Neutron Source Model Research Articles

Neutron Activation Analysis Based on AB-BNCT Treatment Room.

Boron neutron capture therapy (BNCT) is an ideal binary targeted radiotherapy for treating refractory tumors. An accelerator-based BNCT (AB-BNCT) neutron source has attracted more and more attention due to its advantages such as higher neutron yield in the keV energy region, less gamma radiation, and higher safety. In addition to 10 B, neutrons also react with other elements in the treatment room during BNCT to produce many activation products. Due to the long half-life of some activation products, there will be residual radiation after the end of treatment and the shutdown of the accelerator, which has adverse effects on radiation workers. Therefore, the ambient dose equivalent rate in the treatment room needs to be evaluated. The AB-BNCT neutron source model proposed by Li is studied in this paper. Based on the Monte Carlo method, the Geant4 platform was used to simulate the dose induced by radionuclides near the Beam Shaping Assembly (BSA) of the source. It is concluded that the concrete wall contributed the most to the radiation dose. The dose rate of 2.45 μSv h -1 after 13 min of shutdown meets the dose rate limit of 2.5 μSv h -1 , at which point it is safe for workers to enter the treatment room area.

Effects of Neutron Source Models on Neutron Wall Loading in the EAST Tokamak

As a significant input for neutronics analysis of tokamak devices, the plasma neutron source is a bridge connecting fusion plasma physics and engineering design. A plasma neutron source model is established based on the EAST plasma configuration. The maximum neutron wall loading (NWL) is near the outboard midplane and is more than ~20% that of the inboard midplane. The investigations demonstrate that special care should be taken for radiation shielding and protection of the key components located on the outboard midplane and on the inboard midplane. The effect of the tritium accumulation on neutronics analysis should be carefully considered for EAST tokamak D-D plasma operations. The effect of density peaking (DP) on the neutronics analysis is also investigated. It is evident that the peak NWLs are all near the outboard midplane and that the poloidal distributions of the NWL are slightly different for these cases. With increasing DP, both the outboard and inboard peak NWLs decrease. However, the decrease in NWL is very small; NWL decreases only 11% when the neutron source peak increases about 1.5 times.

Evaluation of neutron dose rates at the TCV tokamak facility

A detailed MCNP neutronics model of the building housing the TCV tokamak has been developed. Predicted dose rates have been compared to neutron dose measurements at 12 locations in the building obtained during neutral beam injection into TCV plasmas. This work was motivated both by the necessity to understand neutron transport in a complex environment and to provide an adequate reference model of the building to which additional shielding can be added. The additional shielding is required because of an expected 5-10 fold increase of neutron doses by the addition of a second neutral beam source. The neutron source was estimated using the TRANSP code. Starting from a simple model, which left out or simplified many important structures, the model was iteratively refined. The discrepancies between the predicted and measured doses decreased significantly, down to less than a factor 2 for most of the locations calculated by the most detailed model. A factor 2 is believed to be within the combined errors resulting from uncertainties in the neutron source model, the detector calibration and the uncertainties stemming from the model itself.

Evaluation of the effectiveness of steel for shielding photoneutrons produced in medical linear accelerators: A Monte Carlo particle transport study

PurposeTo confirm the effectiveness of steel as a material for shielding photoneutrons produced in medical linear accelerators and report values for neutron first and second tenth value layer to be used as radiation protection quantities. MethodsMonte Carlo code Geant4 was used to simulate transport of photoneutrons through primary barriers containing layers of concrete and steel. The photoneutron spectrum generated in high energy modern accelerator head components was simulated and projected to barriers of different thicknesses of steel including a control case of no barrier. To derive the neutron tenth value layer (TVLn), absorbed dose was evaluated at 1 cm depth in a water phantom outside the barrier. The fluence and energy spectrum of neutrons leaking outside the barrier were also calculated. The neutron source model was used to quantitatively assess the neutron shielding for a newly designed bunker. ResultsFor an 18 MV photon beam produced by a linear accelerator, the first and second TVLn, for steel as a neutron shielding material were determined to be 15.7–16.1 cm and 28.5–32.4 cm, respectively, depending on the vendor. The lateral distribution of neutrons outside steel barriers is investigated. Also, we show the relative efficiency of steel for fast and epithermal neutrons and determine the neutron energy spectrum outside a steel barrier. ConclusionThis paper agrees with previous studies for Pu-Be sources, and we would advocate using these values for the purposes of high energy linear accelerator bunker design.

Application of local Monte Carlo method in neutronics calculation of EAST radial neutron camera

The Local Monte Carlo (LMC) method is used to solve the problems of deep penetration and long time in the neutronics calculation of the radial neutron camera (RNC) diagnostic system on the experimental advanced superconducting tokamak (EAST), and the radiation distribution of the RNC and the neutron flux at the detector positions of each channel are obtained. Compared with the results calculated by the global variance reduction method, it is shown that the LMC calculation is reliable within the reasonable error range. The calculation process of LMC is analyzed in detail, and the transport process of radiation particles is simulated in two steps. In the first step, an integrated neutronics model considering the complex window environment and a neutron source model based on EAST plasma shape are used to support the calculation. The particle information on the equivalent surface is analyzed to evaluate the rationality of settings of equivalent surface source and boundary. Based on the characteristic that only a local geometric model is needed in the second step, it is shown that the LMC is an advantageous calculation method for the nuclear shielding design of tokamak diagnostic systems.

Investigation of neutron source model based on EAST plasma shape

Plasma neutron source is an important input for neutronics analysis of tokamak devices, and also a bridge connecting plasma physics and fusion engineering design. In order to describe the spatial distribution of plasma neutron source more accurately in tokamak device, a plasma neutron source model based on EAST plasma shape (NSM-PS) is established. The model will help to improve the reliability of neutron transport. A neutron source generator (NSG) is developed to solve the polar distribution of the neutron emissivity based on NSM-PS, and a source card file is generated which can be used directly by Monte Carlo program. The difference between the neutron source model described by the magnetic surface approximation equation and NSM-PS are analyzed in terms of neutron emissivity, neutron spectrum and neutron wall loading. Finally, the full integration of the neutron emissivity calculated by NSG in the polar direction is compared with the neutron rate measured by the neutron flux monitor, which indirectly verifies the reliability of NSM-PS.

Deterministic method for calculating fusion neutron emission probability distribution in tokamak through mesh mapping

The high-fidelity 3D neutronics analysis of fusion systems has been enabled by advanced computer-aided design (CAD)-based simulation tools, and is being driven by an increasing desire to couple the analysis results with other physical field analyses. With increase in the level of output detail, the effects of the fusion neutron source spatial distribution have become increasingly important. The current fine-resolution neutron source model was established based on the Monte Carlo (MC) method, which is a time-consuming approach to reduce the deviations at low-intensity positions. This work presents a deterministic method to calculate the fusion neutron emission probability (FNEP) distribution in a tokamak machine with high resolution. In this deterministic method, the mesh-mapping technique is utilized to transfer neutron intensity data from an irregular (a, φ) mesh to a structured or unstructured (R, Z) mesh. The intensity data on (R, Z) mesh can be used directly by a deterministic neutronics code, or further be integrated and normalized in every grid to obtain the FNEP for an MC neutronics code. This deterministic scheme was implemented in the code EDITON, and the neutron source map can be generated subsequently by this code. EDITON exhibited high efficiency, and its accuracy was verified by comparing the obtained FNEP distribution and neutron wall loading (NWL) with the corresponding data from the TRANSGEN code, which is a tokamak fusion neutron spatial distribution calculation code based on the MC method.

Investigation of ortho↔para hydrogen conversion by collisions with neutrons

ABSTRACT Liquid hydrogen (LH) is constituted by para hydrogen (pH) and ortho hydrogen (oH) molecules. We analyzed the neutron scattering cross sections for pH→oH, pH→pH, oH→pH, and oH→oH reactions in LH. Results show that incident neutron energy lower than 14.7 meV is not sufficient to achieve pH→oH conversion. As energy increases, the cross section of pH→oH reaction increases sharply. If incident neutron energy is larger than 30 meV, branching fraction of pH molecules being converted into oH molecules upon inelastic scattering with the neutrons approaches 100%. As for the opposite process, branching fraction of oH→pH conversion is 10%–20% when incident neutron energy is lower than 30 meV. This conversion fraction oscillates with increasing incident neutron energy, eventually stabilizing at 30%. Based on the cross sections of the four reaction channels, we calculated the corresponding conversion rates based on a reactor cold neutron source model. There is 0.12% pH molecules of LH moderator would be converted to oH for one month operation . If neutron flux increased by 2 orders of magnitude, more than 10% pH would be transformed into oH, which means relative conversion rate of pH→oH induced by neutron inelastic scattering is not negligible.

Fusion Neutron Source Model for the Systems Analysis of a Tokamak Power Plant

In this work, a model has been developed to calculate the neutron wall loading poloidal distribution in a generic tokamak plasma and vessel geometry on the basis of Monte Carlo simulation. Different neutron source radial profiles corresponding to advanced plasma scenarios have been implemented in this model, using combinations of step and parabolic functions.The model has been validated with data from state-of-the-art simulations of ITER wall loading, and a parametric study has been performed over different plasma geometries and radial profiles to check the variability of the neutron poloidal profile.The results show the effect of the different configurations on neutron wall loading. This model can be used for parametric studies for conceptual design or systems analysis activities.

Tokamak D-T neutron source models for different plasma physics confinement modes

Neutronic studies of European demonstration fusion power plant (DEMO) have been so far based on plasma physics low confinement mode (L-mode). Future tokamaks, nevertheless, may likely use alternative confinement modes such as high or advanced confinement modes (H&A-mode). Based on analytical formulae used in plasma physics, H&A-modes D-T neutron sources formulae are proposed in this paper. For that purpose, a tokamak random neutron source generator, TRANSGEN, has been built generating bidimensional (radial and poloidal) neutron source maps to be used as input for neutronics Monte-Carlo codes (TRIPOLI-4 and MCNP5). The impact of such a source on the neutronic behavior of the European DEMO-2007 Helium-cooled lithium–lead reactor concept has been assessed and compared with previous results obtained using a L-mode neutron source. An A-mode neutron source map from TRANSGEN has been used with the code TRIPOLI-4. Assuming the same fusion power, results show that main reactor global neutronic parameters, e.g. tritium breeding ratio and neutron multiplication factor, evolved slightly when compared to present uncertainties margin. However, local parameters, such as the neutron wall loading (NWL), change significantly compared to L-mode shape: from −22% to +11% for NWL.

Numerical benchmarks TRIPOLI − MCNP with use of MCAM on FNG ITER bulk shield and FNG HCLL TBM mock-up experiments

3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium–lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI − MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI − MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.

Preliminary evaluation of the expected radiation damage of the bayonet IFMIF back-plate

This paper summarises and discusses the results of a preliminary damage assessment of the non-seizure coating of the bayonet IFMIF back-plate. Neutron-induced kerma factors, dpa and gas production cross sections libraries were produced in a multigroup structure for neutron energies up to 60 MeV, by processing evaluated nuclear data files with NJOY-99.259 system. The material damage evaluations in terms of heat deposition, displacement and gas production rates were calculated using these libraries and compared with the values obtained using the data contained in the pointwise ACE format files of MCNP5 code package. The calculations were performed with MCNP5 code both using the McEnea and the McDelicious neutron source models to reproduce the energy-angle distributions of the neutrons produced in IFMIF d-Li interactions.

Three-dimensional neutron source models for toroidal fusion energy systems

Developments in computer architecture and neutronics code capabilities have enabled high-resolution analysis of complex 3D geometries. Thus, accurately modeling 3D source distributions has become important for nuclear analyses. In this work two methods are described which generate and sample such 3D sources based directly on the plasma parameters of a fusion device and which facilitate the ability to update the neutron source following changes to the plasma physics configuration. The cylindrical mesh method is for toroidally symmetric machines and utilizes data in a standard file format which represents the poloidal magnetic flux on an R–Z grid. The conformal hexahedral mesh method takes plasma physics data generated in an idealized toroidal coordinate system and uses a Jacobian transformation and a functional expansion to generate the source. This work describes each methodology and associated test cases. The cylindrical mesh method was applied to ARIES-RS and the conformal hexahedral mesh method was applied to a uniform torus and ARIES-CS. The results of the test cases indicate that these improved source definitions can have important effects on pertinent engineering parameters, such as neutron wall loading, and should therefore be used for high-resolution nuclear analyses of all toroidal devices.

Modelling of the Production of Source Neutrons from Low-Voltage Accelerated Deuterons on Titanium-Tritium Targets

Fast quasi-monoenergetic neutrons can be produced by accelerating charged deuterons on tritium solid targets. Benchmark experiments were performed in many laboratories with intense D-T neutron sources. The aim is to validate the computational models and nuclear data for fusion applications. The detailed information on the neutron source term is highly important for the benchmark analyses. At present, the MCNP family of codes cannot explicitly model the D-T reaction for Deuterons in the KeV energy range. The physics for the D-T neutron production was modelled at ENEA (Italy) in the SOURCE and SRCDX subroutines to compile with the MCNP source code. Some improvements to the original subroutines were introduced. The differential cross-sections for the D-T reaction from the ENDF/B-VII library were built into the code. The relativistic approach was implemented for neutron kinematics. The new D-T neutron source model was applied to the MCNP5 simulation of the tungsten integral experiment performed at the OKTAVIAN facility. The uncertainty associated with the realistic D-T reactions was separated from the total uncertainty of the source term. The outcome of the benchmark analysis was an improvement in the quality of the computational model of the experiment.

Evaluation of D(d,n) 3 He reaction neutron source models for BNCT irradiation system design

A mathematical method was developed to calculate the yield, energy spectrum and angular distribution of neutrons from D(d,n) 3 He (D-D) reaction in a thick deuterium-titanium target for incident deuterons in energies lower than 1.0MeV. The data of energy spectrum and angular distribution were applied to set up the neutron source model for the beam-shaping-assembly (BSA) design of Boron-Neutron-Capture-Therapy (BNCT) using MCNP-4C code. Three cases of D-D neutron source corresponding to incident deuteron energy of 1000, 400 and 150 keV were investigated. The neutron beam characteristics were compared with the model of a 2.45 MeV mono-energetic and isotropic neutron source using an example BSA designed for BNCT irradiation. The results show significant differences in the neutron beam characteristics, particularly the fast neutron component and fast neutron dose in air, between the non-isotropic neutron source model and the 2.5 MeV mono-energetic and isotropic neutron source model.

Activation of the IFMIF lithium loop corrosion products

The assessment of the activation of steel corrosion products generated in 1 year of International Fusion Materials Irradiation Facility (IFMIF) lithium loop operation due to the interaction between lithium and Stainless Steel SS-304 has been performed. The paper considers both the neutron and the deuteron activation and it describes the approach used for and presents the results obtained. ANITA-IEAF and ANITA-DEUT activation code packages have been used, respectively, for neutrons and deuterons contribution to the inventories calculations and decay gamma sources production. The activation due to deuterons is about 40 times that due to neutrons. Dose rate evaluations via Vitenea-IEF/SCALENEA-1 external to a pipe of the lithium loop have been performed. The deuterons contribution results to be about two orders of magnitude higher of the corresponding ones due to neutrons. For QA purposes, MCNP-4C2 system has been also used to calculate the neutron contribution to the dose rate. Discrepancies between the two approach results are within 1%. A sensitivity analysis aiming to evaluate the impact of the choice of different neutron source models on the dose rates is presented, with reference to the neutron contribution.

Determination of neutron spectra formed by 40-MeV deuteron bombardment of a lithium target with multi-foil activation technique

Neutron flux spectra at an irradiation field produced by a 40-MeV deuteron bombardment on a thick lithium-target at Forschungszentrum Karlsruhe, Germany, have been determined by the multi-foil activation technique. Twenty-seven dosimetry reactions having a wide energy range of threshold energies up to 38 MeV were employed as detectors for the neutron flux spectra extending to 55 MeV. The spectra were adjusted with the SAND-II code with the experimental reaction rates based on an iterative method. The adjusted spectra validated quantitatively the Monte Carlo deuteron-lithium (d-Li) neutron source model code (M CDeLi) which was used to calculate initial guess spectra and also has been used for IFMIF nuclear designs. Accuracy of the adjusted spectra was∼10% that was suitable for successive integral tests of activation cross section data.

Impact of source term uncertainty on neutronics analysis of D-Li neutron irradiation facility (IFMIF)

Abstract To evaluate the neutron irradiation field for IFMIF, a neutron source model of d-Li target reaction was made for the proposed acceleration energies of 30–40 MeV. Impact of the uncertainty in these source models has been examined on neutronics analysis of IFMIF irradiation field.

Neutronics Analysis of the International Fusion Materials Irradiation Facility High-Flux Test Volume

The purpose of the International Fusion Materials Irradiation Facility (IFMIF) is to provide irradiation conditions of a typical deuterium-tritium (D-T) fusion reactor for small material samples, but with higher irradiation levels. An extensive code and data development has been performed, allowing a comprehensive neutronic analysis of the high-flux test volume. New data evaluations for neutron interactions and responses at high energies (20 to 50 MeV) were performed and processed, and a Monte Carlo neutron source model for the Li(d,xn) reaction was developed for use with the MCNP neutron transport code.The neutron flux density was found to be >1014 n·cm-2·s-1 throughout the anticipated high-flux test volume with a high-energy fraction (>14 MeV) of ~20%. The available test volume with >20 dpa/full-power year in iron was found to be 550 ± 180 cm3. This uncertainty is due almost entirely to the uncertainty in the total neutron yield. Hydrogen and helium production rates were calculated and a helium/dpa ratio between 10 and 12 appm/dpa was found, which is similar to that found in a D-T fusion reactor. IFMIF was found to provide an adequate environment for the simulation of D-T fusion reactors, but more work is required to extend and improve the current data and tools.

Concept and neutron-field characteristics of ESNIT

An Energy Selective Neutron Irradiation Test Facility (ESNIT) proposed at Japan Atomic Energy Research Institute (JAERI) is a high-energy neutron irradiation means to conduct materials R&D for general nuclear energy applications including fusion applications. The concept of ESNIT is characterized especially by selectivity of the peak energy of the neutron energy spectra. The selectivity is very useful to investigate the neutron-energy dependence of radiation damage and irradiation effects. Furthermore, it is expected that the influences due to the high energy tail of neutron energy spectra on the radiation damage and the irradiation effects can be reduced by changing the peak energy. In this paper, the neutron-field characteristics calculated were presented as the first step for calculation of damage parameters and estimation of the influence due to the high energy tail. A neutron source model of the D-Li reactions was made to use in neutron-transport calculations for the neutron field of ESNIT, and the characteristics of neutron spectra and the flux distribution were also calculated for various deuteron energies.