Study on the optimal incident proton energy of ^(7)Li(p,n)^(7)Be neutron source for boron neutron capture therapy

Boron neutron capture therapy(BNCT)is recognized as a precise binary targeted radiotherapy technique that effectively eliminates tumors through the^(10)B(n,α)^(7)Li nuclear reaction.Among various neutron sources,accelerator-based sources have emerged as particularly promising for BNCT applications.The^(7)Li(p,n)^(7)Be reaction is highly regarded as a potential neutron source for BNCT,owing to its low threshold energy for the reaction,significant neutron yield,appropriate average neutron energy,and additional benefits.This study utilized Monte Carlo simulations to model the physical interactions within a lithium target subjected to proton bombardment,including neutron moderation by an MgF_(2)moderator and subsequent BNCT dose analysis using a Snyder head phantom.The study focused on calculating the yields of epithermal neutrons for various incident proton energies,finding an optimal energy at 2.7 MeV.Furthermore,the Snyder head phantom was employed in dose simulations to validate the effectiveness of this specific incident energy when utilizing a^(7)Li(p,n)^(7)Be neutron source for BNCT purposes.

Coin-structured tunable beam shaping assembly design for accelerator-based boron neutron capture therapy for tumors at different depths and sizes

In the past decade,boron neutron capture therapy utilizing an accelerator-based neutron source(ABNS)designed primarily for producing epithermal neutrons has been implemented in the treatment of brain tumors and other cancers.The specifications for designing an epithermal beam are primarily based on the IAEA-TECODC-1223 report,issued in 2001 for reactor neutron sources.Based on this report,the latest perspectives and clinical requirements,we designed an ABNS capable of adjusting the average neutron beam energy.The design was based on a 2.8 MeV,20 mA proton beam bombarding a lithium target to produce neutrons that were subsequently moderated and tuned through a tunable beam shaping assembly(BSA)which can modify the thicknesses and materials of the coin-shaped moderators,back reflectors,filters,and collimators.The simulation results demonstrated that epithermal neutron beams for deep seated tumor treatment,which were generated by utilizing magnesium fluoride with lengths ranging between 28 and 36 cm as the moderator,possessed a treatment depth of 5.6 cm although the neutron flux peak shifts from 4.5 to 1.0 keV.When utilizing a thinner moderator,a less accelerated beam power can meet the treatment requirements.However,higher powers reduced the treatment time.In contrast,employing a thick moderator can reduce the skin dose.In scenarios that required relatively low energy neutron beams,the removal of the thermal neutron filter can raise the thermal neutron flux at the beam port.And the depth of the dose rate peak could be adjusted between 0.25 and 2.20 cm by combining magnesium fluoride and polyethylene coins of different thicknesses.Hence,this device has a better adaptability for the treatment of superficial tumors.Overall,the tunable BSA provides greater flexibility for clinical treatment than common BSA designs that can only adjust the port size.

The study of a neutron spectrum unfolding method based on particle swarm optimization combined with maximum likelihood expectation maximization

The neutron spectrum unfolding by Bonner sphere spectrometer(BSS) is considered a complex multidimensional model,which requires complex mathematical methods to solve the first kind of Fredholm integral equation. In order to solve the problem of the maximum likelihood expectation maximization(MLEM) algorithm which is easy to suffer the pitfalls of local optima and the particle swarm optimization(PSO) algorithm which is easy to get unreasonable flight direction and step length of particles, which leads to the invalid iteration and affect efficiency and accuracy, an improved PSO-MLEM algorithm, combined of PSO and MLEM algorithm, is proposed for neutron spectrum unfolding. The dynamic acceleration factor is used to balance the ability of global and local search, and improves the convergence speed and accuracy of the algorithm. Firstly, the Monte Carlo method was used to simulated the BSS to obtain the response function and count rates of BSS. In the simulation of count rate, four reference spectra from the IAEA Technical Report Series No. 403 were used as input parameters of the Monte Carlo method. The PSO-MLEM algorithm was used to unfold the neutron spectrum of the simulated data and was verified by the difference of the unfolded spectrum to the reference spectrum. Finally, the 252Cf neutron source was measured by BSS, and the PSO-MLEM algorithm was used to unfold the experimental neutron spectrum.Compared with maximum entropy deconvolution(MAXED), PSO and MLEM algorithm, the PSO-MLEM algorithm has fewer parameters and automatically adjusts the dynamic acceleration factor to solve the problem of local optima. The convergence speed of the PSO-MLEM algorithm is 1.4 times and 3.1 times that of the MLEM and PSO algorithms. Compared with PSO, MLEM and MAXED, the correlation coefficients of PSO-MLEM algorithm are increased by 33.1%, 33.5% and 1.9%, and the relative mean errors are decreased by 98.2%, 97.8% and 67.4%.

High-resolution neutronics model for ^(238)Pu production in high-flux reactors

We proposed and compared three methods(filter burnup,single energy burnup,and burnup extremum analysis)to build a high-resolution neutronics model for 238Pu production in high-flux reactors.The filter burnup and single energy burnup methods have no theoretical approximation and can achieve a spectrum resolution of up to~1 eV,thereby constructing the importance curve and yield curve of the full energy range.The burnup extreme analysis method combines the importance and yield curves to consider the influence of irradiation time on production efficiency,thereby constructing extreme curves.The three curves,which quantify the transmutation rate of the nuclei in each energy region,are of physical significance because they have similar distributions.A high-resolution neutronics model for ^(238)Pu production was established based on these three curves,and its universality and feasibility were proven.The neutronics model can guide the neutron spectrum optimization and improve the yield of ^(238)Pu by up to 18.81%.The neutronics model revealed the law of nuclei transmutation in all energy regions with high spectrum resolution,thus providing theoretical support for high-flux reactor design and irradiation production of ^(238)Pu.

Research on perturbation of neutron fluence rate in a closed thermal neutron field due to medium materials

The neutron radiation field has vital applications in areas such as biomedicine,geology,radiation safety,and many others for neutron detection and neutron metrology.Correcting neutron fluence rate perturbation accurately is an important yet challenging problem.This study proposes a correction method that analyzes three physical processes.This method,which transforms the detection process from point detection to area detection,is based on a novel physical model and has been validated through theoretical analyses,experiments,and simulations.According to the average differences between the calculated and experimental results,the new method(1.67%)demonstrated better accuracy than the traditional simulation(2.17%).In a closed thermal neutron radiation field,the detector or strong neutron absorption material significantly perturbs the neutron fluence rate,whereas its impact on the energy spectrum shape and neutron directionality is relatively minor.Furthermore,based on the calculation results of the perturbation rate formula for medium materials with different compositions and sizes,the larger the volume and capture cross section of the medium,the higher the perturbation rate generated in the closed radiation field.

Standard spectrum measurement and simulation of elemental capture spectroscopy log

We have established a set of laboratory measurements which is used for capturing element gammma spectrum. Standard captured gamma ray spectra for ten elements, including Si, Ca, Fe, are obtained using the measurements for the first time in China. We also simulated the capture gamma ray spectra of the ten elements using Monte Carlo methodology with the same parameters of our experimental measurements. Comparing the experiment and simulation results with the data from the International Atomic Energy Agency＇s Nuclear Data Center, we obtained the standard captured gamma ray spectra of the ten elements, which, as calibration spectra, are used to calibrate the raw spectrum in data processing. This method solved the key problem during the conversion from the original measuring spectrum to the yield of each element in the data processing. The method can effectively improve the accuracy of the element yield calculation.

Calculating standard captured γ spectra of formation elements

Multi-element gamma-ray spectrum logging requires standard spectra of individual elements for its interpretation.Since the standard element spectra are usually derived using Monte Carlo simulation of either a pure element or its oxide,we synthesized the standard neutron captured γ spectra（as would be observed using a NaI（Tl） detector） of elements H,Si,Ca and Fe from each element and its oxide.To compare the standard spectra from the elements and oxides,we operated three simulations of sandstone,limestone and mixed formation of sandstone and limestone each with ten different porosities,and used the two kinds of standard spectra to analyze the mixed spectra modeled from sandstone and limestone formations.The results show that the standard element spectra from oxides have more prominent energy peaks than the standard spectra from pure elements.The calculated formation element contents are close to the theoretical values when the standard element spectra from oxides are used to analyze the formation mixed spectra.Therefore,the formation element standard spectra should be calculated from oxide models in the analysis of neutron captured γ spectra by logging tools.

Boron neutron capture therapy for malignant melanoma： first clinical case report in China

A phase Ⅰ/Ⅱ clinical trial for treating malignant melanoma by boron neutron capture therapy（BNCT） was designed to evaluate whether the world＇s first in-hospital neutron irradiator（IHNI） was qualified for BNCT. In this clinical trial planning to enroll 30 patients, the first case was treated on August 19, 2014. We present the protocol of this clinical trial, the treating procedure, and the clinical outcome of this first case. Only grade 2 acute radiation injury was observed during the first four weeks after BNCT and the injury healed after treatment. No late radiation injury was found during the 24-month follow-up. Based on positron emission tomography-computed tomography（PET/CT） scan, pathological analysis and gross examination, the patient showed a complete response to BNCT,indicating that BNCT is a potent therapy against malignant melanoma and IHNI has the potential to enable the delivery of BNCT in hospitals.

Neutron Capture Therapy (NCT) & In-Hospital Neutron Irradiator (IHNI)——a new technology on binary targeting radiation therapy of cancer

BNCT is finally becoming ＂a new option against cancer＂. The difficulties for its development progress of that firstly is to improve the performance of boron compounds, secondly, it is the requirements of quantification and accuracy upon radiation dosimetry evaluation in clinical trials. Furthermore, that is long anticipation on hospital base neutron sources. It includes dedicated new NCT reactor, accelerator based neutron sources, and isotope source facilities. In ad- dition to reactors, so far, the technology of other types of sources for clinical trials is not yet completely proven. The In- Hospital Neutron lrradiator specially designed for NCT, based on the MNSR successfully developed by China, can be installed inside or near the hospital and operated directly by doctors. The Irradiator has two neutron beams for respective treatment of the shallow and deep tumors. It is expected to initiate operation in the end of this year. It would provide a safe, low cost, and effective treatment tool for the NCT routine application in near future.

Study of neutron production and moderation for sulfur neutron capture therapy

Neutron capture therapy with Sulfur-33, similar to boron neutron capture therapy with Boron-10, is effective in treating some types of tumors including ocular melanoma. The key point in sulfur neutron capture therapy is whether the neutron beam flux and the resonance capture cross section of ^(33)S(n;α)^(30) Si reaction at 13.5 keV can achieve the requirements of radiotherapy. In this research,the authors investigated the production of 13.5 keV neutron production and moderation based on an accelerator neutron source. A lithium glass detector was used to measure the neutron flux produced via near threshold^7 Li(p,n)~7 Be reaction using the time-of-flight method. Furthermore, the moderation effects of different kinds of materials were investigated using Monte Carlo simulation.

Analysis of prompt fission neutron spectrum and multiplicity for ^(237)Np(n, f) in the frame of multi-modal Los Alamos model

The improved version of Los Alamos model with the multi-modal fission approach is used to analyse the prompt fission neutron spectrum and multiplicity for the neutron-induced fission of 237Np. The spectra of neutrons emitted from fragments for the three most dominant fission modes （standard Ⅰ, standard Ⅱ and superlong） are calculated separately and the total spectrum is synthesized. The multi-modal parameters contained in the spectrum model are determined on the basis of experimental data of fission fragment mass distributions. The calculated total prompt fission neutron spectrum and multiplicity are better agreement with the experimental data than those obtained from the conventional treatment of the Los Alamos model.

Comparison of neutron energy spectrum unfolding methods and evaluation of rationality criteria

The neutron energy spectrum was measured using a Bonner sphere spectrometer at six locations inside the containment vessel of a nuclear reactor at the Qinshan nuclear power plant. The structures of the neutron spectra obtained by the maximum entropy, iteration, and genetic algorithm methods were consistent with one another and could be interpreted as the spectral superposition of different energy regions. The characteristic parameters of the neutron spectrum, including the fluence rate,average energy, and neutron ambient dose equivalent rate H^(*)(10), were in good agreement among the three methods. In addition, an LB6411 neutron ambient dose equivalent meter was employed to obtain the H^(*)(10) directly for comparison.These findings indicate that neutron spectrum unfolding methods can be used to overcome the problems associated with the response functions of dosimeters to provide more accurate H^(*)(10) values. In this study, the following three evaluation criteria were systematically addressed to ensure the accuracy of the unfolded spectra: count rates of the inverse solutions,neutron spectrum structures, and comparison of key parameters.

Geant4 simulation of multi-sphere spectrometer response function and the detection of 241Am–Be neutron spectrum

This paper is aimed at detecting the neutron spectrum of^(241)Am–Be, a widely used neutron source, with the SP9 ~3He proportional counter, which is a multi-sphere spectrometer system of eight thermal neutron detectors embedded in eight polyethylene(PE) spheres of varying diameters. The transport processes of a neutron in the multi-sphere spectrometer are simulated using the Geant4 code. Two sets of response functions of the PE spheres are obtained for calculating the^(241)Am–Be neutron spectrum.Response Function 1 utilizes the thermal neutron scattering model G4 Neutron HPThermal Scattering for neutron energies of ≤4 eV, and Response Function 2 has no thermal treatment. Neutron spectra of an^(241)Am–Be neutron source are measured and compared to those calculated by using the response functions. The results show that response function with thermal treatment is more accurate and closer to the real spectrum.

Deterministic Parsing Model of the Compound Biological Effectiveness (<i>CBE</i>) Factor for Intracellular ¹⁰Boron Distribution in Boron Neutron Capture Therapy

Purpose: In defining the biological effects of the 10B(n, α)7Li neutron capture reaction, we have previously developed a deterministic parsing model to determine the Compound Biological Effectiveness (CBE) factor in Borono-Phenyl-Alanine (BPA)-mediated Boron Neutron Capture Therapy (BNCT). In present paper, we demonstrate that the CBE factor is directly and unambiguously derivable by the new formula for any case of intracellular 10Boron (10B) distribution, which is founded on this model for tissues and tumor. Method: To determine the CBE factor, we derive the following new calculation formula founded on the deterministic parsing model with three constants, CBE0, F, n and the eigen value Nth/Nmax. where, Nth and Nmax are the threshold value of boron concentration of N and saturation boron density in tissues and tumor. In order to determine these constants and the eigen values, iterative calculation technique was employed for the CEB factor and Nmax data set previously reported. Results and Conclusion: From the iterative calculation results, it is clear that the calculated CBE factor values obtained are almost identical to the original CBE factors and there is a good correlation between the original CBE factors and Nth/Nmax, when CBE0, F and n are given as 0.5, 8 and 3, respectively. These constants provide a better understanding of different types of intracellular10B distribution.

Measurements of the ^(107)Ag neutron capture cross sections with pulse height weighting technique at the CSNS Back-n facility

Silver indium cadmium(Ag–In–Cd) control rod is widely used in pressurized water reactor nuclear power plants,and it is continuously consumed in a high neutron flux environment. The mass ratio of ^(107)Ag in the Ag–In–Cd control rod is 41.44%. To accurately calculate the consumption value of the control rod, a reliable neutron reaction cross section of the ^(107)Ag is required. Meanwhile,^(107)Ag is also an important weak r nucleus. Thus, the cross sections for neutron induced interactions with ^(107)Ag are very important both in nuclear energy and nuclear astrophysics. The(n, γ) cross section of ^(107)Ag has been measured in the energy range of 1–60 eV using a back streaming white neutron beam line at China spallation neutron source. The resonance parameters are extracted by an R-matrix code. All the cross section of ^(107)Ag and resonance parameters are given in this paper as datasets. The datasets are openly available at http://www.doi.org/10.11922/sciencedb.j00113.00010.

Boron neutron capture therapy: moving towards targeted therapy for locally recurrent head and neck squamous cell carcinoma

Locally recurrent head and neck squamous cell carcinoma(HNSCC)is often unresectable,and a repeat course of radiotherapy is associated with incremental toxicities.Boron neutron capture therapy(BNCT)is a novel targeted radiotherapy modality that can achieve a high dose gradient between cancerous and adjacent normal tissues.However,the relationships among the dose resulting from BNCT,tumor response to BNCT,and survival are not completely understood.Recently,a study published in Radiotherapy and Oncology investigated the efficacy of BNCT in the treatment of patients with locally recurrent HNSCC and the factors associated with favorable treatment response and survival.In this article,the findings,strengths and limitations of this study are discussed in depth,and the significance of the study and motivations for future research are highlighted.

<i>In Vitro</i>Evaluation System of Pharmacokinetics and Irradiation Effect in Boron Neutron Capture Therapy (BNCT) Using Three-Dimensional Artificial Human Tumor Tissue Model

Boron neutron capture therapy (BNCT) is based on the incorporation of boron-containing drugs to cancer cells and the nuclear reaction of 10B atoms by thermal neutron irradiation results in tumor degeneration. For the development of this therapy, currently, long time and high cost consuming experiments using many animals are required. In this study, we constructed a new in vitro evaluation system for BNCT by combination of an artificial tumor tissue model, comprised of normal human dermal-derived fibroblast (NHDF) and human pancreatic cancer cell line BxPC3, and the optical plastic material CR-39 as a solid state nuclear track detector. Administration of boronophenylalanine (10BPA) as a boron-containing drug and neutron irradiation up to 2.52 × 1012 n/cm2 to the control tissue constructed by NHDF (NHDF3D) and BxPC3 cell loaded tissue (NHDF3D/BxPC3) resulted in detection of 1.6 times higher number of α-ray/recoiled Li particle tracks in NHDF3D/BxPC3 in comparison to NHDF3D, demonstrating that putative irradiation damage to cancer cells can be evaluated by this system. On a cellular level, the hit number of α-ray/recoiled Li particle tracks per single BxPC3 cells and NHDF was evaluated as 5.46 and 1.71, respectively. The tumor and normal tissue ratio (T/N ratio) was 3.19, which was corresponded with those of BPA as 2 - 4 that reported in the previous studies. This new in vitro evaluation system may provide a useful tool for a low cost, labor-saving, and non-animal method for the development of new boron-containing drugs or improvement of BNCT conditions.

Using activation method to measure neutron spectrum in an irradiation chamber of a research reactor

Neutron spectrum should be measured before test samples are irradiated.Neutron spectrum in an irradiation chamber of a research reactor was measured by using activation method when the reactor is in normal operation under 2 MW.Sixteen kinds of non-fission foils(19 reaction channels) were selected,of which 10 were sensitive to thermal and intermediate energy regions,while the others were of different threshold energy and sensitive to fast energy regions.By measuring the foil radioactivity,the neutron spectrum was unfolded with the iterative methods SAND-Ⅱ and MSIT.Finally,shielding corrections of group cross-section and main factors affecting the calculation accuracy were studied and the uncertainty of solution was analyzed using the Monte Carlo method in the process of SAND-Ⅱ.

Energy spectrum measurement and dose rate estimation of natural neutrons in Tibet region

In this work, natural neutron spectra at nine sites in Tibet region were measured using a multi-sphere neutron spectrometer. The altitude-dependence of the spectra total fluence rate and ambient dose equivalent rate were analyzed. From the normalized natural neutron spectra at different altitudes, the spectrum fractions for neutrons of greater than 0.1 MeV do not differ obviously, while those of the thermal neutrons differ greatly from each other. The total fluence rate, effective dose rate and the ambient dose equivalent rate varied with the altitude according to an exponential law.