Multi-physics modeling of VVERs with high fidelity high-resolution codes

The calculations performed for the design and operation of a Nuclear Power Plant (NPP) are a key factor for their safety analyses. The standard for the computational analysis of NPPs is the so called conventional approach, which relies on coarse mesh diffusion for the neutronic solver and 1D channels for the T/H solver. The recent evolution of computing clusters allows the use of a novel approach, with codes performing first-principle based multi-physics simulations with high-resolution of the calculated parameters. Due to their computational cost, these methods can only be used as an audit tool of the conventional approach. In recent years, the nuclear industry has moved towards the establishment of Best Estimate Plus Uncertainty safety limits. The novel approach is a step forward to that direction. At the same time, VVER technology is expanding, with new reactors being built worldwide. However, there is currently no high-resolution tool available for VVER steady state and cycle analysis. There is a clear need for more refined simulation tools capable of handling hexagonal geometries.The goal of this PhD is the development of a novel computational scheme based on the 3D sub-pin neutron transport code nTRACER-FAST (nTF), coupled to the sub-channel code COBRA-TF (CTF) for VVER full core steady state and cycle analysis. The coupling of a neutronic code with sub-pin resolution to a sub-channel T/H solver, as well as the use of CTF for full core VVER sub-channel analysis are one of the main novelties of this work, to the extent of the writer's knowledge. In this thesis, nTF is verified & validated for VVER 3D core standalone neutronic calculations against data published in international benchmarks. The internal sub-pin coupling of nTF to CTF for hexagonal geometries is fully developed on the course of this study. A double domain decomposition scheme that allows both codes to be executed in parallel is designed. The multi-physics core solver is verified for steady state simulations. nTF/CTF is also compared with nTF using a simplified 1D T/H solver, proving that to achieve accurate predictions, computational tools of different fidelity should not be mixed. Finally, nTF/CTF is developed for 3D full core cycle analysis. The coupled code system is validated with experimental data, achieving the target accuracy set for industrial use. The capability of the solver for sub-pin predictions throughout the depletion cycle is demonstrated.A conventional computational route for VVER analysis, built with CASMO5-VVER as a lattice code for the nodal neutronic solver PARCS, is also established. The use of PARCS with CASMO5-VVER is another novelty of the present work. CASMO5-VVER/PARCS is verified and validated for VVER standalone neutronic and multi-physics steady state simulations and cycle analysis. The modeling options that optimize the use of the code system and the generation of CASMO5-VVER cross-sections for PARCS are described. The comparison of the novel and conventional computational routes demonstrates the accuracy enhancement enabled by the high-resolution core solver, during steady state and depletion calculations. The work performed during this thesis also showed that the nTF/CTF computational requirements are manageable for VVER cycle analysis, when a state-of-the-art computing cluster is used. This illustrates that such high fidelity, high-resolution computational scheme can be used as an audit tool for the conventional route.

Analysis of Advanced Sodium-cooled Fast Reactor Core Designs with Improved Safety Characteristics

Kaichao Sun

In order to meet the steadily increasing worldwide energy demand, nuclear power is expected to continue playing a key role in electricity production. Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years' nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it – after the initial loading – to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor-years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback, viz. there is generally a positive reactivity effect when sodium coolant is removed from the core. Furthermore, this so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed improvements address both neutronics and thermal-hydraulics aspects. Furthermore, emphasis has been placed on not only the beginning-of-life (BOL) state of the core, but also on the beginning of closed equilibrium fuel cycle (BEC) state. An important context for the current thesis is the 7th European Framework Program's Collaborative Project for a European Sodium Fast Reactor (CP-ESFR), the reference 3600 MWth ESFR core being the starting point for the conducted research. The principally employed computational tools belong to the so-called FAST code system, viz. the fast-reactor neutronics code ERANOS, the fuel cycle simulating procedure EQL3D, the spatial kinetics code PARCS and the system thermal-hydraulics code TRACE. The research has been carried out in essentially three successive phases. The first phase has involved achieving a clearer understanding of the principal phenomena contributing to the SFR void effect. Decomposition and analysis of sodium void reactivity have been carried out, while considering different fuel cycle states for the core. Furthermore, the spatial distribution of void reactivity importance, in both axial and radial directions, is investigated. For the reactivity decomposition, two methods – based on neutron balance considerations and on perturbation theory, respectively – have been applied. The sodium void reactivity of the reference ESFR core has been, accordingly, decomposed reaction-wise, cross-section-wise, isotope-wise and energy-group-wise. Effectively, the neutron balance based method allows an in-depth understanding of the "consequences" of sodium voidage, while the perturbation theory based method provides a complementary understanding of the "causes". The second phase of the research has addressed optimization of the reference ESFR core design from the neutronics viewpoint. Four options oriented towards either the leakage component or the spectral effect have been considered in detail, viz. introducing an upper sodium plenum and boron layer, varying the core height-to-diameter (H/D) ratio, introducing moderator pins into the fuel assemblies, and modifying the initially loaded plutonium content. The sensitivity of the principal safety and performance parameters, viz. void reactivity, Doppler constant, nominal reactivity and breeding gain, has been evaluated with respect to each of the options. Finally, two synthesis core concepts – representing different selected combinations of the optimization options – have been proposed. The third and last phase has been to test the proposed optimized designs in terms of the dynamic core response to a representative ULOF accident, in which sodium boiling occurs. Such a hypothetical transient is of prime importance for SFR safety demonstration, since it may lead to a severe accident situation, where both cladding and fuel melt. As compared to the response of the reference ESFR core, certain improvements are indeed achieved with the static neutronics optimization carried out. However, these are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting. Thermal-hydraulic optimization has thus been considered necessary to: 1) prevent sodium flow blockage in the fuel channels and 2) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. Following implementation of appropriate thermal-hydraulics related design changes, one arrives at a final configuration of the SFR core in which – for the selected accident scenario – a new "steady state" involving stable sodium boiling is shown to be achievable, with melting of neither cladding nor fuel. Such satisfactory behavior has been confirmed not only for the beginning-of-life core state, but also for the equilibrium closed fuel cycle.

EPFL2012

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

Understanding the time-dependent behaviour of a nuclear reactor following an intended or unintended change of the reactor conditions is of crucial importance to the safe operation of nuclear reactors. Safety evaluations of nuclear reactors involve the analysis of normal and abnormal reactor operation states with adequate and validated simulation tools. There is a growing interest in supplementing the results of best-estimate reactor calculations with uncertainties. A major source of uncertainty in reactor calculations is the nuclear data. The propagation of uncertainties allows to study their impact on response values of reactor calculations and to establish safety margins in nuclear reactor safety evaluations. For the simulation of reactor transients, the two-step approach consisting of the generation of group constants in infinite lattice geometry and the full-core calculation using a neutron-kinetic/thermal-hydraulic code system has been state-of-the-art. Nowadays, Monte Carlo codes allow for the determination of group constants based on a three-dimensional full-core model. Thus, effects from neighbouring fuel assemblies can be considered. In this thesis, the applicability of the neutron-kinetic code DYN3D coupled with the thermal-hydraulic system code ATHLET for the simulation of transients in small reactor cores was assessed. Experimental reactivity accident tests carried out at the SPERT III research reactor were simulated. Appropriate group constants were generated with the Monte Carlo code Serpent. The overall good agreement between the calculated and the experimental results prove the applicability of the applied codes for the analysis of such reactor transients. Nuclear data uncertainties were propagated through the simulations of two control rod withdrawal transients in a mini-core model of a pressurised water reactor. The random sampling-based method XSUSA in combination with the SCALE code package was used to determine varied two-group constants. Based on these two-group constants, 1,000 DYN3D-ATHLET calculations were performed and statistically analysed to obtain uncertainty information for the reactor power and the reactivity. Since the assumption of normality could not be maintained for the reactor power, distribution-free Wilks tolerance limits were applied. By sensitivity analyses, the number of delayed neutrons per fission of U-235 and the scattering cross sections of U-238 were identified as the major contributors to the reactor power uncertainty. For a potential improvement of transient simulations, an approach was developed that was comprised of the determination of group constants using a Serpent full-core model at selected times during a transient simulation and the updating of the original infinite lattice group constants with these full-core group constants. Various challenges and possible solutions were identified. The applicability of the group constants obtained by a full-core model was assessed by their application in full-core diffusion calculations.

EPFL2021

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

EPFL2021

Analysis of Advanced Sodium-cooled Fast Reactor Core Designs with Improved Safety Characteristics

Kaichao Sun

EPFL2012