Gaussian function model of neutron dose rates during one nuclear fusion plasma discharge was obtained and presented.

Performance Comparison Between SiC and Si Neutron Detectors in Deuterium–Tritium Fusion Neutron Irradiation

Silicon carbide (SiC) detector is considered to be a good substitute for silicon (Si) detector in neutron detection, considering its mature of device fabrication in large sensitive area and thick sensitive volume, high charge-collection efficiency, significant neutron–gamma discrimination, and excellent endurance in harsh environment, and it is expected to meet the demands of neutron detection in increasingly harsh irradiation field. The performance comparison between SiC and Si neutron detector in neutron irradiation is an important issue but is not fully studied: no researcher has ever made a comparative study on them in constant fusion neutron irradiation. In this paper, the SiC and Si detectors irradiated by constant deuterium–tritium fusion neutron irradiation were tested and compared, getting significant difference in performance degradation: a significant degradation was found in the Si detector at a neutron fluence of $1.65\,\,\times \,\,10^{13}$ cm−2, which showed a marked increase in dark current (over four orders of magnitude) and a serve reduction in the peak centroid of the alpha spectrum (over 95%), while nearly no degradation was observed in SiC detector at a much higher neutron fluence ( $3.82\times 10^{13}$ cm−2). The SiC detector is proven to have better radiation resistance than the Si detector at a fluence of around $10^{13}$ cm−2, and thus, it could replace the traditional Si detector in constant neutron irradiation field at least at a fluence of $10^{13}$ cm−2.

Using a Lithium Glass Scintillator to Record Neutron Radiation

Studies have been performed to improve the accuracy of energy measurements of neutron fields and to improve the spectrometric analysis of neutron radiation fields using scintillation detectors. The use of a scintillation detector based on 6Li2O lithium glasses of various diameters and thicknesses for recording neutron radiation energy, 0.4–15·106 eV, is considered. The main characteristics of this scintillation detector and the results of measurements obtained with it are compared with the characteristics and results of measurements of the scintillator using LiI (Eu). Spherical moderators with diameters of 70–300 mm were used for scintillation detector operation.

Degradation of stator insulation of high-voltage asynchronous machines in gamma and neutron radiation field

This paper presents the results of an examination of function stability of high-voltage asynchronous motors exposed to ambient strain caused by combined neutron and gamma radiation. This problem appears in practice when a high-voltage asynchronous motor is used in nu- clear power plants where it can be exposed to this type of ambient strain. The failure of the engine's operation under such conditions may have unexpected consequences. As more than 50 % of failure (malfunction) of high-voltage asynchronous motors is caused by damage to stator insulation, the focus of the paper was on testing the effects of combined neutron and gamma radiation on stator insulation. The tests were carried out under well-controlled laboratory conditions on samples taken from both new and used factory coil windings. Two-layer samples were used to record partial discharge threshold voltage and breakdown voltage. By comparing the experimentally obtained results with the applicable mathematical-statistical procedure, an estimate was made of the aging acceleration of stator insulation and the time duration of reliable operation of high-voltage asynchronous motor was determined by life-time exponent.

Comparison of Biodosimetry Biomarkers for Radiation Dose and Injury Assessment After Mixed-Field (Neutron and Gamma) and Pure Gamma Radiation in the Mouse Total-Body Irradiation Model.

The detonation of a nuclear weapon and the occurrence of a nuclear accident represent possible mass-casualty events with significant exposure to mixed neutron and gamma radiation fields in the first few minutes after the event with the ensuing fallout, extending for miles from the epicenter, that would result primarily in photon (gamma- and/or x-ray) exposure. Circulating biomarkers represent a crucial source of information in a mass-casualty radiation exposure triage scenario. We evaluated multiple blood biodosimetry and organ-specific biomarkers for early-response assessment of radiation exposure using a mouse (B6D2F1, males and females) total-body irradiation model exposed to Co gamma rays over a broad dose range (3-12 Gy) and dose rates of either 0.6 or 1.9 Gy min and compared the results with those obtained after exposure of mice to a mixed field (neutrons and gamma rays) using the Armed Forces Radiobiology Research Institute Co gamma-ray source and TRIGA Mark F nuclear research reactor. The mixed-field studies were performed previously over a broad dose range (1.5-6 Gy), with dose rates of either 0.6 or 1.9 Gy min, and using different proportions of neutrons and gammas: either (67% neutrons + 33% gammas) or (30% neutrons + 70% gammas). Blood was collected 1, 2, 4, and 7 d after total-body irradiation. Results from Co gamma-ray studies demonstrate: (1) significant dose- and time-dependent reductions in circulating mature hematopoietic cells; (2) dose- and time-dependent changes in fms-related tyrosine kinase 3 ligand, interleukins IL-5, IL-10, IL-12, and IL-18, granulocyte colony-stimulating factors, thrombopoietin, erythropoietin, acute-phase proteins (serum amyloid A and lipopolysaccharide binding protein), surface plasma neutrophil (CD45) and lymphocyte (CD27) markers, ratio of CD45 to CD27, procalcitonin but not in intestinal fatty acid binding protein; (3) no significant differences were observed between dose-rate groups in hematological and protein profiles (fms-related tyrosine kinase 3 ligand, IL-5, IL-12, IL-18, erythropoietin, granulocyte colony-stimulating factor, granulocyte-macrophage colony-stimulating factor, CD27, CD45, and ratio of CD45 to CD27) for any radiation dose at any time after exposure (p > 0.148); (4) no significant differences were observed between sex groups in hematological and protein profiles (fms-related tyrosine kinase 3 ligand, IL-18, erythropoietin, granulocyte colony-stimulating factor, granulocyte-macrophage colony-stimulating factor, serum amyloid A, CD45) for any radiation dose at any time after exposure (p > 0.114); and (5) PCT level significantly increased (p < 0.008) in mice irradiated with 12 Gy on day 7 post-total-body irradiation without significant differences between groups irradiated at dose rates of either 0.6 or 1.9 Gy min (p > 0.287). Radiation-quality comparison results demonstrate that: (1) equivalent doses of pure gamma rays and mixed-field radiation do not produce equivalent biological effects, and hematopoietic syndrome occurs at lower doses of mixed-field radiation; (2) ratios of hematological and protein biomarker means in the Co study compared to mixed-field studies using 2× Co doses vs. 1× TRIGA radiation doses (i.e., 3 Gy Co vs. 1.5 Gy TRIGA) ranged from roughly 0.2 to as high as 26.5 but 57% of all ratios fell within 0.7 and 1.3; and (3) in general, biomarker results are in agreement with the relative biological effectiveness = 1.95 (Dn/Dt = 0.67) reported earlier by Armed Forces Radiobiology Research Institute scientists in mouse survival countermeasure studies.

Design and preliminary test of a multi-element tissue-equivalent proportional counter based on the gas electron multiplier

In order to monitor the intense neutron and gamma radiation, multi-element tissue equivalent proportional counters (TEPC) based on the standard gas electron multiplier (GEM) and ceramic thick GEM (THGEM) foils were designed and respectively tested by implanted 55Fe and 241Am sources which would also be used as calibration sources. The performance testing of GEM-TEPC/THGEM-TEPC detectors at various parameters showed that the detectors could work well. The THGEM-TEPC was also used to measure the linear energy spectra in 241Am–Be and intense (108 ncm−2s−1) neutron radiation fields. The results displayed that THGEM-TEPC was available to do microdosimetric research or be a dosimeter in intense neutron/gamma field.

Long-term neutron radiation levels in distressed concrete biological shielding walls.

Neutron radiation can deteriorate mechanical properties of the concrete materials, and thus it is questionable that neutron transport properties of concrete can remain unchanged during the life span of biological shielding walls. one-speed neutron diffusion equation and heat conduction equation were used as governing equations for prediction of neutron radiation and thermal field in concrete, respectively. The potential variations of transport properties due to neutron radiation and elevated temperature were estimated. A simplified example of a typical concrete biological shielding wall was analyzed up to 80 years, and the results were discussed. The radiation damage and radiation heating lead to minor changes of the temperature profile in the concrete. However, neutron radiation and elevated temperature can result in considerable increases of neutron flux and fluence in the concrete. The damage of concrete induced by neutron radiation and elevated temperature can considerably accelerate the penetration of neutron radiation into the concrete. This work is the first attempt to deal with the degradation of neutron and heat transport properties of concrete and its effect on neutron fluence distribution in concrete, and provides a possible way to determine the long-term neutron and thermal fields in concrete biological shielding walls.

Analytical and quasi-Bayesian methods as development of the iterative approach for mixed radiation biodosimetry

The present paper proposes two methods of calculating components of the dose absorbed by the human body after exposure to a mixed neutron and gamma radiation field. The article presents a novel approach to replace the common iterative method in its analytical form, thus reducing the calculation time. It also shows a possibility of estimating the neutron and gamma doses when their ratio in a mixed beam is not precisely known.

The assessment of shutdown dose rate and radioactive waste of HCSB during its replacement in CFETR

Chinese Fusion Engineering Testing Reactor (CFETR) is a new test superconducting tokamak in China aimed to obtain the state-of-the-art technologies and bridge the gaps between ITER and DEMO. The helium cooled solid breeder blanket (HCSB) is one of three main blanket candidates for CFETR. During reactor operations, the highly radioactive nuclides will be produced by the blanket components which are located in the high-energy neutron radiation field. In order to protect the workers from too high radiation and ensure the normal operation of machinery, the Shutdown Dose Rate (SDR) of HCSB during its replacement is evaluated based on the rigorous two-step (R2S) method. Besides, according to the recommended standards provided by Safety and Environmental Assessment of Fusion Power (SEAFP) of the EU fusion program, the initial evaluation of the high level radioactive waste of the blanket is carried out as well. The results show that, based on current design of HCSB, the maximum SDR of the blanket is decreased to 4728 Sv/h after the reactor shutdown one day. And after 50 years of cooling, most of the structural materials can be recycled by simple remote operation.

О некоторых методических вопросах исследования цитогенетических эффектов у пациентов со злокачественными новообразованиями при нейтронной терапии на циклотроне У-120

Purpose: To study dosimetric characteristics of neutron radiation field, to determine their role in the formation of the total cytogenetic effect in the patient’s body and to assess the cytogenetic dosimetry capabilities in improving the quality of NT. Material and methods: A therapeutic beam with the average neutron energy of ~6.3 MeV was obtained from the V-120 cyclotron. The radiation field of the beam was investigated with the help of two ionization chambers with different sensitivity to neutrons. Chamber with high and low sensitivities were made of polyethylene and graphite, respectively. To exclude the uncertainty associated with the change in beam intensity in time, a dosimeter monitor operating in the integral mode was used. Results: The dependence of the monitor factor on the irradiated area was measured. The distributions of the absorbed dose of neutrons and γ-radiation over the depth of the tissue-equivalent medium were found. The contribution of γ-radiation to the neutron dose was increased from ~10 % at the entry to the medium to ~30 % at a depth of 16 cm. Dose distributions of scattered neutron and γ-radiation in the plane of the end face of the forming device were obtained. The contribution of these radiations to the dose received by the patient’s body was estimated. This contribution was shown to be comparable with that from the therapeutic beam. The analysis of the influence of NT on the estimation of the frequency of chromosome aberrations in the blood of patients was carried out. Conclusion: The frequency of chromosome aberrations in the blood of patients was determined by the whole-body dose, including dose due to scattered radiation. When using equal focal doses, the cytogenetic effect was found to be dependent on the area of the irradiated field and the depth of the tumor in the patient’s body. The differences in the RBE of neutrons and γ-radiation as well as the instability of the therapeutic neutron beam intensity create uncertainties that do not allow for the necessary control over the doses using the cytogenetic dosimetry. Therefore, cytogenetic dosimetry should be combined with an effective instrument dosimetry method. The use of biodosimetry based on the assessment of the frequency of chromosome aberrations is promising for controlling the average whole-body dose, on which the overall radiation response of the body depends.

Waste assessment of European DEMO fusion reactor designs

Predicting the amount of waste produced from a fusion power plant is vital to assess the likely environmental impact, disposal costs, and also to satisfy nuclear regulators. Inventory simulations are ideally suited to this task because they can be used to evolve in time, via numerically solving the ordinary differential rate equations, the chemical composition of reactor materials, both during operation, when exposed to neutron irradiation, and afterwards, during post-operation decay. The radiological response of the time-evolving material composition can then be used to assign a waste class to reactor components. Waste assessment has been performed for designs of the European demonstration power plant (DEMO) concept using the FISPACT-II inventory simulator, and using extensive Monte Carlo simulations of the neutron irradiation fields as input. The masses in each waste class (defined using the IAEA infrastructure) have been charted in time for in-vessel components (including the divertor and blanket), ex-vessel regions such as the coils, and for the reactor vacuum vessel (VV). Comparisons are made between the waste-class masses generated for different tritium breeding blanket concepts. Typical predictions include the observation that the majority of the VV will become low-level waste within 100 years, while plasma-facing components and tritium breeding units will remain classified as intermediate-level waste for longer. The waste classification implications of sub-dividing the large VV is considered, highlighting the potential benefits, for waste disposal and recycling prospects, of being able to separate low activity regions of a component from more active regions.

CONSTRUCTION AND TEST OF A PORTABLE TISSUE-EQUIVALENT PROPORTIONAL COUNTER SYSTEM

Tissue-equivalent proportional counter (TEPC) is a very promising dosemeter for radiation protection in mixed radiation fields. In this article, a portable TEPC system including a spherical homemade TEPC detector and the compact low-power integrated circuit prototype was constructed and tested in gamma and neutron radiation fields. The microdosimetric spectra results displayed that the portable TEPC system could deal with four orders of magnitude voltage pulse signals. And the measured dose equivalent values agreed with reference values, which means the portable TEPC system had fulfilled the test and satisfied the requirement for radiation monitor.

Steady state and transient thermal-hydraulic performance of the IFMIF-HFTM irradiation rigs and container with modeling of nuclear heating

During the EVEDA phase of the International Fusion Materials Irradiation Facility (IFMIF), the High Flux Test Module (HFTM) was developed as dedicated irradiation device for Small Scale Testing Technique (SSTT) material specimens in the intensive IFMIF neutron radiation field. The specimens are contained in temperature controlled irradiation rigs. Since one of the requirements for the HFTM is to provide a uniform temperature field for the irradiated specimens, thermal testing was a priority for the performed validation activities.In the HFTM “single-rig” (HFTM-SR) experiments a single rig of 1:1 scale was tested. The heater plates and the specimen region inside the rig were instrumented with thermocouples to monitor heat transfer and specimen temperature spread.In the High Flux Test Module “double compartment” (HFTM-DC) experiments a fully equipped prototype with three heated rigs was tested in the HELOKA-LP helium loop. Special heater cartridges are used to substitute the nuclear heating. These experiments show that the full range of operation temperatures (250–550°C) required for the IFMIF HFTM could be well achieved and well controlled with and without surrogate nuclear heater power. The temperature spread measured inside a capsule is in the range of +/−3K in the lateral direction and +2K/−8K in the vertical direction for the 350°C reference case, fulfilling very well the requirements. This compares well to an allowed +/−19K range according to the requirements.No unforeseen thermal hydraulic effects like oscillations, hysteresis etc. could be detected. To cool down from 350°C to 50°C it takes roughly 315seconds, heating up 135seconds.

Tritium Challenges and Plans for ITER Pellet Fueling and Disruption Mitigation Systems

ITER will use a Pellet Injection System (PIS) for supplying deuterium-tritium (D-T) fuel to the fusion plasma and a Shattered Pellet Injection (SPI) system for rapidly injecting impurities (argon or neon mixed with deuterium) as solidified pellets into the plasma for the purpose of mitigating the harmful effects of plasma disruptions as part of a Disruption Mitigation System (DMS). Both systems are being designed to handle significant amounts of tritium in the process streams and are exposed to similar environmental conditions during operation, including exposure to gamma and neutron radiation and significant magnetic fields. Multiple barriers to prevent the potential release of the tritium inventory into the environment are included in the designs. The unique environmental conditions present some challenges and are currently being addressed during the design phase.

Pulsed neutron fields measurements around a synchrotron storage ring

A measurement campaign was performed for characterizing the neutron ambient dose equivalent, H*(10), in selected positions at ISA, Aarhus, Denmark, around the ASTRID and ASTRID2 storage rings. The neutron stray radiation field is characterized here by very intense radiation bursts with a low repetition rate, which result in a comparatively low average H*(10) rate. As a consequence, devices specifically conceived for operating in pulsed neutron fields must be employed for efficiently measuring in this radiation environment, in order to avoid severe underestimations of the H*(10) rate. The measurements were performed with the ELSE NUCLEAR LUPIN 5401 BF3-NP rem counter, a detector characterized by an innovative working principle that is not affected by dead time losses. This allowed characterizing both the H*(10) and the time structure of the radiation field in the pre-selected positions.

Performance tests and comparison of microdosimetric measurements with four tissue-equivalent proportional counters in scanning proton therapy

This paper compares the performance of four different Tissue-Equivalent proportional counters (TEPC) first in standard radiation fields, with gamma and neutron sources, then in the mixed and complex/intense neutron and photon stray radiation field of a scanning proton therapy facility. The paper focuses on the dead time correction and introduces a new spectra processing methodology to enable the comparison of the four TEPCs while accounting for their different gas filling and gain, lineal energy range of the spectrum and the analysis methodology. Measurements with 137Cs and/or 60Co gamma sources demonstrate variable low-LET threshold for each TEPC while data acquired with a 252Cf neutron source show comparable response of the four TEPCs for high-LET particles. Meanwhile, in the scattered field of proton therapy, microdosimetric spectra measured at different positions and orientations around the patient show a majority of high-LET events at the smallest angle with respect to the beam axis while low-LET particles were mainly dominant at 90° from the beam axis. The introduced processing methodology led to good overlapping of microdosimetric spectra for the four systems.

ATLAS-TPX: a two-layer pixel detector setup for neutron detection and radiation field characterization

A two-layer pixel detector setup (ATLAS-TPX), designed for thermal and fast neutron detection and radiation field characterization is presented. It consists of two segmented silicon detectors (256 × 256 pixels, pixel pitch 55 μm, thicknesses 300 μm and 500 μm) facing each other. To enhance the neutron detection efficiency a set of converter layers is inserted in between these detectors. The pixelation and the two-layer design allow a discrimination of neutrons against γs by pattern recognition and against charged particles by using the coincidence and anticoincidence information. The neutron conversion and detection efficiencies are measured in a thermal neutron field and fast neutron fields with energies up to 600 MeV. A Geant4 simulation model is presented, which is validated against the measured detector responses. The reliability of the coincidence and anticoincidence technique is demonstrated and possible applications of the detector setup are briefly outlined.

Monte Carlo based investigation of a universal two-component albedo neutron dosimeter in a deep geological disposal system for high-level nuclear waste

Albedo neutron dosimeter utilizes the property of the human body to moderate incident high energy neutrons, which are scattered and reflected as thermal and low energy neutrons that can be measured by thermoluminescence detector (TLD) with high sensitivity. The current study is devoted to the assessment of possible utilization of the albedo neutron dosimeter as a personal monitoring device in a deep geological disposal system for high-level nuclear waste. When a nuclear waste package is transported to its final disposal position in a deep geological repository, such as in a horizontal emplacement drift in a rock salt mine, the question about the appropriate calibration factor of the albedo neutron dosimeter arises. Neutrons backscattered by the surrounding host rock layers might contribute significantly to the composition of the radiation field and in turn influence the response of the neutron detector. To account for different radiation field compositions, a field calibration technique providing field-specific calibration factors was applied in the current study. The field calibration of an albedo neutron dosimeter in a deep geological repository was performed with Monte Carlo method. The radiation field was simulated by a shielding cask loaded with a 252Cf neutron source that is placed in a horizontal drift of a rock salt mine. Calibration factors were calculated at various distances to the shielding cask. In the investigated case, the neutron radiation field in a geological repository can be assigned to the calibration factor N1, which is recommended for reactors and accelerators with heavy shielding.