Accelerator-based Neutron Source Research Articles

Research and Development of Accelerator Based Neutron Sources

Research and Development of Accelerator Based Neutron Sources

Neutron Source for Material and Component Tests by Using IFMIF/EVEDA Prototype Accelerator

A neutron source for material and component tests is an essential tool for the DEMO reactor development. An accelerator-based neutron source such as IFMIF is regarded as the most promising one in Japan and the EU. The construction plan of IFMIF is still open due to the influence of the large cost overrun of ITER procurements. Japan Atomic Energy Agency (JAEA) has a plan of a neutron source for material and component tests using an IFMIF/EVEDA prototype accelerator and a lithium test loop for the IFMIF target facility. Expected performances of three options; 9 MeV and upgrading to 26 or 40 MeV of deuteron beam, are discussed. At the back plate position of the target, 1.5, 14, and 25 dpa/fpy are expected for 9, 26, and 40 MeV case, respectively. The option of 40 MeV is desirable, however, the option of 26 MeV is acceptable for blanket functional tests and material tests.

Neutron intensity monitor with activation foil for p-Li neutron source for BNCT – Feasibility test of the concept

Proton-lithium (p-Li) reaction is being examined worldwide as a candidate nuclear production reaction for accelerator based neutron source (ABNS) for BNCT. In this reaction, the emitted neutron energy is not so high, below 1MeV, and especially in backward angles the energy is as low as about 100keV. The intensity measurement was thus known to be difficult so far. In the present study, a simple method was investigated to monitor the absolute neutron intensity of the p-Li neutron source by employing the foil activation method based on isomer production reactions in order to cover around several hundreds keV. As a result of numerical examination, it was found that 107Ag, 115In and 189Os would be feasible. Their features found out are summarized as follows: 107Ag: The most convenient foil, since the half life is short. 115In: The accuracy is the best at 0°, though it cannot be used for backward angles. And 189Os: Suitable nuclide which can be used in backward angles, though the gamma-ray energy is a little too low. These would be used for p-Li source monitoring depending on measuring purposes in real BNCT scenes.

Development of beryllium-based neutron target system with three-layer structure for accelerator-based neutron source for boron neutron capture therapy

The iBNCT project team with University of Tsukuba is developing an accelerator-based neutron source. Regarding neutron target material, our project has applied beryllium. To deal with large heat load and blistering of the target system, we developed a three-layer structure for the target system that includes a blistering mitigation material between the beryllium used as the neutron generator and the copper heat sink. The three materials were bonded through diffusion bonding using a hot isostatic pressing method. Based on several verifications, our project chose palladium as the intermediate layer. A prototype of the neutron target system was produced. We will verify that sufficient neutrons for BNCT treatment are generated by the device in the near future.

Additive effect of BPA and Gd-DTPA for application in accelerator-based neutron source

Because of its fast metabolism gadolinium as a commercial drug was not considered to be suitable for neutron capture therapy. We studied additive effect of gadolinium and boron co-administration using colony forming assay. As a result, the survival of tumor cells with additional 5ppm of Gd-DTPA decreased to 1/10 compared to the cells with boron only. Using gadolinium to increase the effect of BNCT instead of additional X-ray irradiation might be beneficial, as such combination complies with the short-time irradiation regimen at the accelerator-based neutron source.

Mock-up experiment at Birmingham University for BNCT project of Osaka University – Neutron flux measurement with gold foil

Mock-up experiment for development of accelerator based neutron source for Osaka University BNCT project was carried out at Birmingham University, UK. In this paper, spatial distribution of neutron flux intensity was evaluated by foil activation method. Validity of the design code system was confirmed by comparing measured gold foil activities with calculations. As a result, it was found that the epi-thermal neutron beam was well collimated by our neutron moderator assembly. Also, the design accuracy was evaluated to have less than 20% error.

Broad energy range neutron spectroscopy using a liquid scintillator and a proportional counter: Application to a neutron spectrum similar to that from an improvised nuclear device

A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100keV up to 9MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n)3He and D(d,n)3He, generating neutron energies in the range between 0.2 and 8MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9Be(d,n) reaction using a 5MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima.

Application of an ultraminiature thermal neutron monitor for irradiation field study of accelerator-based neutron capture therapy.

Phantom experiments to evaluate thermal neutron flux distribution were performed using the Scintillator with Optical Fiber (SOF) detector, which was developed as a thermal neutron monitor during boron neutron capture therapy (BNCT) irradiation. Compared with the gold wire activation method and Monte Carlo N-particle (MCNP) calculations, it was confirmed that the SOF detector is capable of measuring thermal neutron flux as low as 105 n/cm2/s with sufficient accuracy. The SOF detector will be useful for phantom experiments with BNCT neutron fields from low-current accelerator-based neutron sources.

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. The particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.

Absolute determination of the neutron source yield using melamine as a neutron detector

A new approach to absolute determination of the neutron source yield is presented. It bases on the application of melamine (C3H6N6) to neutron detection combined with Monte Carlo simulations of neutron transport. Melamine has the ability to detect neutrons via 14N(n, p)14C reaction and subsequent determination of 14C content. A cross section for this reaction is relatively high for thermal neutrons (1.827 b) and much lower for fast neutrons. A concentration of 14C nuclei created in the irradiated sample of melamine can be reliably measured with the aid of the accelerator mass spectrometry (AMS). The mass of melamine sufficient for this analysis is only 10 mg. Neutron detection is supported by Monte Carlo simulations of neutron transport carried out with the use of MCNP-4C code. These simulations are aimed at computing the probability of 14C creation in the melamine sample per the source neutron. The result of AMS measurements together with results of MCNP calculations enable us to determine the number of neutrons emitted from the source during the irradiation of melamine.The proposed method was applied for determining the neutron emission from a commercial 252Cf neutron source which was independently calibrated. The measured neutron emission agreed with the certified one within uncertainty limits. The relative expanded uncertainty (k=2) of the absolute neutron source yield determination was estimated at 2.6%.Apart from calibration of radionuclide neutron sources the proposed procedure could facilitate absolute yield measurements for more complex sources. Potential applications of this methodology as it is further developed include diagnostics of inertial confinement fusion and plasma-focus experiments, calibration of neutron measurement systems at tokamaks and accelerator-based neutron sources as well as characterization of neutron fields generated in large particle detectors during collisions of hadron beams.

Computational Study of Integrated Neutron/Photon Imaging for Illicit Material Detection

The feasibility of integration of photon and neutron radiography for nondestructive detection of illicit materials was examined. The MCNP5 code was used to model a radiography system consisting of accelerator-based neutron and photon sources and the imaging detector array, with an object under scrutiny placed between them. For this examination, the objects consisted of a matrix of low-Z and high-Z materials of various shapes and density. Transmission-radiography computations were carried out using 2.5-MeV deuterium-deuterium and 14-MeV deuterium-tritium neutron sources, and a 0.3-MeV photon source. The radiography tallies for both neutron and photon sources were modeled for the same geometry of the system. The photon-to- neutron transmission ratios were determined for each pixel of the detector array and utilized to identify the presence of specific materials in the radiographic images. By focusing on the inherent difference between neutron and photon interactions, it was possible to determine the shape and material composition of complex objects present within a pallet or a shipping container. The use of a single imaging array of scintillation detectors for simultaneous measurements of fast neutrons and photons is discussed, and its function in the dual neutron/photon radiography applications is addressed.

Application of Advanced Nuclear Emulsion Technique to Fusion Neutron Diagnostics

In order to measure the 2.5 MeV neutrons produced by DD nuclear fusion reactions, we have developed a compact neutron detector based on nuclear emulsion. After optimization of development conditions, we evaluated the response of the detector to an accelerator-based DD neutron source. The absolute efficiency at an energy of 2.5 MeV was estimated to be (4.1±0.2)×10-6 tracks/neutron.

Neutronic Analysis of the IFMIF Tritium Release Test Module Based on the EVEDA Design

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based intense neutron source to test fusion reactor materials under irradiation conditions expected to be experienced by a future fusion power plant (DEMO). The Tritium Release Test Module (TRTM) is intended for the irradiation of solid breeder ceramics as well as beryllium involving in-situ tritium release measurements in IFMIF. During the EVEDA (Engineering Validation Engineering Design Activities) phase, a detailed engineering design for the TRTM has been elaborated. A new 3-dimesional Monte Carlo geometry model of TRTM was prepared for a neutronic analysis directly from engineering CAD data using the McCad conversion software developed at KIT. The analysis was performed with the latest version of the Monte Carlo code McDeLicious, an enhancement to MCNP5 for IFMIF neutronics calculations, using a state-of-the-art nuclear data library FENDL-3. The result emphasizes the importance of the neutron reflector which should be placed behind TRTM in order to make the irradiation properties close to the European HCPB DEMO. Although the achievable dpa is lower than that expected in DEMO, the T/dpa and He/dpa values can be simulated very well when the neutron reflector is appropriately designed, in particularly by utilizing beryllium.

Engineering Validation and Engineering Design of Lithium Target Facility in IFMIF/EVEDA Project

EVEDA Lithium Test Loop (ELTL) has been designed and constructed, has operated a liquid lithium flow test facility with the world’s highest flow rate and has succeeded in generating a 100-mm-wide and 25-mm-thick free-surface lithium flow along a concave back plate steadily at a high speed of 20 m/s at 300°C for the first time in the world. This result will greatly advance the development of an accelerator-based neutron source to high energy and high density, one of the key objectives of the fusion reactor materials development under the BA (Broader Approach) Activities. Recent related engineering validation and engineering design of the lithium facility has been evaluated.

Development of a New Method for Measurement of Neutron Detector Efficiency up to 20 MeV

A new approach to neutron detector efficiency detemination has been taken. A neutron detector has been calibrated with a 252Cf source at low energy. The calibration can be extended to energies above 8 MeV with accelerator-based neutron sources. This techniques uses the fact that the cross section for a symmetric reaction with nucleus of atomic number A yielding a final nucleus with atomic number (2A−1) and a neutron A+A→(2A−1)+n. This reaction must be symmetric about 90∘ in the center-of-mass system. The laboratory energies for the neutrons at the paired energies differ substantially. Thus, an efficiency known at one of the two angles can be used to determine the efficiency to higher energies or, for a negative Q, to lower neutron energies.

Design of Neutron Reflectometer in the Angle-resolved Dispersive Mode for Compact Accelerator Based Neutron Sources

Compact accelerator-driven neutron sources take advantage of their readiness of operation and high adaptability to various experiments, although their neutron flux is low. We are now working on installing a reflectometer to Kyoto University Accelerator based Neutron Source (KUANS). Neutron reflectometer (NR) is one of the most simple but powerful spectrometer, in which neutron reflectivity from the surface of a sample is measured and analyzed to give so called ‘depth profile’ of the sample, which represents the distribution of materials along the surface normal direction. Usually, reflectometer requires narrow well-collimated beam. For the present reflectometer, however, incident neutron beam is narrowed at the sample position but not so much collimated to allow variety of incident angle to the sample simultaneously assuming the specular reflection at the sample surface. This multi-incident angle measurement helps to compensate weak intensity of KUANS. The motivation of this study is development of the NR with the readiness of operation for a preliminary experiment. In this study, the design of the NR is presented and the reflectivity is estimated via Monte-Carlo simulation. It was found that measurable lowest reflectivity is down to 10-2.

Boron Neutron Capture Therapy (BNCT) - Low-Energy Neutron Spectrometer for Neutron Field Characterization -

Boron Neutron Capture Therapy (BNCT) is known to be a promising and new cancer therapy which can kill tumor cells suppressing damage to normal tissues. Recently, accelerator based neutron sources (ABNS) for BNCT are under development especially in Japan. For characterization of the neutron field, we carried out the series study concerning thermal/epi-thermal neutron spectrometry especially for ABNS based BNCT. In the present study, we tested a commercially available position sensitive 3He proportional counter as the neutron detector combined with a measuring system using a multi-parameter MCA and a thermal/epi-thermal column designed and constructed for test measurements. From the results, the characteristics of the spectrometer were obtained through the basic experiments. Also, spectrum measurements were carried out with the thermal/epi-thermal neutron field. Agreement between the calculation and measurement was acceptable for both cases, showing that measurement would be possible for neutron energy from 0.01 eV to 10 keV. The spectrometer is hence expected to be applied to characterization measurement of the neutron field in the real BNCT scene in the future. c © 2014 The Japan Society of Plasma Science and Nuclear Fusion Research

Properties and Possible Applications of Kyoto University Accelerator Based Neutron Source (KUANS)

The results of preliminary experiments of Kyoto University Accelerator based Neutron Source (KUANS) is reported. The Time of Flight spectrum showing clear peak of thermal neutron is presented and neutron flux of 4000 n/cm2s at the 1.24 m from the moderator surface is evaluated. Preliminary results of time-resolved neutron radiography are also presented, which shows 10 minutes-exposure is enough to make clear images. Finally, some applications suitable for KUANS, such as reflectometer are suggested.

High-power electron beam tests of a liquid-lithium target and characterization study of 7Li(p,n) near-threshold neutrons for accelerator-based boron neutron capture therapy

A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center (SNRC). The target is intended to demonstrate liquid-lithium target capabilities to constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals. The lithium target will produce neutrons through the 7Li(p,n)7Be reaction and it will overcome the major problem of removing the thermal power >5kW generated by high-intensity proton beams, necessary for sufficient therapeutic neutron flux.In preliminary experiments liquid lithium was flown through the target loop and generated a stable jet on the concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power densities of more than 4kW/cm2 and volumetric power density around 2MW/cm3 at a lithium flow of ~4m/s, while maintaining stable temperature and vacuum conditions. These power densities correspond to a narrow (σ=~2mm) 1.91MeV, 3mA proton beam. A high-intensity proton beam irradiation (1.91–2.5MeV, 2mA) is being commissioned at the SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator.In order to determine the conditions of LiLiT proton irradiation for BNCT and to tailor the neutron energy spectrum, a characterization of near threshold (~1.91MeV) 7Li(p,n) neutrons is in progress based on Monte-Carlo (MCNP and Geant4) simulation and on low-intensity experiments with solid LiF targets. In-phantom dosimetry measurements are performed using special designed dosimeters based on CR-39 track detectors.

Measurement and estimation of isomeric cross section of 137Ba(n, n′)137mBa reaction using accelerator based neutron source

The isomeric cross section of 137Ba(n, n′)137mBa reaction over the neutron energy from 1keV to 4MeV has been calculated using TALYS-1.2 nuclear code. The variations of cross section with level densities and effective imaginary potentials have been studied to obtain the best fit with the excitation function. The reaction cross section for Barium was also measured in continuous neutron energy using the Microtron accelerator based neutron source. In this case, the bremsstrahlung radiation emitted by impinging 6MeV electrons on the e–γ primary target (tungsten) was allowed to fall on the γ–n secondary target (beryllium) to produce continuous neutrons and further barium was irradiated at 00 with respect to the incoming neutrons. It is observed that the experimental cross section for Barium is in good agreement with TALYS code and also evaluated data of EXFOR. Deviation factor has been determined for measuring and theoretical cross sections by TALYS-1.2 found to be around 14.71.