Specifications for a coupled neutronics thermal-hydraulics SFR test case

Coupling neutronics/thermal-hydraulics calculations for the design of nuclear reactors is a growing trend in the scientific community. This approach allows to properly represent the mutual feedbacks between the neutronic distribution and the thermal-hydraulics properties of the materials composing the reactor, details which are often lost when separate analysis are performed. In this work, a test case for a generation IV sodium-cooled fast reactor (SFR), based on the ASTRID concept developed by CEA, is proposed. Two sub-assemblies (SA) characterized by different fuel enrichment and layout are considered. Specifications for the test case are provided including geometrical data, material compositions, thermo-physical properties and coupling scheme details. Serpent and ANSYS-CFX are used as reference in the description of suitable inputs for the performing of the benchmark, but the use of other code combinations for the purpose of validation of the results is encouraged. The expected outcome of the test case are the axial distribution of volumetric power generation term (q'''), density and temperature for the fuel, the cladding and the coolant

Neutronics Studies on the NIST Reactor Using the GA LEU fuel

The National Bureau of Standards Reactor (NBSR) located on the National Institute of Standards and Technology (NIST) Gaithersburg campus, is currently underway of fuel conversion from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. One particular challenging part of the conversion of the NBSR is the high average flux level (2.5×1014 n/cm2-s) required to maintain experimental testing capabilities of the reactor, without significant changes to the external structures of the reactor. Recently the General Atomics (GA) Training Research Isotopes General Atomics (TRIGA) fuel has shown some promising features as a LEU candidate for the high performance research reactors such as the NBSR. The GA fuel has a long history of success in conversion of research reactors since it was developed in 1980s. The UZrH compound in the GA fuel has seen success in long term TRIGA reactors, and is a proven safe LEU alternative. This study performs a neutronics evaluation of the TRIGA fuel under the schema of the NBSR’s heavy conversion requirements in order to examine whether the TRIGA fuel is a viable option for conversion of the NBSR. To determine the most optimal path of conversion, we performed a feasibility study with particular regard to the fuel dimensions, fuel rod configurations, cladding, as well as fuel structure selection. Based on the outcome of the feasibility study, an equilibrium core is then generated following the NBSR’s current fuel management schema. Key neutronics performance characteristics including flux distribution, power distribution, control rod (i.e., shim arms) worth, as well as kinetics parameters of the equilibrium core are calculated and evaluated. MCNP6, a Monte Carlo based computational modeling software was intensively used to aid in these calculations. The results of this study will provide important insight on the effectiveness of conversion, as well as determine the viability of the conversion from HEU to LEU using the GA fuel

Conceptual design of the EU-DEMO dual coolant lithium lead equatorial module

© 20xx IEEE. Personal use of this material is permitted. Permission from IEEE must be obtained for all other uses, in any current or future media, including reprinting/republishing this material for advertising or promotional purposes, creating new collective works, for resale or redistribution to servers or lists, or reuse of any copyrighted component of this work in other works.Within the framework of EUROfusion Program, the Dual Coolant Lithium Lead (DCLL) is one of the four EU breeder blanket concepts that are being investigated as candidates for DEMO. DCLL uses PbLi as the main coolant, tritium breeder, tritium carrier, and neutron multiplier. The main structures, including the first wall, are cooled with helium. The EU program proposed for the next years will consider a DCLL version limited to 550 °C in order to allow the use of conventional materials and technologies. During the first year of EUROfusion activities, a draft design of the DCLL has been proposed. The main blanket performances were adapted to the new specifications and the CAD model of DEMO. The breeder zone has been toroidally divided into four parallel PbLi circuits, separated through stiffening grid radial walls. The PbLi flow routing has been designed to maximize the amount of thermal power extracted by flowing PbLi and to avoid the occurrence of reverse flows due to volumetric heating. Thermal hydraulics, magnetohydrodynamic and neutronics calculations have been performed for the first draft design. The new DCLL design employs Eurofer-alumina-Eurofer sandwich as flow channel insert (FCI).Postprint (published version

Collision statistics for random flights with anisotropic scattering and absorption

For a broad class of random walks with anisotropic scattering kernel and absorption, we derive explicit formulas that allow expressing the moments of the collision number $n_V$ performed in a volume $V$ as a function of the particle equilibrium distribution. Our results apply to arbitrary domains $V$ and boundary conditions, and allow assessing the hitting statistics for systems where the typical displacements are comparable to the domain size, so that the diffusion limit is possibly not attained. An example is discussed for one-dimensional (1d) random flights with exponential displacements, where analytical calculations can be carried out.Comment: 9 pages, 5 figure

Comparative analysis of neutronics/thermal-hydraulics multi-scale coupling for LWR analysis

The aim of the research described in this paper is to perform consistent comparative analyses of two different approaches for coupling of two-scale, two-physics phenomena in reactor core calculations. The physical phenomena of interest are the neutronics and the thermal-hydraulics core behaviors and their interactions, while the spatial scales are the “global” (assembly/channel-wise) and the “local” (pin/sub-channel-wise). The objective is three-fold: qualification of coupled code systems by consistent step-by-step cross-comparison (in order to understand the prediction deviations in both neutronics and thermal-hydraulics parameters); assessment of fine scale (local/subchannel-wise) thermal-hydraulic effects; and evaluation of the impact of on-line modeling of interactions of the two spatial scales. The reported work is within the cooperation between the Universidad Politécnica de Madrid (UPM), Spain and the Pennsylvania State University (PSU), USA. The paper first presents the two multi-scale coupled code systems followed by cross-comparisons for steady state calculations. Selected results are discussed to highlight some of the issues involved in comparative analysis of coupled multi-scale simulations. The transient comparisons are subject of future work and publications

A new model with Serpent for the first criticality benchmarks of the TRIGA Mark II reactor

We present a new model, developed with the Serpent Monte Carlo code, for neutronics simulation of the TRIGA Mark II reactor of Pavia (Italy). The complete 3D geometry of the reactor core is implemented with high accuracy and detail, exploiting all the available information about geometry and materials. The Serpent model of the reactor is validated in the fresh fuel configuration, through a benchmark analysis of the first criticality experiments and control rods calibrations. The accuracy of simulations in reproducing the reactivity difference between the low power (10 W) and full power (250 kW) reactor condition is also tested. Finally, a direct comparison between Serpent and MCNP simulations of the same reactor configurations is presented

Gas core reactors for actinide transmutation and breeder applications

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions

Development of an Analytic Nodal Diffusion Solver in Multigroups for 3D Reactor Cores with Rectangular or Hexagonal Assemblies.

More accurate modelling of physical phenomena involved in present and future nuclear reactors requires a multi-scale and multi-physics approach. This challenge can be accomplished by the coupling of best-estimate core-physics, thermal-hydraulics and multi-physics solvers. In order to make viable that coupling, the current trends in reactor simulations are along the development of a new generation of tools based on user-friendly, modular, easily linkable, faster and more accurate codes to be integrated in common platforms. These premises are in the origin of the NURESIM Integrated Project within the 6th European Framework Program, which is envisaged to provide the initial step towards a Common European Standard Software Platform for nuclear reactors simulations. In the frame of this project and to reach the above-mentioned goals, a 3-D multigroup nodal solver for neutron diffusion calculations called ANDES (Analytic Nodal Diffusion Equation Solver) has been developed and tested in-depth in this Thesis. ANDES solves the steady-state and time-dependent neutron diffusion equation in threedimensions and any number of energy groups, utilizing the Analytic Coarse-Mesh Finite-Difference (ACMFD) scheme to yield the nodal coupling equations. It can be applied to both Cartesian and triangular-Z geometries, so that simulations of LWR as well as VVER, HTR and fast reactors can be performed. The solver has been implemented in a fully encapsulated way, enabling it as a module to be readily integrated in other codes and platforms. In fact, it can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. Verification of performance has shown that ANDES is a code with high order definition for whole core realistic nodal simulations. In this paper, the methodology developed and involved in ANDES is presented