Neutronic Design Research Articles

Reactivity Effect Evaluation of Fast Reactor Based on Angular-Dependent Few-Group Cross Sections Generation

Among all the possible occurring reactivity effects of a fast reactor, the situations whereby the control rod was inserted, or the coolant was voided could lead to strong anisotropy of neutron flux distribution, therefore the angular dependence on neutron flux should be considered during the few-group cross-sections generation. Therefore, the purpose of this paper is to compare the influence whether the angular dependence on neutron flux is considered in the calculation of few-group cross sections for the reactivity effect calculation. In the study, the 1-D SN finite difference neutron transport equation solver was implemented in the TULIP of SARAX code system so that the high-order neutron flux could be obtained. Meanwhile, the improved Tone’s method was also applied. The numerical results were obtained based on three experimental FR cores, the JOYO MK-I core, ZPPR-9 core, and ZPPR-10B core. Both control rod worth and sodium void reactivity were calculated and compared with the measurement data. By summarizing and comparing the results of 46 cases, significant differences were found between different consideration of the neutronic analysis. The consideration of angular dependence on neutron flux distribution in the few-group cross-sections generation was beneficial to the neutronic design analysis of FR, especially for the reactivity effect calculation.

Neutronic Design for Heat Pipe Reactor With Annular and Accident Tolerant Fuels

Several core designs of heat pipe reactors with megawatt power were proposed for extreme environments, such as the deep space, the deep sea, and the earthquake locations. However, the existing designs have either the difficulty of manufacture or potential issues of transport. In the present work, a heat pipe design is proposed with an annular fuel element to replace the cylindrical and hexagon fuel elements. In addition, candidate accident tolerant fuels, such as the UN and U3Si2 fuels, are implemented. The neutronic properties of the new reactor design are systematically investigated by the OpenMC Monte Carlo code simulations. It is found that BeO presents a better effect of reducing the axial power deviation than Al2O3. The criticality of the proposed design is verified by two configurations of control drums. The depletion calculations show that each design can operate for decades of years.

Simulating and benchmarking neutron total scattering instrumentation from inception of events to reduced and fitted data

In the design and realization of modern neutron scattering instrumentation, particularly when designing beamline concepts from the ground up, it is desirable to fully benchmark against realistically simulated data. This is especially true for total scattering beamlines, where the future deliverable data is to be analysed in both reciprocal- and real-space representations, and needs must be carefully balanced to ensure sufficient range, resolution and flux of the instrument. An approach to optimize the design of neutron scattering instrumentation via a workflow including ray-tracing simulations, event-based data reduction, heuristic analysis and fitting against realistically simulated spectra is demonstrated here. The case of the DISCOVER beamline concept at the Spallation Neutron Source is used as an example. The results of the calculations are benchmarked through simulation of existing instrumentation and subsequent direct comparison with measured data. On the basis of the validated models, the ability to explore design characteristics for future beamline concepts or future instrument improvements is demonstrated through the examples of detector tube size and detector layout.

Design of an inelastic neutron scattering system based on DT neutron generator for metallic materials analysis

Design of an inelastic neutron scattering system based on DT neutron generator for metallic materials analysis

Evaluation of Fixed Absorber Reactivity Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, fixed absorber worth was measured as the difference of core reactivity measured by control rod worth between cores with and without a single fixed absorber or three fixed absorbers. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as validation data are investigated through a comparison with calculations using the latest neutronics design methodology developed at Japan Atomic Energy Agency. Calculated-to-experiment values and their uncertainties of fixed absorber worth were 1.00 ± 0.05 and 1.02 ± 0.04, respectively. Through this study, the measurements and calculations were found consistent and reliable.

Design, fabrication and comprehensive properties of the novel thermal neutron shielding Gd/316L composites

The constitutive equation of the 155/157Gd areal density and thermal neutron shielding rate was established by Monte Carlo N Particle Transport Code (MCNP) and used to design the novel Gd/316L stainless steel thermal neutron shielding materials. When the 155/157Gd areal density is greater than 0.01 g/cm2, the thermal neutron absorption rate of the Gd/316L composite reaches approximately 99%. The novel thermal neutron shielding materials with different Gd (1 wt.%, - 5 wt.%,) contents were prepared at 950 °C by spark plasma sintering. The transition layer composed of FeNi3, GdNi, and Gd3Ni can be formed during the sintering process. Meanwhile, the mechanical properties of Gd/316L composites decrease with the increase of the Gd content. In H3BO3 solution, pitting corrosion and galvanic corrosion occur at the interface between Gd and 316L matrix.

New 3D Diffusion Code Based on The Nodal Polynomial Expansion

Nodal Expansion Method (NEM) is widely employed in the neutronic design of nuclear reactors core. The main reason is its higher computational performance and efficiency when compared with the conventional fine mesh difference method or finite element method. The NEM diffusion calculation uses coarse spatial meshes and the size can lie in assembly scale. This is the key for the computational efficiency and high accuracy calculations. The NEM consists mainly of basis polynomial function expansion for each nodal direction. The Nodal Expansion Method (NEM), as proposed by Finnemann in 1977, has been used to solve the multigroup neutron diffusion equations in three-dimensional (3D) rectangular geometry. The weighted residual technique has been applied to determine the higher order coupling coefficients. Resulting from the combination of nodal and finite element methods, NEM provides rigorously accurate equations obtained by integrating the neutron balance equation. The two group coarse mesh 3D IAEA benchmark has been simulated by NEM3D-1A using different nodes sizes. The VENTURE

Massimo Salvatores: integral experiments and their use for the validation of nuclear data and the neutronic design of advanced nuclear systems

Among the many domains of reactor physics on which Massimo Salvatores gave his considerable contributions, he was particularly passionate about integral experiments. In this paper, we make a review of selected experimental campaigns among the numerous ones he has promoted, conceived, designed, directed, or analyzed. They have been regrouped in a temporal sequence corresponding to the different periods of Massimo's career, which exceeded 50 years. When possible, for each of the experiments we provide a brief description, the goal for which it was conceived and carried out, and the practical impact on validation and design improvement. Finally, the conclusions offer thoughts and suggestions for the future of the integral experiments and a possible way of honoring the invaluable legacy that Massimo Salvatores has left to us.

NEUTRONIC BENCHMARK ON HOLOS-QUAD MICRO-REACTOR CONCEPT

The Holos-Quad micro-reactor concept is proposed by HolosGen LLC for civilian applications to generate 22 MWt with a lifetime of approximately 8 effective full power years (EFPYs). The design is based on a very innovative high-temperature gas-cooled reactor concept using four Subcritical Power Modules (SPMs) that fit into one commercial 40-foot transport ISO container. Neutronics benchmarks were developed based on a preliminary version of the Holos-Quad design to confirm feasibility of the neutronics design of this non-traditional high temperature gas-cooled (HTGR)-type micro-reactor concept. Calculations were performed using Monte Carlo codes (SERPENT and OpenMC) as well as the PROTEUS high-fidelity deterministic code for two exercises based on a unit cell model and a full core model. The results obtained showed good agreement with all of the evaluated parameters for the unit cell problem and the full core problem. SERPENT and OpenMC display consistently good agreement in eigenvalue within 150 pcm for the unit cell benchmark and less than 270 pcm for the full core benchmark. PROTEUS eigenvalues showed relatively larger differences but still reasonable agreement with the SERPENT solutions. For the depletion benchmark, the observed eigenvalue differences between SERPENT and OpenMC were within 300 pcm throughout the depletion up to 100 GWd/MT, ensuring that the two codes used equivalent input parameters such as the recoverable heat values for fission reactions. This benchmark exercise confirms the neutronic feasibility and core performance of the preliminary Holos-Quad design.

Neutronics analysis of thorium-uranium based molten salt blanket in stellarator-type FFHR

• Detailed CAD model has been designed for thorium-uranium based molten salt blanket. • Complex geometry structure has been analyzed in coupling with Monte Carlo calculation. • Neutronics features have been simulated in consideration of the burnup effect. • The thickness and the 6 Li enrichment have been optimized in the simulation process. The stellarator-type Fusion-Fission Hybrid Reactor (FFHR) is an alternative conceptual design for the tokamak-based devices in order to realize the stable power output. It can bring the viable energy source for the fusion technology in a foreseeable future, and has been widely studied with a lot of advanced concepts being proposed. However, there are some challenges in the neutronics design of the FFHR blanket for the purpose of obtaining the appropriate Tritium Breeding Ratio (TBR) and blanket energy multiplication factor, since the blanket system has a complicated three-dimensional structure in helical shape, which can not be easily translated into native geometry representation syntax of most Monte Carlo neutron transport codes. In this paper, a liquid fuel, F-Li-Be molten salt, with uranium and thorium dissolved has been proposed for the preliminary design of stellarator-type FFHR blanket. Besides, the CAD-PSFO code [ 19 ] has been employed for the simulation of complex geometry structure in the helical blanket, and the OMCB package [ 22 ] has been used in the burnup analysis of the neutronics features in the molten salt blanket. Moreover, detailed geometry design of the blanket arrangement and thickness have been studied in order to meet the requirement of energy multiplication factor in the stellarator-type FFHR. Also, the composition and concentration of the fuel material and 6 Li in molten salt have been discussed for the purpose of tritium self-sufficiency with steady energy supply in the whole lifetime.

Neutronic design and dynamic analysis of a 450 MWth graphite molten salt reactor core

Molten salt reactor (MSR) has good economic potential, inherent safety and nonproliferation characteristics. In this paper, the 450 MWth graphite molten salt reactor is taken as the research object. First, the neutronic design of a 450 MWth graphite molten salt reactor is introduced in terms of the core material, conceptual design, static physical characteristics and core design parameters. Then, the model of the MSR core is established by combining the neutron dynamics model, the thermal model and the auxiliary calculation. Finally, the transient characteristics of the 450 MWth MSR core under the conditions of step reactivity insertion and main pump coast-down are simulated and analyzed. The results show that under the aforementioned conditions, the core power can be stable in the controllable level range, and the core temperature is in a safe range.

Neutronic Design of the Bunker Shielding for the European Spallation Source

Neutronic Design of the Bunker Shielding for the European Spallation Source

Neutronics analysis of the stellarator-type fusion-fission hybrid reactor based on the CAD optimization method

The stellarator plasmas devices have some advantages in steady state operation with less magneto hydrodynamic activities, and the development of stellarator-type FFHR (Fusion-Fission Hybird Reactor) can provide viable energy in the foreseeable future with less requirements to the plasmas technology compared with the pure fusion system. However, the neutronics design of the stellarator-type FFHR has some challenges and difficulties, since the construction of helical structures with large number of high order spline surfaces are the extremely complicated process, and the physical parameters of the complex model have to be changed frequently in the primary design stage, which make the corresponding manual input simulation work into the time consuming and error prone tasks. In this context, the CAD optimization method has been used in converting the dense mesh oriented CAD structure to the low poly style, and the CAD-PSFO code has been employed for the modeling transformation process in Monte Carlo calculation. The thorium resource has been selected as the breeding material for the stellarator-type FFHR, and the polonium and natural abundance of uranium have been used as the seed fuel for the initial loading in order to get the high energy multiplication factor. The OMCB (OpenMC basedBurnupcode) has been introduced to study the burnup features, and stellarator-type FFHR has been designed as the multi-purposes facility that can keep tritium self-sufficiency and get the high breeding ratio with steady energy output.

Neutronics design for molten salt accelerator-driven system as TRU burner

Treatment of plutonium separated from spent fuel has been one of the most important issues in the utilization of nuclear power in Japan. This study investigates a molten salt accelerator-driven system (ADS) to transmute transuranic (TRU) nuclides to address the issue. This study proposes the MARDS (Molten salt AcceleratoR Driven System) concept which has four features: employment of chloride, no dedicated spallation target, non-contact between beam window and fuel salt and modularized core. Neutronics calculation for the MARDS concept was performed for a condition of 400 MW thermal power with 800 MeV proton beam. The calculation results showed that the maximum proton beam current was about 7 mA and about 100–120 kg plutonium could be transmuted during one-year operation. The result indicates that the MARDS concept can employ a circular accelerator instead of LINAC and transmute TRU effectively.

Benchmark experiment of large-angle scattering reaction cross section of iron at 14 MeV using two shadow bars – Comparison of experimental results with ENDF/B-VIII –

ABSTRACT It is important to perform neutron transport simulations with accurate nuclear data in the neutronics design of a fusion reactor. However, absolute values of large-angle scattering cross sections vary among nuclear data libraries even for well-examined nuclide of iron. Benchmark experiments focusing on large-angle scattering cross sections were thus performed to confirm the correctness of nuclear data libraries. The series benchmark experiments were performed at a DT neutron source facility, OKTAVIAN of Osaka University, Japan, by the unique experimental system established by the authors’ group, which can extract only the contribution of large-angle scattering reactions. This system consists of two shadow bars, target plate (iron), and neutron detector (niobium). Two types of shadow bars were used and four irradiations were conducted for one experiment, so that contribution of room-return neutrons was effectively removed and only large-angle scattering neutrons were extracted from the measured four Nb reaction rates. The obtained experimental results were compared with calculations for five nuclear data libraries including JENDL-4.0, JEFF.-3.3, FENDL-3.1, ENDF/B- VII, and recently released ENDF/B-VIII. It was found from the comparison that ENDF/B-VIII showed the best result, though ENDF/B-VII showed overestimation and others are in large underestimation at 14 MeV.

Optimization of AHWR core design: neutronics and thermal-hydraulics analysis and calculations

Advanced Heavy Water Reactor(AHWR) is being designed by BARC for thorium utilisation and demonstrate all aspects of the thorium fuel cycle. The AHWR is a vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The equilibrium fuel cycle of AHWR is being optimised for a discharge burn-up of 40 GWd/Te with Plutonium as makeup fuel in a closed fuel cycle. The 233U content in the Uranium and fissile Plutonium content in the Plutonium have been assumed to be about 78% and 75% respectively. One of the important passive safety features of AHWR is heat removal through natural circulation which is governed by a strong neutronic and thermal hydraulic coupling. The neutronics design simulations were done for average coolant condition. In order to study the effect of natural circulation and its consequences on the coolant behaviour and the resulting power distribution, the neutronics and thermal–hydraulic coupled calculations were carried out. The neutronics calculations code (FEMFOL) and thermal–hydraulic calculations code (ARTHA) were externally coupled to estimate the coolant density distribution and the critical heat flux ratio (CHFR) for the fuel. This paper gives the details of the critical heat flux ratio estimation and the effect of density distribution on the core power distribution at some selected operating phases during core follow-up of AHWR.

Structural pre-conceptual design studies for an EU DEMO equatorial EC port plug and its port integration

Abstract For the EU DEMO Tokamak, Electron Cyclotron (EC) launching systems for plasma heating and stabilization are under development. Various concepts for the optical system are currently studied of which the Mid Steering Antenna (MSA) with a steering mirror at a recessed position behind the breeding blanket (BB), the Open Ended Waveguide (OEWG) concept with quasi-optical beam propagation by fixed mirrors only and the Remote Steering Antenna (RSA), currently seen as back-up solution, are the basic ones. In addition, hybrid solutions, which means a port plug with different launcher concepts, are taken into consideration. In parallel, design drafts for a generic equatorial port plug are sketched with the aim to provide a versatile structural system, which allows customized installation of the potential optical systems. This paper presents a pre-conceptual hybrid design of an equatorial EU DEMO EC port plug based on MSA for plasma stabilization and OEWG for plasma heating, taking into account the exact port position with respect to the toroidal field coil, mechanical integrity, heat dissipation, neutronic shielding requirements, design integration and maintenance concepts.

A novel design of neutron shielding composite materials with three-dimensionally interwoven structure and excellent properties

Neutron shielding materials should have low density, high strength and excellent neutron shielding performance. Ceramic/epoxy composites are one of the candidates for such materials. In this study, neutron shielding composites with three-dimensionally interwoven structure were prepared with matrix of epoxy resin with B4C fine-powders and reinforcement of Al2O3–ZrO2 3D reticulated ceramic with Gd2O3-CA6 layer. The 3D reticulated ceramic was fabricated by polymer replica method. Vacuum infiltration with Al(OH)3/CaCO3/Cd2O3 slurry was used to enhance the properties and structure of 3D reticulated ceramic. Furthermore, the epoxy with 15 wt% B4C powders was also vacuum infiltrated into the ceramic skeletons to form neutron shielding composite materials with three-dimensionally interwoven structure. The as-prepared composite materials shielding rates of fast neutrons and thermal neutrons reached 26.5% and 89.4%, respectively. The bulk density and compressive strength reached 1.65 g/cm3 and 71.14 MPa, respectively. This study demonstrated an effective technique for preparing neutron shielding materials with excellent properties.

CALCULATION OF NEUTRON FLUX DISTRIBUTION AT PIERCING BEAM PORTS OF PLATE TYPE RESEARCH REACTOR BANDUNG

Based on a strategic plan of TRIGA 2000 Bandung’s future operation, BATAN has already decided to implement an option to convert the fuel elements core of TRIGA 2000 from using the cylindrical type of elements produced by General Atomic to MTR plate type of fuel elements produced by local fuel element manufacture. The core design calculation has proved that the core configurations of 5 x 5 matrix using local plate type fuel elements met the requirement of core neutronics design. In addition to the current core configuration, further study must be added to consider the use of beam ports as utilization facilities in the design. The neutron flux distribution at piercing beam port has been calculated based monte carlo algorithm using TRIGA MCNP and MCNP software. The calculation result showed that at piercing beam port surface neutron flux distribution is not quite symmetric. The highest neutron flux at piercing beam port is , where as the flux of neutron thermal energy group is . These results are considerably appropriate for such core configuration and as a result, they can be used as a basic data for designing Plate Type Research Reactor Bandung, especially for neutron diffraction experiment

Preliminary performance analysis and optimization based on 1D neutronics model for Indian DEMO HCCB blanket

India, under its breeding blanket R&D program for DEMO, is focusing on the development of two tritium breeding blanket concepts; namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder (HCCB). The study presented in this paper focuses on the neutronic design analysis and optimization from the tritium breeding perspective of the HCCB blanket. The Indian concept has an edge-on configuration and is one of the variants of the helium-cooled solid breeder blanket concepts proposed by several partner countries in ITER. The Indian HCCB blanket having lithium titanate (Li2TiO3) as the tritium breeder and beryllium (Be) as the neutron multiplier with reduced-activation ferritic/martensitic steel structure aims at utilizing the low-energy neutrons at the rear part of the blanket. The aim of the optimization study is to minimize the radial blanket thickness while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of the HCCB blanket. It is found that inboard and outboard blanket thicknesses of 40 cm and 60 cm, respectively, can give a tritium breeding ratio (TBR) >1.3, with 60% 6Li enrichment, which is assumed to be sufficient to cover potential tritium losses and associated uncertainties. The results also demonstrated that the Be packing fraction (PF) has a more profound impact on the TBR as compared to 6Li enrichment and the PF of Li2TiO3.