Uncertainty and sensitivity analyses for steady-state thermal–hydraulics of research reactors

Structured treatment of model uncertainty in complex thermal-hydraulics codes: Technical challenges, prospective and characterization

Summary form only given, as follows. A self-consistent two-dimensional radio-frequency glow discharge model has been developed in cylindrical coordinates using a fluid model. The objective of the study is to provide insights to charged species dynamics and investigate their effects on deposition process for a polyatomic depositing gas discharge. Swarm data as a function of electron energy for methane are provided as input to the model. A power-law scheme is used for the discretization of convection-diffusion terms in the model. The necessary dc bias for the discharge in the asymmetric reactor geometry is predicted by a trial-and-error method such that the cycle-averaged current to the powered electrode becomes zero. The simulations are performed for different experimental design and operating conditions. The model predictions of electron density profile and self-generated dc bias compared well with the experimental results. Comparisons were first made with the data obtained by Sugai and coworkers from Nagoya University, Japan. Computations were carried out for the same geometry and operating conditions of the experimental reactor to obtain the radial and axial variations of plasma variables. The contours of cycle averaged electron and positive ion densities are shown. The present model was used to simulate a second experimental reactor (with a different geometric configuration) at Chemical Vapor Deposition Laboratory, Drexel University. The model predictions of electron density, dc bias and power compared well with the experimental measurements. The validated model was then used to predict the temporal and spatial variations of plasma variables for different reactor operating conditions. The radial variations of species fluxes to the cathode are also presented at different operating conditions of the reactor as they are important for thin carbon film deposition process.

In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing conditions, such as a control-rod movement, induced by an operator, or an accident scenario. The current methods for transient analysis apply deterministic solvers to calculate the neutronic response, but not only do these deterministic solvers always need to discretise the problem, they also often apply more fundamental approximations such as homogenization or diffusion. Therefore a stochastic method is needed, which only has a statistical uncertainty. A challenge for applying the Monte Carlo method on transient analysis, is the different types of particles which must be simulated in a single simulation. It is common for a neutronics calculation to simulate the prompt neutrons, however the simulation of precursors is new. The main challenge when simulating precursors lies in their long lifetime, which is seconds, whereas the prompt neutrons have a lifetime of microseconds. This has been solved by dividing the transient problem into time intervals and forcing the precursors to produce a delayed neutron is all time intervals. This will ensure prompt-neutron chains in all intervals. The prompt neutrons will form prompt-neutron chains, which poses another challenge for the Monte Carlo method. The chain length of these prompt neutrons varies a lot and the can also split into many branches, which averages out in a real power reactor, but is difficult to simulate on a computer. To improve the chain-length statistics, a new variance reduction technique is developed, which dictates that a single neutron will emerge from each interaction. This emerging neutron can be the result of a scattering event or a fission event. Finally, a dynamic simulation scheme, which can run in parallel, has been devised, simulating all time intervals consecutively. Also, a method for sampling the initial conditions is created and this scheme is implemented in a purpose-built Monte Carlo code and in a general-purpose Monte Carlo code. The two codes have been tested in several cases and they behave as expected, agreeing with deterministic method where expected, but deviating when the deterministic methods are no longer valid. In order to simulate realistic transients in a power reactor, feedback has to be taken into account. To achieve this, the dynamic Monte Carlo method has been coupled to a thermal-hydraulics code, using an explicit scheme. The results of this coupled simulation are compared to a state-of-the-art coupled diffusion calculation in a NURISP-benchmark calculation and the results agree well, except for a small deviation towards the end. The exact source of this deviation should be further investigated, but the difference is small, when realising that there is already a deviation in the steady-state results. The results demonstrate the feasibility of performing a fully dynamic transient analysis, using only Monte Carlo for the neutronics part of the calculation. Transients can now be calculated, with detailed modelling of any complex geometry and with continuous energy, which is especially useful for newly developed reactor types and one-of-a-kind research reactors.

Fostering applications of neutron scattering techniques in developing countries: IAEA's role

Understanding of thermal-hydraulic instabilities and the mutual interactions at supercritical pressure

Development of a Multi-Compartment Containment thermal hydraulic code: PARIRODHAN

This paper describes the development of a new thermal-hydraulic (TH) module for the Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) multiphysics code for the full-core simulation of liquid metal–cooled fast reactors, developed at Politecnico di Torino. The code performs steady-state and transient neutronic (NE) and TH coupled calculations while maintaining a relatively low computational cost thanks to the adoption of simplified physical models. The NE module implements the nodal formulation of the multigroup neutron diffusion equations with delayed neutron precursors whereas the TH module treats the reactor hexagonal assemblies as separate channels, which are individually modeled as one-dimensional in the axial direction, accounting for the thermal coupling in the horizontal direction. The new TH module is more robust and portable while providing improved performance with respect to the previous implementation—also thanks to the adopted OpenMP parallelization. Some physical aspects that were previously neglected, such as the thermal inertia of nonfuel rods, have also been included in the model. The development was carried out in accordance with current best practices for code design, implementation, and testing, thus rendering the code easier to be maintained and possibly to be extended in the future. The code usability has also been improved by means of a set of Python classes purposely developed to simplify the input generation and postprocessing phases. This can potentially widen the usage of FRENETIC within the fast reactor community for the simulation of full-core coupled NE-TH transients and/or as a platform to test new solution methods. The paper also includes the application of this new FRENETIC version to a representative configuration of the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) core design. First, the nominal operating conditions of the ALFRED reactor were simulated, proving that the new upgrade of the code is faster and more robust with respect to the previous one. Then, two accidental transients where NE-TH feedback effects are relevant—an unprotected loss of flow and a total instantaneous blockage of a single fuel assembly—were considered.

This paper discusses key issues and highlighted topics in thermal hydraulics in connection with Japan’s current sodium-cooled fast reactor development efforts, with particular focus on design study and related research of the Japan Sodium-Cooled Fast Reactor (JSFR). Several innovative technologies, i.e., compact reactor vessel, two-loop system, full natural-circulation decay heat removal, and recriticality-free core, have been investigated to reduce construction cost and to achieve a high level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are ongoing. In this paper, the progress of design study is reviewed, and key issues are discussed. Then, research and development activities on the thermal hydraulics are highlighted in connection with the phenomena taking place in natural-circulation decay heat removal and innovative design and enhanced safety features.

Taillefer is a versatile Python tool for carrying out Sensitivity Analysis (SA) and uncertainty propagation (UP) studies based on Monte Carlo sampling. Developed with the primary goal of investigating sensitivities and uncertainties of steady-state thermal-hydraulic (SSTH) safety parameters of the high-performance research reactors Forschungs Neutronenquelle Heinz Maier-Leibnitz (FRM II) in Garching, Germany, and the Réacteur à Haut Flux (RHF) in Grenoble, France, it can also be used for a large variety of other modeling problems. The work presented here aims to explain the underlying mathematical background of SA and UP studies with Taillefer and to show some steps to verify these routines. Furthermore, a real-life application example is provided that demonstrates Taillefer’s use in SSTH analysis of the RHF. For this purpose, Taillefer is coupled to the external thermal-hydraulic software PLTEMP/ANL, which is one of the codes used at FRM II and RHF to access SSTH performance and safety parameters. Determining these crucial quantities is part of identifying possible low-enriched uranium (LEU) core designs that are suitable to replace the currently used highly enriched uranium fuels of the two reactors, supporting global nonproliferation efforts. Taillefer is a powerful tool in these conversion studies, as it increases the reliability of the LEU safety parameters by providing information about sensitivities and uncertainties in addition to the nominal values predicted by the thermal-hydraulic software.

A tightly coupled scheme for neutronics and thermal–hydraulics using open-source software

The International Atomic Energy Agency (IAEA) has established a designation for an International Centre based on Research Reactors (ICERR). The intention of this designation is to provide a vehicle for IAEA member states to access international research-reactor and ancillary nuclear research and development infrastructure. The U.S. Department of Energy (DOE) has made a commitment to world leadership in the development of advanced nuclear energy, science, and technology. To this end, DOE has established programs and initiatives to enhance this leadership role. ICERR designation constitutes an important step in achieving the DOE vision. DOE, represented by Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL), is submitting this application for re-designation as an IAEA ICERR. Both INL and ORNL have a decades-long and storied history that supports nuclear research, development, and deployment both nationally and internationally. Both have a history of safe and efficient nuclear operations and have demonstrated a track record of international collaboration and cooperation. INL and ORNL are home to two primary and two secondary research reactors. The primary reactors are • Advanced Test Reactor at INL • High-Flux Isotope Reactor at ORNL. Secondary reactors include: • Neutron Radiography Reactor at INL • Transient Reactor Test (TREAT) Facility at INL • INL Advanced Test Reactor: Critical Facility (ATRC). ICERR designation will also include numerous ancillary facilities at both laboratories, including the following: • Multiple post-irradiation examination facilities at both INL and ORNL • Radiochemistry hot-cell facilities at ORNL • INL nuclear-fuel-fabrication and materials-science facilities • INL and ORNL radiological and nuclear analytical capabilities.

A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the subchannel thermal-hydraulic core analysis code MI based on the momentum integral model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/MI for the accident analysis of the training and research reactor at the Budapest University of Technology and Economics (BME Reactor). An efficient and flexible cross-section generation procedure based on the WIMS code was developed and included in POWEX-K/MI. The code uses an implicit difference approach for the diffusion equations and thermal hydraulics. The neutronic part of the code was tested by simulating a start-up experiment performed at the BME reactor. The obtained results agree well with the experimental results. The code system was then applied to analysing power excursion accidents initiated by ramp reactivity insertions of $0.82 and $1.2. The results show that the reactor is inherently safe in case of such accidents, i.e., no core melt occurs even if the safety rods do not fall into the core.

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Molten salt reactor (MSR) using liquid fuel is one of the Generation-IV candidate reactors. Its liquid fuel characteristics are fundamentally different from those of the conventional solid-fuel reactors, especially the much stronger neutronics and thermal hydraulics coupling is drawing significant attention. In this study, the fundamental thermal hydraulic model, neutronic model, and some auxiliary models were established for the liquid-fuel reactors, and a time-dependent coupled neutronics and thermal hydraulics code named COUPLE was developed to solve the mathematic models by the numerical method. After the code was verified, it was applied to the molten salt fast reactor (MSFR) to perform the steady state calculation. The distributions of the neutron fluxes, delayed neutron precursors, velocity, and temperature were obtained and presented. The results show that the liquid fuel flow affects the delayed neutron precursors significantly, while slightly influences the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank, which leads to the maximal temperature about 1100 K at the centre of the vortex. The results can provide some useful information for the reactor optimization.

In the present scenario, the International Thermonuclear Experimental Reactor (ITER) is in progress and efforts are being made to extend ITER to DEMOnstration Power plant (DEMO) with the purpose to harness the fusion energy for peaceful and constructive purposes. ITER uses the sueperconducting magnet systems for trapping and maneuvering plasma inside the giant tokamak machines. Superconductivity only entails under the critical conditions of temperature, magnetic field and current density. If any of these parameters surpass the critical value, the quench ensues in the cable. In the real system scenarios, electromagnetic, radiative & conductive loads are present, which may affect the performance of the machine. The magnet designer can change the parameters in terms of current and magnetic field to ensure that the critical conditions are met, but when it comes to temperature, it is really hard to locate the hot spot locations. To envisage such situations, it is essential to study thermal hydraulics and hence system code is needed for the whole superconducting magnet system. For such purposes, Cryogenic Circuit, Conductor and Coil code (4C) was built, imbibing the thermal hydraulic modeling of the cable in Conduit conductors (including strands) coupled with structures (which give support to the winding pack, wound by conductors) and an external cryogenic circuit. During the thesis work, the pre-existing 4C code, was modeled for various coils with or without external cooling circuit as required. Henceforth, it was implemented against the wide range of transients, fast transients (~ few seconds) as well as slow transients (~ few days or weeks) to test the flexibility of the code in studying the wide range of physical phenomenon on different time scales. Eventually, the results were validated against the available experimental data to observe the precision of the model in 4C code. One of the analysis carried out during this dissertation included the cool down of one of the non planar coils (NPC) used in Wendelstein 7-X, which was an example of slow transient. In the domain of fast transients, an extensive analysis of AC losses in one of poloidal field (PF) coils with an external cooling circuit, was performed for Korean Superconducting Tokamak Advanced Research (KSTAR), functional fusion device