A methodology to study multiple designs of a nuclear fission power system for space applications and to obtain optimal systems in light of various possible constraints or objectives is developed in this work. Indeed, mass and size are key parameters to minimize when designing space nuclear power system, that depend on multiple technological options and other input parameters. Thus, a multiobjective constrained optimization method, combining simplified physics models for preliminary design sizing, LHS design-of-experiment, surrogate models and a genetic algorithm is employed to obtain a Pareto front of optimal space nuclear fission power system designs. The methodology is applied to the case of thermal spectrum, HALEU-fueled heat pipe reactors coupled to thermoelectric generators, with a target power of 10 kWe. Such a system was already proposed by CEA (ECSPLORER), and will serve as a reference point for this study. The selected optimization variables of this study are the fuel and heat pipe diameters, the moderator thickness in the reactor, and the converter cold and hot temperatures. The influence of constraints on the design space is analyzed, such as the criticality of the reactor, the heat pipe operating limits and the material temperature limits. The optimization objectives vary from 2 to 4, including the mass of the overall system, the radiator size, the mission lifetime and the maximum electrical power obtainable with the design. The corresponding Pareto solutions show that 10 kWe systems can achieve specific masses of about 189 kg/kWe, or an overall radiator height less than 3.2 m, which outperform ECSPLORER. These performances are significantly reduced however for higher power outputs, whereas mission lifetime can be extended beyond 10 years with little degradation of the other objectives. The methodology could be extended to additional objectives or applied to other designs of space nuclear fission power systems.

The performance of NTP engines consistent with the two HALEU core options described in the literature is assessed. The engine cycle and performance were evaluated using hydrogen as the propellant. The specific impulse is 896 s, with a thrust ranging from 66 kN (UN cermet core) to 113 kN (NERVA-like core). Next, calculations for a manned Mars mission are presented. Various operational concepts are investigated by varying the number of engines, their thrust, the reactor cooldown propellant flow rate, and the number of tanks. Gravity losses are quantified for each mission maneuver. A conjunction-class mission with four engines and four tanks (Delta V of 11.8 km/s) requires a vehicle mass of 450 tons carrying 320 tons of hydrogen. An opposition class mission with six engines and six tanks (Delta V of 15.5 km/s) multiplies the hydrogen mass by a factor of two (with an initial vehicle mass exceeding 800 tons) compared to a conjunction class mission. For reference, the International Space Station weighs approximately 420 tons. This study confirms NTP interest since the propellant mass needed for combustion engines is about five times larger for conjunction class missions and eight times larger for opposition class missions compared to NTP engines.

This study focuses on an Energy Conversion System (ECS) designed to fully compensate the reactivity swing during reactor operation. A methodology is proposed to assess the contribution of the ECS to the core reactivity and to propose a core with produced power uniquely driven by the control of the ECS. The aim is to control the core inlet and mean temperatures in order to take advantage of reactivity feedbacks in the core. PWR and SFR cores, connected to Rankine and Brayton cycles, are analysed. A static method is proposed to take into account the reactivity feedback effects on the core and the CYCLOP code (a software package developed by CEA) is used to model the Energy Conversion Cycles. The conclusion proposes a control of the ECS and the associated cycle efficiency to compensate the reactivity loss and to drive the reactor produced power.

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

Download This Paper Open PDF in Browser Add Paper to My Library Share: Permalink Using these links will ensure access to this page indefinitely Copy URL Copy DOI

In the framework of energy transition, the massive insertion of Variable Renewable Energies leads to think to new ways of controlling and stabilizing electrical networks. Small Modular Reactors (SMRs) could be a sustainable solution if they are proven to be flexible enough. Previous studies have shown the positive influence of SMRs on grids at short-time scales due to inertia and frequency regulation phenomena. In addition, this article aims at studying the influence of constrained grid events on nuclear systems safety and operation. The models of a power system dynamics (PowerFactory) and a nuclear dynamics software (CATHARE) are compared and chained. Two application cases are carried out to quantify the impact and the relevance of this chaining: a short-circuit and a load loss. This article finally concludes that this chaining is relevant to accurately simulate nuclear reactor behavior following grid events. Moreover, a chaining could be insufficient for electrical simulations after severe events such as short-circuits or for high nuclear insertion's rate in an energy mix, a coupling, i.e co-simulation, could be considered.

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commission), a methodology is proposed to early consider safety requirement in the undergoing reactor design process. Before the use of mechanistic tools (CATHARE, SIMMER, EUROPLEXUS, etc.) whose input deck elaboration requires an advanced knowledge of the reactor design, the methodology proposed in this article involves several physical tools simulating phenomena likely to govern the choice of design parameters. These tools are mostly based on reduced order models (ROM), meaning that they mainly involve low-dimensional modelings (mostly 0D and 1D) that are validated versus experimental results. They are gathered in a platform that covers all kind of accidental phenomenology, from the initiator (pump trip, reactivity insertion, local flow blockages, etc.) until the reach of a stable and coolable state after corium relocation. Thus, enabling a large number of simulations in a reasonable computational time, this fast-running platform makes possible the characterization of some major accident transient bifurcations (such as boiling onset, boiling stabilization, primary power excursion, molten fuel vaporization, corium axial relocation in transfer tubes, etc.), in terms of probability of occurrence and of consequences on the transient evolution. It also enables to identify the main physical parameters causing the bifurcations in order to allow a straight feedback on the core design and to give some orientations for future R&D studies. In this article, a focus is firstly made on some scenario bifurcations to illustrate the platform capabilities. The boiling onset and possible reactor state stabilization are studied. The possibility of primary power excursion depending on the core design and the fuel vaporization possibilities are assessed. This platform is also of great help in order to rapidly compare several core designs when facing accidental transients. An example is given by comparing a CADOR core design to an ASTRID-like core design through flow stability maps in natural circulation conditions obtained by the fast-running tool platform. These applications demonstrate the efficiency of the presented methodology integrating safety at the very first design stages, and at facilitating the safety-oriented design of SFRs.

In the current renewable energies’ expansion framework, the increasing part of intermittent electricity production sources (solar or wind farms) in the energy mix and the reducing part of thermal power stations that are nowadays useful to ensure grid stability will lead to a complete paradigm shift concerning the means to ensure grid stability. Nuclear energy, which is carbon-free and dispatchable, may be a sustainable solution to this grid reliability issue if it is adequately designed and implemented on the grid. Several solutions aiming at improving the future nuclear power flexibility are currently under investigation in the literature, among them are those based on Small Modular Reactor (SMR) plants. In order to demonstrate their potential ability to stabilize electric grids, it is necessary to perform electrical dynamic simulations taking into account a spatial and temporal discretization of the grid. In this paper, such calculations are performed using the PowerFactory software. This tool can reproduce electrical grids thanks to models of turbo generators, lines, transformers, loads, I&C systems, etc. The objective is to assess to what extent the innovative SMR features may enhance the frequency control of a grid. For this purpose, a short-circuit event and three frequency stability criteria are firstly defined. Then, a verification of the correct behaviour of the IEEE 39-bus (or New England) grid with regulations is carried out. The relevance of implementing Small Modular Reactors (SMR) instead of large power plants on such frequency stability criteria on this grid is finally assessed, in order to conclude in a preliminary way the possible contribution of small reactors to the future grid’s sustainability.

The objective of the paper is to study the potential behaviour of a power system with high share of nuclear and less thermal plants, in which variable RES insertion increases − for example the French case −, in order to determine the specifications for the design of a potential nuclear reactor with high “manoeuvrability”. Moreover, the flexible reactor may participate more in the supply − demand balance and in particular during large frequency fluctuations caused by the high variability of RES. The studies are carried out with the PowerFactory software, which make it possible to highlight specific needs regarding the power ramp for an “ideal” flexible nuclear reactor. Using a benchmark network, the Kundur “2 areas-4 machines”, the flexibility requirements are obtained as a function of the grid disturbances. For this purpose, the penetration of variable RES is progressively increased, while nuclear power is reduced and thermal power plants are totally suppressed. The study shows that a drop in RES production directly impacts the minimal frequency. A faster response speed of nuclear power makes it possible to restore this stability and return to normal operating conditions imposed by the grid operator. This paper describes therefore the process of obtaining the flexibility criterion for different cases of insertion of variable RES.

The massive penetration of renewable energy sources (RES) that are variable and not “dispatchable”, may weaken the power system supply-demand balance. Nuclear power plants (NPP) contribute in part to this daily and seasonal balance thanks to the “load-following” mode in France for example, but there are still limits to their use. These limits prevent a nuclear power modulation as efficient and quickly as the conventional thermal power plants. The need in terms of power ramps for nuclear in a constrained power system has been quantified in previous studies. Nuclear may compensate for the removal of thermal power plants, in order to fulfill energetic strategies of CO2 reduction. The possibility that nuclear reactors can achieve power ramps of significant values (>5%Pn/min) is put forward and could make possible to replace the services currently provided by thermal power plants. The objective of the study is then to use these power system requirements as the main input parameter for the modelling of a current simplified nuclear reactor capable of responding to frequency control within a specific hypothesis framework. In this paper, a French 1300 MW pressurized water reactor is modelled. Parametric studies are carried out in order to reveal technical and technological constraints when increasing electric power ramp. The study explores ways of design, which may influence reactor flexibility, such as the neutron parameter, Doppler coefficient, or the thermohydraulic parameter, delay in the primary loop.

Sodium cooled fast neutron reactors (SFR) are one of the selected reactor concepts in the framework of the Generation IV International Forum. In this concept, unprotected loss of cooling flow transients (ULOF), for which the non-triggering of backup systems is postulated, are regarded as potential initiators of core melting accidents. During an ULOF transient, spatial distributions of fuel, structure and sodium temperatures are affected by the core cooling flow decrease, which will modify the spatial and energy distribution of neutron in the core due to the spatial competition of neutron feedback effects. As no backup systems are triggered, sodium may reach its boiling temperature at some point, leading to local sodium density variations and making the transient fluctuate in a two-phase flow physics where thermal-hydraulics and neutronics may interact with each other. The transient phenomenology involves several physic disciplines at different time and spatial scales, such as core neutronics, coolant thermal-hydraulics and fuel thermo-mechanics. This paper presents the results of thermal-hydraulic/neutronic coupled simulations of an ULOF transient on the SFR project ASTRID. These coupled calculations are based on the supervisor platform SALOME to link the neutron code APOLLO3® to the system thermal-hydraulic code CATHARE3. The physical approach used by the coupling to describe the neutron kinetic is a quasi-static adiabatic one, updating the normalized spatial power distribution periodically by performing static neutron calculations, while a point kinetic model associated to a neutron feedback model calculates the power amplitude variations.