D. Lathouwers

Validation of the DALTON-THERMIX Code System with Transient Analyses of the HTR-10 and Application to the PBMR

Abstract The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of the new three-dimensional diffusion code DALTON, which is coupled to the existing thermal-hydraulic code THERMIX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5. The behavior of pebble-bed reactors during a loss of forced cooling (LOFC) transient is of particular interest since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature. For validation purposes, calculation results of normal operation, an LOFC transient, and a control rod withdrawal transient without SCRAM have been compared with experimental data obtained in the High Temperature Reactor–10 (HTR-10). The code system has been applied to the 400-MW(thermal) pebble bed modular reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system. It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100°C) as compared to its nominal value in the HTR-10 but reaches a maximum value of 1648°C during the depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976°C) during normal operation.

Iterative computation of time-eigenvalues of the neutron transport equation

Abstract The Implicitly Restarted Arnoldi Method (IRAM) is applied to the computation of prompt time-eigenvalues of the neutron transport equation. Derivation of the eigenvalue problem is based on a least-squares functional combined with a spherical harmonics angular discretization and spatial finite elements. The method is applied to a mono-energetic homogeneous slab and compared to semi-analytical results. The results are found to be accurate if the angular discretization is sufficiently refined. The scheme is also applied to a model ADS geometry in both one and three energy groups. The IRAM is found to be very efficient but the solution of fixed source problems that are part of the algorithm need to be accelerated in the multigroup case to obtain an overall efficient method.

Development of a Three-Dimensional Time-Dependent Calculation Scheme for Molten Salt Reactors and Validation of the Measurement Data of the Molten Salt Reactor Experiment

Abstract This paper presents the development, validation, and results of a three-dimensional, time-dependent, coupled-neutronics–thermal-hydraulic calculational scheme for channel-type molten salt reactors (MSRs). The reactor physics part is based on diffusion theory, extended by a term representing the flow of the fuel through the core. The calculation of the temperature field is done by modeling all fuel channels, which are coupled to each other by a three-dimensional heat conduction equation. For the purpose of validation, the results of the MSR Experiment (MSRE) natural-circulation experiment and the thermal feedback coefficients of the reactor have been calculated and compared. With the aid of a code system developed to implement this scheme, calculations were carried out for the normal operating state of the MSRE and some debris-induced channel-blocking-incident transients. In the case of the MSRE, it is shown that the severity of such an incident strongly depends on the degree of channel blocking and that high-temperature gradients in the moderator can connect thermally the adjacent fuel channels. Results are included for an unblocking transient (i.e., the debris suddenly exits the core, and the fuel flow reverts to the normal operating pattern), and it was demonstrated that during the unblocking large power peaks can be induced.

Space-dependent kinetics simulation of a gas-cooled fluidized bed nuclear reactor

In this paper we present numerical simulations of a conceptual helium-cooled fluidized bed thermal nuclear reactor. The simulations are performed using the coupled neutronics/multi-phase computational fluid dynamics code finite element transient criticality which is capable of modelling all the relevant non-linear feedback mechanisms. The conceptual reactor consists of an axi-symmetric bed surrounded by graphite moderator inside which 0.1 cm diameter TRISO-coated nuclear fuel particles are fluidized. Detailed spatial/temporal neutron flux and temperature profiles have been obtained providing valuable insight into the power distribution and fluid dynamics of this complex system. The numerical simulations show that the unique mixing ability of the fluidized bed gives rise, as expected, to uniform temperature and particle distribution. This uniformity enhances the heat transfer and therefore the power produced by the reactor.

Core Design and Fuel Management Studies of a Thorium-Breeder Pebble Bed High-Temperature Reactor

Abstract An inherently safe thorium-breeder pebble bed reactor has great potential to improve the safety and sustainability of nuclear energy. The aim of this work is to determine the conditions under which breeding is possible in a thorium-breeder pebble bed reactor (PBR) and to present possible core designs for such a reactor. A method is developed to calculate the equilibrium core configuration of a thorium-breeder PBR, consisting of a driver channel and a breed channel. The SCALE system is used for cross-section generation and fuel depletion, and a two-dimensional (r,z)-flux profile is obtained using the DALTON neutron diffusion code. With the code scheme, the influence of several geometrical, operational, and fuel management parameters on breeding capability can be studied. Four fuel reprocessing schemes are investigated. The first scheme recycles breeder pebbles into the driver channel after some delay for additional 233Pa decay. The second scheme reprocesses the discharged breeder pebbles to make driver pebbles with higher 233U content. The third scheme also reprocesses the uranium isotopes from the discharged driver pebbles. Criticality, and thus breeding, can only be achieved in practice for this case. The fourth scheme, which adjusts the driver pebble residence time to find a critical core, is used to design a thorium-breeder PBR under practical operating conditions. A breeder reactor can even be achieved for a 150-cm core diameter, the same as for the uranium-fueled HTR-PM, but the design presented operates at a significantly lower reactor power, 71 MW(thermal) compared with 250 MW(thermal).

What is wise in the production of 99Mo? A comparison of eight possible production routes

The present paper addresses eight possible routes of producing 99Mo, and discusses both yield and 99Mo specific activities (SA) in the context of anticipated worldwide demand. Target dimensions are modelled by considering both limits set by cooling and by inside-target radiation attenuation characteristics. Energy deposition profiles are set up by MCNP6, reaction probabilities are taken from TALYS/TENDL and JANIS codes, and both are used in arriving at the produced 99Mo. The outcomes suggest that U neutron-fission may remain one of the most relevant and efficient means of producing 99Mo at the world-demand level, but that within this domain new developments may surface, such as ADSR or AHR production modes. Accelerator-based 99Mo production is discussed as asking for developments in both target cooling and new concepts in post-EOB upgrading of 99Mo SA, and/or new concepts for 99Mo/99mTc-generators, the latter possibly in both volumes (mass) and 99Mo capacities.

Dynamics modeling and stability analysis of a fluidized bed nuclear reactor

Abstract A theoretical model describing the coupling of neutronics, thermohydraulics and fluidization in a fluidized bed nuclear reactor is presented. The stability of the system is investigated by linearizing and perturbing the system around its equilibrium points and identifying the root loci of the sytem. It is found that within the operational range, the eigenvalues are located in the negative part of the phase plane, implying linear stability. Simulations of transient conditions are performed, viz. a hypothetical startup transient and a quasistatic transient related to noise resulting from stochastic movements of the fuel particles. These simulations show that although the total power of the reactor may reach high values, the fuel temperature is well below safety limits at all times.

Fast and accurate sensitivity analysis of IMPT treatment plans using Polynomial Chaos Expansion

The highly conformal planned dose distribution achievable in intensity modulated proton therapy (IMPT) can severely be compromised by uncertainties in patient setup and proton range. While several robust optimization approaches have been presented to address this issue, appropriate methods to accurately estimate the robustness of treatment plans are still lacking. To fill this gap we present Polynomial Chaos Expansion (PCE) techniques which are easily applicable and create a meta-model of the dose engine by approximating the dose in every voxel with multidimensional polynomials. This Polynomial Chaos (PC) model can be built in an automated fashion relatively cheaply and subsequently it can be used to perform comprehensive robustness analysis. We adapted PC to provide among others the expected dose, the dose variance, accurate probability distribution of dose-volume histogram (DVH) metrics (e.g. minimum tumor or maximum organ dose), exact bandwidths of DVHs, and to separate the effects of random and systematic errors. We present the outcome of our verification experiments based on 6 head-and-neck (HN) patients, and exemplify the usefulness of PCE by comparing a robust and a non-robust treatment plan for a selected HN case. The results suggest that PCE is highly valuable for both research and clinical applications.

An Investigation of Power Stabilization and Space-Dependent Dynamics of a Nuclear Fluidized-Bed Reactor

Abstract Previous work into the space-dependent kinetics of the conceptual nuclear fluidized bed has highlighted the sensitivity of fission power to particle movements within the bed. The work presented in this paper investigates a method of stabilizing the fission power by making it less sensitive to fuel particle movement. Steady-state neutronic calculations are performed to obtain a suitable design that is stable to radial and axial fuel particle movements in the bed. Detailed spatial/temporal simulations performed using the finite element transient criticality (FETCH) code investigate the dynamics of the new reactor design. A dual requirement of the design is that it has a moderate power output of ˜300 MW(thermal).

A space-angle DGFEM approach for the Boltzmann radiation transport equation with local angular refinement

In this paper a new method for the discretization of the radiation transport equation is presented, based on a discontinuous Galerkin method in space and angle that allows for local refinement in angle where any spatial element can support its own angular discretization. To cope with the discontinuous spatial nature of the solution, a generalized Riemann procedure is required to distinguish between incoming and outgoing contributions of the numerical fluxes. A new consistent framework is introduced that is based on the solution of a generalized eigenvalue problem. The resulting numerical fluxes for the various possible cases where neighboring elements have an equal, higher or lower level of refinement in angle are derived based on tensor algebra and the resulting expressions have a very clear physical interpretation. The choice of discontinuous trial functions not only has the advantage of easing local refinement, it also facilitates the use of efficient sweep-based solvers due to decoupling of unknowns on a large scale thereby approaching the efficiency of discrete ordinates methods with local angular resolution. The approach is illustrated by a series of numerical experiments. Results show high orders of convergence for the scalar flux on angular refinement. The generalized Riemann upwinding procedure leads to stable and consistent solutions. Further the sweep-based solver performs well when used as a preconditioner for a Krylov method.