Molten-Salt Reactor Experiment

The Molten-Salt Reactor Experiment (MSRE) was an experimental molten salt reactor research reactor at the Oak Ridge National Laboratory (ORNL). This technology was researched through the 1960s, the reactor was constructed by 1964, it went critical in 1965, and was operated until 1969. The costs of a cleanup project were estimated at about $130 million. The MSRE was a 7.4 MWth test reactor simulating the neutronic "kernel" of a type of inherently safer epithermal thorium breeder reactor called the liquid fluoride thorium reactor. It primarily used two fuels: first uranium-235 and later uranium-233. The latter 233UF4 was the result of breeding from thorium in other reactors. Since this was an engineering test, the large, expensive breeding blanket of thorium salt was omitted in favor of neutron measurements. In the MSRE, the heat from the reactor core was shed via a cooling system using air blown over radiators. It is thought similar reactors cou

Improved methodology for analysis and design of Molten Salt Reactors

Rodrigo Gonzalez Gonzaga De Oliveira

Molten Salt Reactors is a class of advanced reactors that is typically characterised by the presence of a molten inorganic salt of a fissile material as fuel. At the moment, this class of reactors is considered one of the promising options in the long run of nuclear power.In the context of safety assessment of the MSFR, a task on analysis of selected transients was planned by the SAMOFAR project, to which this thesis should contribute. In addition, fluids with high melting temperature bring with them challenges regarding solidification of the fluid and possible flow blockage, to which a method was implemented during this thesis to evaluate the consequences. The open core cavity with curved shape adds another issue where the core is highly turbulent and requires the use of unstructured mesh codes for routine analysis, adding uncertainty and computational burden to the analytical workflow. Using ATARI, the tool developed during this thesis, we:1. Analyse the MSFR together with partners in support of a safety assessment task.2. Consider the impact and how to include solidification/melting in an analytical methodology.3. Evaluate options to manage/reduce uncertainties and the burden of routine use of high-fidelity codes.After a careful derivation of the conservation laws and assembling the code main algorithms, ATARI undergoes verification in a limited scope using the Stefan problem, manufactured solutions and energy balances.The nuclear cavity benchmark was performed between SAMOFAR partners showing equivalency of the codes within the scope of the benchmark. Then the participating institutions perform analysis of the MSFR for the safety assessment task, which ATARI is unable to followup with the transient analysis. The reason for failure is identified as the requirement of appropriate acceleration schemes by OpenFOAM-based multiphysics solvers, and the lack of one in ATARI.In the lack of accurate data regarding the heat exchangers, the impact of solidification is evaluated in a few simplified cases of a pipe with square cross-section and heat exchanger. For different salt compositions and different boundary conditions, the heat exchanger is found to recover from blockage if pump maintains its momentum and some cross-flow is allowed. The recovery of cases with abrupt solidification is questionable due to the required pressure head.The use of flow baffles is investigated as an option to structure the flow inside the reactor and reduce uncertainties originating from turbulence modelling while allowing the use of structured mesh codes for analysis. The concept is tested in a chloride based MSR, which retains its breed and burn mode while showing a quasi-1D flow appropriate for modelling with legacy codes.We recommend that the nuclear cavity benchmark should be expanded to include a more demanding transient case, representative of MSFR transient analysis requirements. A case is made for homogenised and economical models as used in ATARI to enable the practical use of coupling methods that promote the simultaneous solution of equations. A niche for ATARI is found as a special-mission code for analysis of challenging cases with deforming structure and flow field that justifies the use of unstructured meshes and the added computational burden. Due to design uncertainties and their nature, it is recommended to consider solidification implicitly when analysing the MSFR fuel circuit.

EPFL2021

Conceptual Design of a Sustainable Waste Burning Molten Salt Reactor

Boris Aviv Hombourger

For an energy source to qualify as sustainable, it must maximize the efficiency with which it uses natural resources while minimizing the amounts of waste it produces. Currently deployed nuclear power plants however use a very limited amount of the energy contained in natural nuclear fuels such as Uranium and Thorium and produce long-lived nuclear waste that must be dealt with. Molten Salt Reactors (MSRs) are one family of advanced reactors designs with the potential for safe and sustainable energy production due to their use of a liquid molten salt fuel. This thesis aimed at reviewing MSR designs that have been proposed in the past, investigating alternative design possibilities and proposing a concept capable of disposing of existing waste in a sustainable manner. One possibility to do so is to design a reactor to run on an isobreeding closed fuel cycle while using existing actinide waste as a start-up inventory, which is the route investigated in this thesis. For this purpose, the EQL0D fuel cycle procedure was developed using the MATLAB environment and the Serpent 2 Monte Carlo neutron transport code to simulate the start-up, transition and equilibrium of MSRs. The procedure was then applied to investigate the performance at equilibrium of several candidate fuel salts and moderators in an infinite lattice and in a closed Uranium-Plutonium and Thorium-Uranium fuel cycle. The results showed that only an isobreeding closed Uranium-Thorium fuel cycle can be achieved in moderated MSRs, while both isobreeding Uranium-Plutonium and Thorium-Uranium closed fuel cycles can be sustained in fast-spectrum MSRs, which additionally show a substantial reactivity margin that enable them to dispose of poorly fissile nuclides. The transition behavior of selected fluoride- and chloride-fueled reactors was then investigated. The fluoride-fueled cores all feature a closed Thorium-Uranium fuel cycle and included two historical graphite-moderated designs (the single- and two-fluid Molten Salt Breeder Reactors) and a modern fast-spectrum design (Molten Salt Fast Reactor). The chloride-fueled cores were proposed based on the results obtained in the infinite lattice study. The results obtained indicated that the transuranic-burning capabilities of fluoride-fueled reactors are substantially limited by the low solubility of transuranic trifluorides in fluoride salt mixtures, while no such limitation was found in chloride salts. On the other hand, chloride salts necessitate substantially larger cores and inventories. Finally, the possibility of operating MSRs on two types of open cycles without fuel processing, breed-and-burn and once-through, was investigated to alleviate concerns related to the uncertainties related to the technical feasibility of molten salt fuel processing in closed fuel cycles. The results obtained show that it is possible to operate chloride-fueled MSRs in a sustainable breed-and-burn fuel cycle and also on an open cycle with appreciable burn-ups achieved.

EPFL2018

Appraisal of Core Analysis Methods against Zero-Power Experiments on Full-Scale BWR Fuel Assemblies

Flavio Dante Giust

The present doctoral research aims at the appraisal of nodal core simulators used for the calculation of commercial boiling water reactor (BWR) cores, against measurements carried out at the Paul Scherrer Institute under the LWR-PROTEUS experimental programme. The research focuses mainly on the prediction of radial and axial total-fission rate (i.e. power) distributions in the vicinity of core heterogeneities, caused by features such as the presence of control blades, enrichment boundaries or partial length rods. As such, this thesis complements previous validation work performed for LWR calculational methods on the basis of integral experiments in zero-power research reactors, the latter corresponding largely to the validation of two-dimensional, reflected fuel-assembly calculations. Two commercial code systems have been investigated currently: HELIOS/PRESTO-2, presently used for core monitoring at the Leibstadt Nuclear Power Plant (Switzerland), and CASMO-5/SIMULATE-5, which represents the most recent generation of the widely applied CASMO/SIMULATE system. Six different LWR-PROTEUS configurations have been modelled, for which the nodal reconstructed total-fission rate distributions have been compared against experimental results, as well as against reference, three-dimensional whole-reactor Monte Carlo calculations using the MCNPX code. To start with, a methodology has been developed and tested for appropriate representation of the multi-zone experimental configurations, as employed in the LWR-PROTEUS programme, by means of reduced-geometry models set up using the investigated nodal code systems. The approach adopted has been to apply, in each case, case-dependent three-dimensional boundary conditions to the nodal code's modelling of the 3x3 array of BWR assemblies constituting the LWR-PROTEUS test zone. This has been done in terms of so-called Partial Current Ratios (PCRs), which describe the relation between the incoming and outgoing neutron currents across the test-zone boundary. These PCRs have been derived from the fore-mentioned MCNPX calculation of the multi-zone PROTEUS reactor, thus allowing for the adequate description of the three-dimensional effects associated with the interaction between the test zone and its surroundings. The results obtained for the reference LWR-PROTEUS configuration, with a test zone consisting of an unperturbed regular array of identical, axially uniform BWR fuel assemblies (of type SVEA-96+), showed that both the investigated nodal methodologies reproduce the experimental total-fission rate distribution with very good accuracy, equivalent to that achievable with high-order transport calculations. This has provided adequate indication that the developed methodology of using three-dimensional PCRs, calculated by means of an appropriate whole-reactor MCNPX model, offers a reliable platform for the desired validation. The full insertion of a L-shaped hafnium control blade, especially fabricated for the LWR-PROTEUS programme, has permitted the study of the behaviour of calculated total-fission rate distributions in the presence of strong radial and azimuthal flux gradients. In this case, the use of reflective boundary conditions in the lattice calculations, combined with the azimuthal non-uniformity of the currents that couple the test zone with the outer zones of the reactor, produce systematic deviations which, together, lead to some loss of accuracy in comparison with the regular, uncontrolled case. The three-dimensional effects of the changes in enrichment and gadolinium content between two different axial sections of the fuel assembly have also been studied. In this case, the calculations of the axial total-fission rate distributions are seen to generally reproduce the experimental data with good accuracy, although certain systematic effects are observed. In particular, at local level, and very near the strong axial heterogeneity caused by the enrichment boundary, relatively large deviations have been found to occur. However, due to their very local character, these deviations are not considered to represent a relevant issue in connection with power reactor monitoring and design. Also for the demanding case of the LWR-PROTEUS configuration with a partially inserted control blade, the results obtained in this thesis show a very good behaviour. As in the boundary enrichment case, however, some very localized deviations occur in the vicinity of the strong axial flux gradients. In a separate phase of the LWR-PROTEUS programme, BWR assemblies with partial length rods (PLRs), viz. of type SVEA-96 Optima2, were studied in the central test zone. The results of the comparisons made currently for the corresponding LWR-PROTEUS configurations have shown that, also for fuel assemblies with axial heterogeneities of this type, the performance of both the investigated nodal methodologies is very satisfactory for predicting the global distribution of the total-fission rate. The local three-dimensional effects caused by the PLRs are also predicted with good accuracy, similar, in fact, to that obtained with high-order three-dimensional transport calculations. The present research shows that the two nodal code systems investigated – HELIOS/PRESTO-2 and CASMO-5/SIMULATE-5 – reproduce the LWR-PROTEUS experimental results with high accuracy, both for the radial and the three-dimensional (i.e. at pellet level) comparisons. Significant deviations occur almost exclusively within a short distance from the nodal interface, in cases featuring strong gradients across the core midplane. The axial and radial shapes of the total-fission rate distributions are well predicted in most cases, which represents an important observation concerning the monitoring of operational limits in power reactor cores. Thus, overall – within the range of applicability of the experimental conditions studied in the LWR-PROTEUS programme – the results of the comparisons performed in this thesis confirm that nodal methodologies with pin-power reconstruction have a very high level of performance. In fact, the accuracy that they can achieve for the assembly-internal total-fission rate distribution is, in general, higher than that practically achievable in an operating nuclear power plant. This is due to unavoidable, additional uncertainties in the characterisation of a BWR core under power reactor conditions. Finally, the present research has provided quantitative insights into the applicability of integral data, produced in critical facilities such as PROTEUS, for validating the calculation of power-reactor core heterogeneities. One has thus been able to demonstrate that the experimental evidence from the LWR-PROTEUS programme indeed constitutes a very valuable basis for the validation and appraisal of three-dimensional nodal methodologies with pin-power reconstruction.

EPFL2012