A methodology to assist the design of liquid metal reactors, passively cooled by natural circulation duringoff-normal conditions, is derived from first principle physics. Based on this methodology, a preliminarydesign of a small LFR is accomplished and presented with accompanying neutronic and reactor dynamiccharacterizations. The benefit of using this methodology for reactor design compared to other availablemethods is discussed.

This paper describes the design, implementation and characterisation of an Autonomous Reactivity Control (ARC) system in a small modular lead-cooled fast reactor. The aim of this work was to demonstrate the applicability of the ARC system and to study its dynamic behaviour during an anticipated transient without scram. A simplified one-dimensional model was developed to calculate the heat transfer within the ARC system, and the reactivity worth as a function of the neutron poison’s insertion into the active core was obtained via static neutronic calculations. By coupling the aforementioned models, the ARC’s time-dependent reactivity was derived as a function of the coolant outlet temperature variation. This model was implemented into the BELLA multi-point dynamics code and transient simulations were run. A control rod ejection accident was studied leading to an unprotected transient overpower scenario, in which 350 pcm reactivity was inserted during one second. It was shown that the ARC system provides a forceful negative reactivity feedback and that steady-state temperatures after the transient were reduced by almost 300 K compared to an identical transient without its action. In this scenario, the ARC system managed to stabilise the coolant outlet temperature at a value 100 K above nominal conditions. The implementation of an ARC system provided the reactor with a passively actuated self-regulating reactivity control system able to insert large amounts of negative reactivity in a short amount of time.

Passive safety systems in a nuclear reactor allow to simplify the overall plant design, beside improving economics and reliability, which are considered to be among the salient goals of advanced Generation IV reactors. This work focuses on investigating the application of a self-actuated, gravity-driven shutdown system in a small lead-cooled fast reactor and its dynamic response to an initiating event. The reactor thermal-hydraulics and neutronics assessment were performed in advance. According to a first-order approximation approach, the passive insertion of shutdown assembly was assumed to be influenced primarily by three forces: gravitational, buoyancy and fluid drag. A system of kinematic equations were formulated a priori and a MATLAB program was developed to determine the dynamics of the assembly. Identifying the delicate nature of the balance of forces, sensitivity analysis for coolant channel velocities and assembly foot densities yielded an optimal system model that resulted in successful passive shutdown. Transient safety studies, using the multi-point dynamics code BELLA, showed that the gravity-driven system acts remarkably well, even when accounting for a brief delay in self-actuation. Ultimately the reactor is brought to a sub-critical state while respecting technological constraints.

Breeding of nuclear fuel from fertile nuclides may allow to extend known nuclear fuel resources from a century to thousands of years. In this article, we show that the requirement for a breeder reactor fuel to feature an effective neutron production per absorption larger than 2 (eta > 2) breaks down for fertile and fissionable nuclides meeting two criteria related to the relative magnitude of capture and fission cross sections. Moreover, we find that a breeding ratio larger than unity can be achieved for fuels consisting of a single nuclide, in spite of the this nuclide featuring eta < 2. In particular, neptunium is identified as a nuclear fuel that can sustain a reactivity increase over time up to a burn-up exceeding 240 GWd/ton in a fast neutron spectrum.

In this paper, a novel approach for transmutation of americium in fast reactors is presented. Using enriched uranium as fissile support, rather than plutonium, it is shown that a minor actinide burning rate of 25 kg/TWh(th) is possible to achieve in a passively safe, critical lead-cooled reactor. Moreover, the plutonium produced by transmutation of Am-241 features up to 38% (PU)-P-238, making it difficult to use for weapons production.

In this paper, the conceptual design of a small lead-cooled nuclear reactor intended to replace diesel-power in off-grid applications is presented. In a vessel of dimensions making it transportable by air, the targeted design performance is to produce 3 MW of electrical power for up to 30 years without reloading of fuel. Consequently, the inner vessel can be sealed, delaying malevolent access to the nuclear fuel and improving security. Alumina forming alloys are applied to ensure long term corrosion protection of fuel cladding tubes, steam generator tubes and primary vessel over the operational temperature regime. Moreover, decay heat can be removed in a completely passive manner by natural convection from the core to the primary coolant and by thermal radiation from the primary vessel to the environment. Finally, the source term is such that relocation of population residing beyond 1 km from the reactor will not be required even in the case of a complete core melt.

SEALER is a small lead-cooled reactor, designed by LeadCold Reactors for commercial power production in off-grid applications, such as Arctic communities and mining operations. For the purpose of safety-informed design, and independent verification of transient performance, KTH and LeadCold are developing BELLA, a lumped-parameter dynamics code for simulation of protected and un-protected transients in liquid metal cooled reactors. In the present contribution, results from a preliminary benchmark on UTOP and ULOF transients between BELLA and SAS4A/SASSYS-1 are presented. Whereas the codes predict very similar fuel and clad temperatures during the UTOP simulation, discrepancies in natural convection mass flow following a loss-of-flow event are identified and discussed.

In this paper, the multi-point dynamics code BELLA and its benchmarking with respect to SAS4A/SASSYS-1 is described for a small fast reactor cooled with natural convection of lead (ELECTRA). It is shown that BELLA is capable of reproducing the magnitude of mass-flow, reactivity, power and temperature excursions during design extension conditions with an accuracy better than 10%. Hence, the BELLA code can be used for safety-informed design and stability analyses of fast reactor systems, permitting to isolate essential phenomena and trends of significance for their safety assessment. (C) 2015 Elsevier Ltd. All rights reserved.