Implementation of optically simulated luminescent dosimeter for quality control of gamma ray dose of an accelerator-based neutron source.

BACKGROUND: Neutron beams utilized for performing BNCT are composed of a mixture of neutrons and gamma rays. Although much of the dose delivered to the cancer cells comes from the high LET particles produced by the boron neutron capture reaction, the dose delivered to the healthy tissues from unwanted gamma rays cannot be ignored. With the increase in the number of accelerators for BNCT, a detector system that is capable of measuring gamma ray dose in a mixed neutron/gamma irradiation field is crucial. Currently, BeO TLDs encased in quartz glass are used to measure gamma ray dose in a BNCT irradiation field. However, this type of TLD is no longer commercially available. A replacement dosimetry system is required to perform the recommended ongoing quality assurance of gamma ray measurement for a clinical BNCT system. PURPOSE: The purpose of this study is to evaluate the characteristics of a BeO OSLD detector system under a mixed neutron and gamma ray irradiation field and to assess the suitability of the system for routine quality assurance measurements of an accelerator-based BNCT facility. METHODS: The myOSLD system by RadPro International GmbH was evaluated using the accelerator-based neutron source designed for clinical BNCT (NeuCure BNCT system). The readout constancy, linearity, dose rate effect, and fading effect of the OSLD were evaluated. Free-in-air and water phantom measurements were performed and compared with the TLD results and Monte Carlo simulation results. The PHITS Monte Carlo code was used for this study. RESULTS: The readout constancy was found to be stable over a month-long period and similar to the TLD results. The OSLD readout signal was found to be linear, with a high coefficient of determination (R2 ≥ 0.999) up to a proton charge of 3.6 C. There was no significant signal fading or dose rate dependency. The central axis depth dose and off-axis dose profile measurements agreed with both the TLD and Monte Carlo simulation results, within one standard deviation. CONCLUSION: The myOSLD system was characterized using an accelerator system designed for clinical BNCT. The experimental measurements confirmed the OSLD achieved similar, if not superior to, the currently utilized dosimetry system for routine QA of an accelerator-based BNCT system. The OSLD system would be a suitable replacement for the current TLD system for performing routine QA of gamma ray dose measurement in a BNCT irradiation field.

Dissolution of Molybdenum in Hydrogen Peroxide: A Thermodynamic, Kinetic and Microscopic Study of a Green Process for 99mTc Production.

99mTc-based radiopharmaceuticals are the most commonly used medical radioactive tracers in nuclear medicine for diagnostic imaging. Due to the expected global shortage of 99Mo, the parent radionuclide from which 99mTc is produced, new production methods should be developed. The SORGENTINA-RF (SRF) project aims at developing a prototypical medium-intensity D-T 14-MeV fusion neutron source specifically designed for production of medical radioisotopes with a focus on 99Mo. The scope of this work was to develop an efficient, cost-effective and green procedure for dissolution of solid molybdenum in hydrogen peroxide solutions compatible for 99mTc production via the SRF neutron source. The dissolution process was extensively studied for two different target geometries: pellets and powder. The first showed better characteristics and properties for the dissolution procedure, and up to 100 g of pellets were successfully dissolved in 250-280 min. The dissolution mechanism on the pellets was investigated by means of scanning electron microscopy and energy-dispersive X-ray spectroscopy. After the procedure, sodium molybdate crystals were characterized via X-ray diffraction, Raman and infrared spectroscopy and the high purity of the compound was established by means of inductively coupled plasma mass spectroscopy. The study confirmed the feasibility of the procedure for production of 99mTc in SRF as it is very cost-effective, with minimal consumption of peroxide and controlled low temperature.

Operational status of and upgrade plan for the 100-MeV proton linac at the Korea multi-purpose accelerator complex.

The report presents the operation status of and upgrade plan for the 100-MeV proton linac at the Korea Multi-purpose Accelerator Complex (KOMAC). First, an operation history of the 100-MeV linac since its commissioning in 2013, such as operation hours, user services, machine availabilities, and downtimes, is discussed. Second, the status of the beamlines in service or under development is described in a detailed manner. Finally, the Korea Spallation Neutron Source (KSNS), which is part of the upgrade plan for the 100-MeV proton linac to expand its utilization fields, is discussed.

Effects of Boron Neutron Capture Therapy on the Growth of Subcutaneous Xenografts of Human Colorectal Adenocarcinoma SW-620 in Immunodeficient Mice.

Boron neutron capture therapy (BNCT) can become an instrument for patients with malignant neoplasms of the rectum and colon. Here we evaluate the effectiveness of BNCT performed at the accelerator based epithermal neutron source at G. I. Budker Institute of Nuclear Physics, Siberian Division of Russian Academy of Sciences, in relation to subcutaneous xenografts of human colon adenocarcinoma SW-620 in SCID mice. Utilization of BNCT with boronоphenylalanine (BPA) and sodium borocaptate (BSH), which were injected intravenously into the retroorbital sinus, resulted in a significant decrease in tumor volumes compared to the control group (no radiation).

Feasibility study on the use of 230 MeV proton cyclotron in proton therapy centers as a spallation neutron source for BNCT.

AIM: The feasibility of using 230 MeV proton cyclotrons in proton therapy centers as a spallation neutron source for Boron Neutron Capture Therapy (BNCT) was investigated. BACKGROUND: BNCT is based on the neutron irradiation of a 10B-containing compound located selectively in tumor cells. Among various types of neutron generators, the spallation neutron source is a unique way to generate high-energy and high-flux neutrons. MATERIALS AND METHODS: Neutron beam was generated by a proton accelerator via spallation reactions and then the produced neutron beam was shaped to be appropriate for BNCT. The proposed Beam Shaping Assembly (BSA) consists of different moderators, a reflector, a collimator, as well as thermal and gamma filters. In addition, the simulated Snyder head phantom was utilized to evaluate the dose distribution in tumor and normal tissue due to the irradiation by the designed beam. MCNPX2.6 Monte Carlo code was used to optimize BSA as well as evaluate dose evaluation. RESULTS: A BSA was designed. With the BSA configuration and a beam current of 104 nA, epithermal neutron flux of 3.94 × 106 [n/cm2] can be achieved, which is very low. Provided that we use the beam current of 5.75 µA, epithermal neutron flux of 2.18 × 108 [n/cm2] can be obtained and the maximum dose of 38.2 Gy-eq can be delivered to tumor tissue at 1.4 cm from the phantom surface. CONCLUSIONS: Results for 230 MeV protons show that with proposed BSA, proton beam current about 5.75 µA is required for this purpose.

MTS-6 detectors calibration by using 239Pu-Be neutron source.

BACKGROUND: Thermoluminescent detectors, type MTS-6, containing isotope 6Li (lithium) are sensitive in the range of thermal neutron energy; the 239Pu-Be (plutonium-and-beryllium) source emits neutrons in the energy range from 1 to 11 MeV. These seemingly contradictory elements may be combined by using the paraffin moderator, a determined density of thermal neutrons in the paraffin block and a conversion coefficient neutron flux to kerma, not forgetting the simultaneous registration of the photon radiation inseparable from the companion neutron radiation. The main aim of this work is to present the idea of calibration of thermoluminescent detectors that consist of a 6Li isotope, by using 239Pu-Be neutron radiation source. MATERIAL AND METHODS: In this work, MTS-6 and MTS-7 thermoluminescent detectors and a plutonium-and-beryllium (239Pu-Be) neutron source were used. Paraffin wax fills the block, acting as a moderator. The calibration idea was based on the determination of dose equivalent rate based on the average kerma rate calculated taking into account the empirically determined function describing the density of thermal neutron flux in the paraffin block and a conversion coefficient neutron flux to kerma. RESULTS: The calculated value of the thermal neutron flux density was 1817.5 neutrons/cm2/s and the average value of kerma rate determined on this basis amounted to 244 µGy/h, and the dose equivalent rate 610 µSv/h. The calculated value allowed for the assessment of the length of time of exposure of the detectors directly in the paraffin block. CONCLUSIONS: The calibration coefficient for the used batch of detectors is (6.80±0.42)×10-7 Sv/impulse. Med Pr 2017;68(6):705-710.

Calculation of response parameters for a neutron long counter instrument.

The primary goal of this study was to develop a simulation model of a long counter available at Canadian Nuclear Laboratories (CNL). Using the Monte Carlo N-Particle version 6 (MCNP6) code, the model was used to calculate, as a function of incident energy, the number of counts recorded per source neutron, effective centre, and sensitivity. This study also carried out measurements of the neutron emission rate of and direct neutron flux at 2 m from an in-house 252Cf neutron source.

Pencil beam kernel-based dose calculations on CT data for a mixed neutron-gamma fission field applying tissue correction factors.

Objective.For fast neutron therapy with mixed neutron and gamma radiation at the fission neutron therapy facility MEDAPP at the research reactor FRM II in Garching, no clinical dose calculation software was available in the past. Here, we present a customized solution for research purposes to overcome this lack of three-dimensional dose calculation.Approach.The applied dose calculation method is based on two sets of decomposed pencil beam kernels for neutron and gamma radiation. The decomposition was performed using measured output factors and simulated depth dose curves and beam profiles in water as reference medium. While measurements were performed by applying the two-chamber dosimetry method, simulated data was generated using the Monte Carlo code MCNP. For the calculation of neutron dose deposition on CT data, tissue-specific correction factors were generated for soft tissue, bone, and lung tissue for the MEDAPP neutron spectrum. The pencil beam calculations were evaluated with reference to Monte Carlo calculations regarding accuracy and time efficiency.Main results.In water, dose distributions calculated using the pencil beam approach reproduced the input from Monte Carlo simulations. For heterogeneous media, an assessment of the tissue-specific correction factors with reference to Monte Carlo simulations for different tissue configurations showed promising results. Especially for scenarios where no lung tissue is present, the dose calculation could be highly improved by the applied correction method.Significance.With the presented approach, time-efficient dose calculations on CT data and treatment plan evaluations for research purposes are now available for MEDAPP.

The spectrum-averaged cross section investigation of 117Sn(n,n')117mSn and 67Zn(n,p)67Cu reactions.

The spectrum averaged cross sections (SACS) in standard neutron field, e.g. 252Cf(s.f.), is a preferable tool for cross section evaluation and validation. A set of reaction measurements with high energy thresholds was previously performed. The presented work focuses on lower energy threshold reactions, namely on the inelastic scattering of the tin foil, more specifically the reaction 117Sn(n,n')117mSn, and the zinc foil reaction, namely 67Zn(n,p)67Cu. These reactions are of special interest due to their intermediate energy range, which is essential in classical reactor dosimetry and fast reactor dosimetry. The experiments were carried out in a standard neutron field formed by 252Cf(s.f.) source in Rez. The experimental results were compared with calculations using MCNP6.2, ENDF/B-VII.1 transport library, and ENDF/B-VIII.0 and IRDFF-II cross section data library. Additionally, the calculations using CEA code DARWIN/PEPIN2 using JEFF-3.0/A were executed. The obtained experimental SACS of previously measured reactions were in good agreement with the SACS calculated using the IRDFF-II library. Additionally, the calculational reaction rate of 67Zn(n,p)67Cu was in accordance with the experimental data in case of ENDF/B-VIII.0 nuclear data library. Moreover, the calculational results of 117Sn(n,n')117mSn obtained by DARWIN/PEPIN2 code (using JEFF-3.0/A nuclear data library) are closest to the experimental results.

Design of a mixed material moderator in a beam-shaping assembly for proton accelerator-based boron neutron capture therapy.

Boron Neutron Capture Therapy is being promoted with the development of accelerator neutron sources, and many new accelerator-based BNCT facilities are being built. In Particle Accelerator Facility project of Sun Yat-sen University, we plan to build a terminal for BNCT research based on an 8 MeV, CW 3 mA proton accelerator. In this paper, we present a beam-shaping assembly for this proton accelerator with such low 24 kW beam power, using composite moderator materials composed of five elements: Mg, Al, F, O, and Li. The calculation result of FLUKA with ENDF/B and JENDL libraries shows that the epithermal neutron beam flux is 1.57×109n/cm2/s with the CW 3 mA proton beam. The fast neutron component and the gamma ray component under free-air condition are 1.49×10-13Gy∙cm2 and 8.12×10-14Gy∙cm2 respectively, in line with IAEA-TECDOC-1223 design recommendations. The thermal analysis shows that the maximum temperature of beryllium target is 706.5 K, and the structure materials of BSA do not melt.

Effects of neutron sources and moderator materials on therapeutic dose delivery to deep-seated lesions in proton accelerator-based boron neutron capture therapy.

This study aimed to identify the optimal conditions for delivering sufficient doses to deep-seated lesions within short irradiation times for two boron carriers of different T/N ratios. The therapeutic depth and irradiation time of a neutron beam for beam shaping assemblies (BSAs) with a Li or Be target and a MgF2 or CaF2 moderator were examined with the fast-neutron dose per epithermal neutron (FNR) as a parameter. When T/N = 3.61, the therapeutic depth was almost saturated at an FNR of about 10 × 10-13 Gy cm2; when the FNR value was about 10 × 10-13 Gy cm2, the therapeutic depth of the neutron beam for the BSA with a Be target and a MgF2 moderator was almost identical to that for the neutron beam for the BSA with a Be target and a CaF2 moderator, and slightly greater than those for the neutron beams for the BSAs with a Li target and a MgF2 or CaF2 moderator; moreover, the irradiation time of the neutron beam for the BSA with a Be target and a MgF2 moderator was shorter than that for the neutron beam for the BSA with a Be target and a CaF2 moderator. When T/N = 100, the therapeutic depths of the neutron beams for the BSAs varied greatly depending on the FNR, and were greater than the corresponding values for T/N = 3.61. We therefore concluded that the BSA with a Be target and a MgF2 moderator that produced a neutron beam with an FNR of about 10 × 10-13 Gy cm2 is optimal for delivering sufficient doses to deep-seated lesions in short irradiation times when T/N = 3.61, and stricter control over FNR is required when T/N = 100.

Measurement of Isfahan heavy water zero-power reactor kinetic parameters using advanced pulsed neutron source method in a near critical state.

The pulsed neutron source (PNS) technique was used to determine the prompt neutron decay constant for two different lattice pitches in the HWZPR heavy water zero power reactor. The results were compared to the variance-to-mean ratio (VTM) method. The neutron mean generation time was also calculated for both pitches, and the results were compared to previous Monte Carlo calculations. The findings of this research can be used as a benchmark nuclear codes to validate kinetic parameters.

Fast neutron induced reaction cross sections on natural manganese and tantalum.

The cross sections for the 55Mn(n,2n)54Mn, 181Ta(n,2n)180gTa, and 181Ta(n,p)181Hf reactions were measured to be 705.1 ± 26.1 mb at 14.0 MeV, 1362.7 ± 87.2 mb at 13.6 MeV, and 2.31 ± 0.09 mb at 13.6 MeV, respectively, by using an off-line γ-ray spectroscopic technique. The neutrons were produced via the 3H(d,n)4He reaction. The monitor reactions 27Al(n,α)24Na and 93Nb(n,2n)92mNb were used for neutron flux determination. The results from the present work were compared with those of the literature and the evaluated data from ENDF/B-VIII.0, JEFF-3.3, JENDL-5, CENDL-3.2, and BROND-3.1 libraries. Besides, the cross sections were also estimated with the TALYS-1.96 nuclear model code using different level density models for a better description of the present work and literature data. The present experimental results were found to be in good agreement with most of the available literature data and with the evaluated data.

Excitation functions for fast neutron induced reactions on iron and lead.

Cross sections of the 54Fe(n,p)54Mn, 54Fe(n,α)51Cr, 56Fe(n,p)56Mn and 204Pb (n,2n)203Pb reactions induced by D-T neutrons were obtained with activation method and γ-ray spectrometry technique. Experimental values measured in this work are consistent with most of the previous literature data. These reactions cross sections were theoretically calculated by using the TALYS-1.96 and EMPIRE-3.2.3 codes from threshold up to 20 MeV, and significant discrepancies were found between calculated results and experiment data. In addition, experimental values are compared with evaluated nuclear data of the CENDL-3.2, ENDF/B-VIII.0, JENDL-5, BROND-3.1 and JEFF-3.3 libraries, and significant difference was found for the 54Fe(n,α)51Cr reaction in ENDF/B-VIII.0 library but not for other reactions.

Energy measurement of 241Am-9Be neutrons by tagged neutron time-of-flight.

Neutron energy spectrum of a 241Am-9Be source was measured in the energy range of 0.3 to ∼ 6.5 MeV using γ-ray tagged neutron time-of-flight method. BC501A type organic liquid scintillator detector was used at a flight path of 175 cm to detect neutrons with good energy resolution. The de-excitation γ-ray emitted in coincidence with neutrons was detected using a fast BaF2 detector. The measured data has been compared with the ISO 8529-2 standard neutron reference spectrum and found good agreement in the overlapping energy region. Present measurement applied efficiency correction to the data and extended up to 0.3 MeV in the lower energy region compared to earlier reported measurement using similar technique.

A Review of Planned, Ongoing Clinical Studies and Recent Development of BNCT in Mainland of China.

Boron neutron capture therapy (BNCT) is a promising cancer treatment modality that combines targeted boron agents and neutron irradiation to selectively destroy tumor cells. In mainland China, the clinical implementation of BNCT has made certain progress, primarily driven by the development of compact neutron source devices. The availability, ease of operation, and cost-effectiveness offered by these compact neutron sources make BNCT more accessible to cancer treatment centers. Two compact neutron sources, one being miniature reactor-based (IHNI-1) and the other one being accelerator-based (NeuPex), have entered the clinical research phase and are planned for medical device registration. Moreover, several accelerator-based neutron source devices employing different technical routes are currently under construction, further expanding the options for BNCT implementation. In addition, the development of compact neutron sources serves as an experimental platform for advancing the development of new boron agents. Several research teams are actively involved in the development of boron agents. Various types of third-generation boron agents have been tested and studied in vitro and in vivo. Compared to other radiotherapy therapies, BNCT in mainland China still faces specific challenges due to its limited clinical trial data and its technical support in a wide range of professional fields. To facilitate the widespread adoption of BNCT, it is crucial to establish relevant technical standards for neutron devices, boron agents, and treatment protocols.

Accurate activity determination of a californium neutron source.

Accurately counting analog events requires constructing an electronics chain that produces one count for each input pulse. In this work we review the use of Nuclear Instrumentation Module electronic units for counting neutron capture events in a 3He tube. We identify two unique types of false trigger events in a leading-edge discriminator and show how a dual timer module can be used to produce a veto window to exclude these events. We use the constructed electronics chain to build an apparatus to measure neutron pulses from a 252Cf neutron source. We compare the measurements with a Monte Carlo N-Particle (MCNP) model to determine the activity of the neutron source. Furthermore, by making additional measurements with borated polyethylene attenuators between the source and detector, we are able to determine the boron concentration of the polyethylene. This technique provides accurate determination of the source activity to a precision of 2.8% at the k = 1 level. The method used is simple, inexpensive, and requires no additional calibrated instruments.

Initial evaluation of accelerator-based neutron source system at the Shonan Kamakura General Hospital.

An accelerator-based boron neutron capture therapy (AB-BNCT) system was installed at the Shonan Kamakura General Hospital (SKGH). We confirmed that a stable operation was possible for 1 h at a current of 30 mA. The evaluated thermal neutron flux was 2.8 × 109 cm-2 s-1 and in good agreement (±5%) with the calculated values. The daily variation was within ±2%. The ambient dose rate due to residual radioactivity after irradiation was approximately 5 µSv/h using a lead shutter.

Operational data for fault prognosis in particle accelerators with machine learning.

This paper presents real operational data collected from the power systems of the Spallation Neutron Source facility, which provides the most intense neutron beam in the world. The authors have used a radio-frequency test facility (RFTF) and simulated system failures in the lab without causing a catastrophic system failure. Waveform signals have been collected from the RFTF normal operation as well as during fault induction efforts. The dataset provides a significant amount of normal and faulty signals for the training of statistical or machine learning models. Then, the authors performed 21 test experiments, where the faults are slowly induced into the RFTF system for the purpose of testing the models in fault prognosis to detect and prevent impending faults. The test experiments include interesting combinations of magnetic flux compensation and start pulse width adjustments, which cause gradual deterioration in the waveforms (e.g., system output voltage, system output current, insulated-gate bipolar transistor currents, magnetic fluxes), which mimic the fault scenarios. Accordingly, this dataset can be valuable for developing models to predict impending fault scenarios in power systems in general and in particle accelerators in specific. All experiments occurred in the Spallation Neutron Source facility of Oak Ridge National Laboratory in Oak Ridge, Tennessee of the United States in July 2022.

Measurement of Cumulative fission product yields on 235U induced by 2.8 MeV neutrons.

Off-line gamma-ray spectrometry was used to accurately measure the Cumulative fission product yields (CFPYs) of fission products in the 235U (n, f) reaction induced by 2.8 MeV neutrons. The 2.8 MeV quasi-monoenergetic neutron beam was produced by the CPNG-600 Cockcroft Walton accelerator at the China Institute of Atomic Energy （CIAE）and the gamma spectra were measured by the HPGe γ-ray Spectrometer. After fully considering and revising the sources of uncertainty, high-precision CFPYs of 4 fission products were obtained. This study has important applications in reactor design and operation and is conducive to the establishment of an evaluated nuclear database.