Neutron Streaming Research Articles

Monte Carlo intercomparison and benchmark of the neutron streaming in the ramified access maze of the CERN High-energy AcceleRator Mixed field (CHARM) facility at CERN

Experiments of high-energy neutron streaming were performed in the access maze of the CERN High-energy AcceleRator Mixed-field (CHARM) facility where high-intensity proton beams of 24 GeV/c impact on a copper target. The streaming of the secondary neutrons through the various legs of the ramified access maze of the facility was measured using aluminium activation detectors installed at 12 different locations: for the first time, the extended coverage of measurement locations allowed to assess the streaming in more distant areas of the maze. The attenuation profile along the maze measured via the production of 24Na was also used to benchmark results from Monte Carlo calculations performed with FLUKA, PHITS, and GEANT4: over almost the five orders of magnitude of the measured production yield, an agreement within a factor 1.4, 1.7, and 2.5 respectively was found for the three codes.

Nuclear analyses for the integration of ITER equatorial Port 2

The present work is devoted to nuclear analyses in support of the ITER diagnostic Equatorial Port 2 (EP#2) integration. ITER EP #2 is a diagnostic port based on the long-modular Diagnostic Shielding Module (DSM) housing the following systems: Disruption Mitigation System (DMS) in DSM#1 and #3 and X-Ray Crystal Spectroscopy Core (XRCS-Core) in DSM#2. Ensuring adequate radiation shielding is a major challenge since the diagnostic systems require several apertures from the Vacuum Vessel (VV) through the Port Interspace (PI) and up to the Port Cell (PC). In the present study, a three-dimensional MCNP model of EP#2 has been developed, starting from the latest design available from Preliminary Design Review stage (PDR), and successively integrated into the reference 40° ITER C-Model. Comprehensive nuclear analyses have been carried out employing the D1SUNED v3.1.4 code based on the MCNP Monte Carlo transport code. Relevant nuclear quantities during and at the end of plasma operations have been evaluated: i.e., neutron and gamma fluxes and energy spectra along the port from the Diagnostic First Wall (DFW) up to the Bio-Shield Plug (BP), nuclear heating, neutron damage, helium and tritium production, and shutdown dose rate. This analysis allowed the identification of potentially critical areas, and therefore, the implementation of additional shielding options aimed at reducing the neutron streaming and mitigating of the radiation field in the PI region. In this work, the results of the analyses are presented and discussed. Some solutions to mitigate nuclear loads and to improve the shielding in PI area are proposed and their impact has been assessed. Finally, some recommendations for the optimization of the design of EP#2 are provided as well.

In-vessel final design of the ITER Wide Angle Viewing System in equatorial port 12

The ITER Wide Angle Viewing System (WAVS) is a major diagnostic for machine protection, plasma control and physics analysis. It will measure the surface temperature of the plasma facing components by infrared thermography and will image the plasma emission in the visible range. The global system is composed of 15 lines of sight located in four equatorial ports, covering more than 80% of the tokamak internal surface.This article focuses on the In-Vessel components of the three lines of sight of the diagnostic in equatorial port 12, which have to be installed for the “first plasma” of ITER.The main components of the system are the First Mirror Unit (FMU) and the Hot Dog-Leg (HDL). The FMU contains two mirrors and all associated equipment allowing their support, protection, cleaning and monitoring. It is a critical component since it has to sustain severe constraints (harsh nuclear environment, plasma proximity, complex interfaces) while it has to achieve challenging optical performances. All elements of the FMU are actively cooled. Furthermore, a pneumatic shutter to avoid direct contamination of the mirrors during first wall conditioning phases can close the FMU. The HDL is composed of two mirrors generating a labyrinth avoiding the direct neutron streaming and transmitting the optical beam toward the windows and the Ex-Vessel components of the WAVS.The main design aspects of the FMU and HDL are presented in this paper. These In-Vessel components have been presented to their Final Design Review (FDR) in July 2022 and their design is mature enough to outline the main features to comply with all the challenging requirements of the ITER machine.

Improved Methods for Calculating the Neutron Diffusion Coefficient of General Pebble Bed

There exists a neutron streaming effect in the high-temperature gas-cooled pebble-bed reactor (HTGR) pebble-bed core caused by the spatial heterogeneity of the neutron’s free path, which has a remarkable impact on neutron leakage. It is necessary to take into consideration the streaming effect in evaluating the homogenized diffusion coefficient of the pebble bed, prior to the whole-core diffusion calculation. In this paper, two methods are proposed for calculating the homogenized multigroup diffusion coefficient of the pebble bed based on migration area conservation theory and Benoist’s theory, respectively. Compared with existing methods, the newly proposed methods are adaptable to a general pebble bed consisting of multitype pebbles and nonvacuum atmosphere. Numerical results demonstrate the proposed methods’ effectiveness and consistency in evaluation of the pebble-bed homogenized diffusion coefficient.

Design of optical transmission line of ITER poloidal polarimeter

ITER Poloidal Polarimeter is laser-aided diagnostics that measures the change of polarization of injected far-infrared laser beam in order to identify plasma current profile. PoPola consists of thirteen measurement chords, and each chord is comprised of a set of optical components. This study shows the design of optics in the ITER vacuum vessel and those in high radiation area outside the vacuum vessel. The design takes into account zero-offset configuration, the minimization of the neutron streaming, remote alignment, accessibility, manufacturability and radiation hardness. The laser power loss due to vignetting loss and absorption is evaluated, and the signal-to-noise ratio at a detector is estimated at 50,000. This study evaluates the acceptable errors of manufacturing and assembly of the optical transmission line on the condition that the demonstrated alignment method using the two steering mirrors outside the vacuum vessel is applied. The evaluation results show that the medium tolerance class of ISO 2768 is basically enough for manufacturing ITER poloidal polarimeter. Only for assembling the in-vessel mirrors to the mirror module, the mirror angles need to be carefully adjusted with accuracy of 0.1° Finally, radiation hardness of a steering mirror is presented. The steering mirror consists of an angle sensor and a piezo actuator. The angle sensors are developed by the authors and free from semiconductor and soft resin. The prototype test results show that angle sensor can measure the mirror angle in the range of ±2.3° with accuracy of 0.37%. As for the piezo actuator, the irradiation test was carried out. The test results show that the piezo actuator is able to operate after the neutron fluence of 3.5 × 1015 n/cm2 and the gamma dose of 0.99 MGy.

Verification of the Discrete Ordinates Goal-Oriented Multi-Collision Source Algorithm with Neutron Streaming Problems

The shielding calculation of neutron streaming problems with ducts is characterized by the strong anisotropy of angular flux, which poses a challenge for the analysis of nuclear installations. The discrete ordinate method is one of the most commonly deterministic techniques to solve the neutron transport equation, in which the accuracy and efficiency neutron are crucial to ensure the reliability of the streaming shielding simulation. We implemented the goal-oriented multi-collision source algorithm in the 3D transport code ARES. This algorithm can determine the importance factor based on the adjoint transport calculation, obtain the response function to enable problem-dependent, goal-oriented spatial decomposition, and provide the error estimation as a driving force behind the dynamic quadrature to optimize the source iteration. This study focuses on verifying the goal-oriented multi-collision source algorithm under the neutron streaming problems, and the capabilities of the algorithm have been tested on IRI-TUB benchmark of SINBAD database. The numerical results show that the algorithm can effectively control the angular discretization error for the neutron streaming problems, which is more economical than the traditional discrete ordinate calculation.

Steady state thermo-mechanics and material property definition framework for analyzing DCLL blanket in the fusion nuclear science facility

A thermo-mechanics model that relies on creating the material property definition framework (MPDF) and multiphysics coupling of the heat transfer and the solid mechanics modules is developed to determine the structural integrity of the recently designed dual cooled lead lithium (DCLL) inboard blanket (IB) for the Fusion Nuclear Science Facility under steady state loads. The MPDF is called to supply fusion relevant neutron irradiation and temperature induced changes in material properties during multiphysics finite element runs, and PbLi temperature profiles are used to approximate Magnetohydrodynamics effect and the nuclear volumetric heating on the PbLi. Neutron irradiation and temperature induced reduction of the yield and ultimate strengths of F82H steel at the first wall (FW) are quantified for one year. A blanket in an assembly with gaps between blanket sectors and another blanket in an assembly with no gaps between blanket sectors, both exposed to radiation damage that lasted for one year are analyzed. Analysis using the elastic ITER structural design criteria for in-vessel components (ITER SDC-IC) design rules and a linear isotropic-hardening-type elastoplastic material model are used where most appropriate. The IB blanket with gaps between blanket sectors will withstand the steady state combined thermal and coolant loads for one year operational period but will fail if no gaps are allowed between blanket sectors. It is recommended that a gap of about 7.62 mm should be provided between IB blanket sectors during assembly which would close up during service, stop neutron streaming, reduce stresses and reduce bending of the FW into the scrape-off layer.

Development of a thermo-mechanical behaviour model adapted to the ITER vacuum vessel material

The ITER machine has been classified as a Basic Nuclear Installation French nuclear regulator (INB n°174), which implies that it will be the first fusion reactor to go through complete French nuclear licencing. The combination of mechanic and electromagnetic phenomena with the heat loads caused by neutron streaming requires a multi-physics approach to the damage assessment, which has not yet been implemented in the common nuclear codes and standards.The general damage prevention methodology consists in guaranteeing the structural integrity of a component. The development of design rules has mainly two origins: prevention of damage from monotonic mechanical loads and prevention of damage from repeated application of loads. In most cases, structural integrity is justified within a linear elastic behaviour but when this route is not enough to respect the design criteria, several non-linear approaches to the material's mechanical behaviour can be considered, requiring more elaborated demonstration of the design compliance. Nevertheless, the models proposed in the nuclear model database are sometimes not sufficient to properly describe the experimentally observed cyclic plasticity behaviour and, in particular, the ratcheting and shakedown phenomena.According to ITER community experts in materials and analyses, a thermo-mechanical behaviour model fitting the ITER Tokamak materials data will guarantee the best prediction of the damage considering a nuclear and a multi-physic loading condition.This paper describes the assessment of the non-linear behaviour of Vacuum Vessel (VV) material with a strong thermo-mechanical coupling and a damage parameter to prevent crack initiation. More precisely, Chaboche's models available in the literature (elasto-(visco)-plasticity models, with various types of hardening) have been enriched in order to explicitly take into account the influence of the temperature on the mechanical behaviour and, reciprocally, the influence of the mechanical behaviour on the temperature. Mechanical cycling tests have been performed on the VV constitutive material, emphasizing on the progressive deformation state up to failure mode, i.e., ratcheting. The proposed models have been tested on a homogeneous problem and the results compared with uniaxial test results; this type of simulation is commonly called “0D”analysis. The last part of this document describes the finite element implementation of the constitutive material model and its application to the ITER VV welded support.

The neutron streaming correction method in the homogenization of Pebble-bed HTGR reflector

In pebble-bed high temperature gas-cooled reactors (HTGRs), the outside graphite reflector generally contains void regions, including the boring channels for control rods, absorber balls and coolant flow. Deterministic reactor physics calculation of HTGR is usually based on regular geometries, which requires pre-homogenization to the materials with void regions. However, the volume-weighted homogenization to the HTGR reflector can greatly overestimate the keff, due to the neutron streaming effect in the reflector channels. This paper proposes a convenient method of correcting the neutron streaming effect in the HTGR reflector, by adjusting the homogenized transport cross sections based on preservation of the neutron transmission probabilities. The proposed method has been implemented in the PANGU code for pebble-bed HTGR physics analysis and fuel cycle simulations. Numerical results suggest that the proposed method can effectively reduce the discrepancy in the homogenization of the HTGR reflector.

Neutron streaming analyses and shielding optimization through ECRH openings in DTT Tokamak building

The Divertor Tokamak Test facility (DTT) will be a superconductive machine that aims at studying the power exhaust issue in a DEMO relevant environment. A significant power across the separatrix (compared to the radius of the machine), PSep/R ≅15 MW/m is of primary importance to fulfill this requirement and it can be obtained by injecting to the plasma up to 45 MW of additional power. This amount of power will be reached with a mix of Heating and Current Drive (HCD) systems: Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating (ICRH) and Negative Neutral Beam Injection (NNBI).Power supply and wave generators of HCD will be hosted in remote-control buildings around the DTT Tokamak hall, and thus several ducts have been designed to connect the systems to the machine. The evacuated transmission lines will pass through dedicated holes in the main building making almost empty and big penetrations which cause a large neutron streaming through the walls.This work is devoted to a three-dimensional neutronics study, in support of the DTT licensing procedure, addressing the neutron and gamma streaming through the ECRH penetrations in order to optimize and determine the effectiveness of several proposed shielding options required to protect workers, population and auxiliary systems.Neutron and gamma transport simulations were carried out with the MCNP5 Monte Carlo code. Surface sources of neutrons and gammas have been used to propagate the radiation outside the penetrations. The ADVANTG hybrid transport code has been also used for variance reduction techniques in order to obtain statistically significant results in regions located several meters away from the plasma source. The results of the calculations are presented and discussed as well as the effectiveness of the proposed options to mitigate the radiation streaming and to reduce the doses outside the tokamak hall.

Measurements and Monte Carlo simulations of high-energy neutron streaming through the access maze using activation detectors at 24 GeV/c proton beam facility of CERN/CHARM

ABSTRACT A measurement of high-energy neutron streaming was performed through a maze at the CERN (Conseil Européen pour la Recherche Nucléaire) High-energy AcceleRator Mixed-field (CHARM) facility. The protons of 24 GeV/c were injected onto a 50-cm-thick copper target and the released neutrons were streamed through a maze with several corridor-legs horizontally designed with the shield walls in the facility. Streaming neutrons were measured by using aluminum activation detectors placed at 10 locations in the maze. From the radionuclide production rate in the activation detectors, the attenuation profile along the maze was obtained for the reaction of 27Al(n,α)24Na. Monte Carlo simulations performed with two codes, the Particle and Heavy Ion Transport System (PHITS) and CERN FLUktuierende KAskade (FLUKA), gave good agreements with the measurements within a factor of 1.7 for the production rates ranging over more than 3 orders of magnitude.

Neutron-gamma flux distributions in cylindrical beam tubes of pool-type research reactor

In nuclear reactor, a variety of ducts are equipped in the radiation shield. Beam tubes, used for experimental facilities for research and development are one of them. It is very important to estimate the neutron and gamma photon flux profile inside as well as at the exit of the duct. This is required from experimental as well as radiation protection point of view. One way to estimate these flux profiles is to use transport method and another way is to use semi-empirical formula which describes energy-space distributions of streamed neutrons and gamma photons in various beam tubes with different shapes. Deterministic transport methodology using discrete ordinates approximation cannot be accurately model the streaming problems. This is due to inability to incorporate sufficient number of angular mesh points in the direction of streaming paths, which leads to ray effect (Miller and Lewis in Computational methods of neutron transport, Wiley, New York, 1985). This ray effect arises in the problem (radiation shielding etc.) having very little scattering and localized sources. In such problems anomalies (unphysical oscillations in the computed scalar flux) likely to arise in the scalar flux distribution. As a result, considerable underestimation occurs. Monte Carlo transport methods seem to be the best method in dealing streaming problems. However, reliable results with small statistical errors are not necessarily obtained in reasonable computation time. So, the use of semi-empirical formula will give the result with suitable accuracy and in less time. In this paper, streaming of neutrons as well as gamma photons through the beam tubes has been calculated using semi-empirical formula and compared with Monte Carlo results. Here, we have considered cylindrical ducts and the results obtained by semi-empirical formula are comparable with those from Monte Carlo code. The results can be verified further when the neutron-gamma flux measurement is done.

Analysis of DD, TT and DT Neutron Streaming Experiments with the ADVANTG Code

The paper presents an analysis of DD, TT and DT neutron streaming benchmark experiments with the recently released hybrid transport code ADVANTG (AutomateD VAriaNce reducTion Generator). ADVANTG combines the deterministic neutron transport solver Denovo with the Monte Carlo transport code MCNP via the principle of variance reduction. It automatically produces weight-window and source biasing variance reduction parameters based on the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using this novel hybrid methodology Monte Carlo simulations of realistic complex fusion streaming geometries have become possible. In this paper the experimental results from the 2016 DD campaign using measurements with TLDs and activation foils up to 40 m from the plasma source are analyzed. New detailed models of the detector assemblies were incorporated into the JET 360° MCNP model for this analysis. In preparation of the TT and DTE2 campaigns at JET a pre-analysis for these campaigns is also presented.

Nuclear analysis of the Single Module Segment WCLL DEMO

The latest design of the Water Cooled Lithium Lead blanket for the European DEMO reactor is based on a Single Module Segment (SMS) concept for the blanket segments: the removal of the gaps between the modules (present in the former Multi Module structure design), with the consequent reduction of the neutron streaming, improves the tritium production and the shielding capabilities. A nuclear analysis has been carried out with the Monte Carlo N-Particle transport code version 5 (MCNP5) and the Joint Evaluated Fission and Fusion nuclear data libraries version 3.2 (JEFF 3.2). A detailed three-dimensional MCNP DEMO model with the Single Module layout has been generated. Three-dimensional neutron and gamma transport simulations have been carried out in order to assess the tritium production, nuclear power deposition in each subcomponent of the blanket and nuclear heating density, neutron flux, neutron damage (dpa) and Helium production, relevant for the design of the system. The results confirm the fulfilment of tritium self-sufficiency and shielding requirements of DEMO. Furthermore, a parametric analysis has been carried out by varying the thickness of the first wall tungsten layer to evaluate how the blanket performances are influenced by the tungsten layer thickness.

Addressing the feasibility of inboard direct-line injection of high-speed pellets, for core fueling of DEMO

Pellet injection represents, to date, the most promising option for core fuelling of the EU-DEMO tokamak. Simulations with the HPI2 pellet ablation/deposition code indicate, however, that sufficiently deep fuel deposition requires injection from the High Field Side (HFS) at velocities ≳1 km/s. Two complementary inboard injection schemes are being explored: one makes use of guide tubes with curvature radii ≥6 m in the attempt of preserving pellet integrity at speeds of ˜1 km/s, the other is investigating the feasibility of injecting high-speed (˜3 km/s) pellets along “direct line of sight” (DLS) trajectories, from either the HFS or a vertical port. Options using quasi-vertical DLS paths routed across the upper vertical port have been explored first, as they can be more easily integrated, Unfortunately, the radial position of the available vertical access (≳9 m from the machine axis) turns out to be unfavorable; further simulations with the HPI2 code predict indeed that vertical injection may be effective only if pellets trajectories are well inboard the magnetic axis. High-speed injection through oblique inboard “DLS” paths, not interfering with the Central Solenoid (CS), are instead predicted to yield good performance, provided that the injection location is ≲2.5 m from the equatorial mid-plane. The angular spread of high-speed free-flight pellets, recently measured using an existing facility, turns out to be enclosed within ˜ 0.7°. This scatter cone may require significant cut off volume of the Breeding Blanket (BB). Moreover, DLS in-vessel conical penetrations may increase the neutron flux outside of the bio-shield, and also result in a significant heat load in the cryogenic pellet source. These issues are being investigated, to identify suitable shielding strategies; preliminary results are reported. The suitability of straight guide tubes to reduce the scatter cone, and hence the corresponding open cross section on BB penetration and the neutron streaming, will be explored as a further step.

Neutrons along the labyrinth of high energy medical accelerators: Effect of wall materials

Neutron streaming along the labyrinth of a generic bunker of an 18MV medical accelerator was evaluated. Monte Carlo simulations using MCNP code were performed to calculate neutron ambient dose equivalent along the labyrinth. The effect of plain, borated and barites concrete wall material, as well as borated concrete and plywood (Celotex), as neutron absorbing wall liners, was examined. The results of the study suggest that plywood can provide a cost effective material to attenuate neutron streaming along the labyrinth.

Neutron streaming calculations in the JET entrance labyrinth

Neutron streaming along the entrance labyrinth of the Joint European Torus (JET) was evaluated. Monte Carlo simulations using MCNP code were performed to calculate neutron fluence and ambient dose equivalent along the length of the labyrinth. The results of this work aim to assist operational radiation protection activities in the JET facility and contribute to the validation of the safety assessment calculations made for ITER.

Experiment and analysis of neutron streaming in iron-polyethylene multi-layer shielding assembly

Experiment and analysis of neutron streaming in iron-polyethylene multi-layer shielding assembly

Stray millimeter-wave radiation loads on ITER fused silica windows

In ITER, the window assemblies will be subjected to the combination of loads due to stray radiation and neutron streaming. Therefore, adequate tests are needed to qualify the windows for the hostile environmental conditions. In ITER, two main systems will generate high power waves in the millimeter range: the Collective Thomson Scattering (CTS) system, with a total power of 1 MW at 60 GHz, and the Electron Cyclotron Resonant Heating (ECRH) system, with a maximum power of 20 MW at 170 GHz. In particular operating conditions, a significant power fraction can be deflected in other ports as stray radiation. High levels of stray radiation could damage diagnostic windows that are safety relevant components. The radiation absorbed by the fused silica window assemblies on ports #11 and #12 have been evaluated in detail. The possible effects of gamma and neutron flux on the dielectric permittivity and loss tangent, determining the absorption, have also been reviewed.

Radiation shielding requirements for the full power operation of the linear IFMIF prototype accelerator (LIPAc) at Rokkasho

The detailed shielding analysis has been performed in order to find a configuration of the radiation shielding in the accelerator building necessary for the operation of the Linear IFMIF Prototype Accelerator (LIPAc), which is under construction in Rokkasho, Japan, at the full power of 1.125 MW to satisfy the radiation dose limit prescribed by Japanese regulations. The final design of the beam dump constructed by CIEMAT, Spain was fully taken into consideration. The impact of all the features potentially causing neutron streaming were investigated and the optimization of local additional shieldings were conducted. As the result, the most important one was found to be the air ventilation duct shield, which completely covers the ducts and creates an additional labyrinth. By considering the additional shielding configuration determined in the present study, the legal criteria at the controlled area boundary can be completely satisfied.