A semi-analytical method for modelling station blackout performance in liquid metal reactors is developed, permitting to identify key factors determining peak temperatures during the transient, and hence to design associated passive safety systems. It is shown that integrity of the fuel cladding during this transient can be ensured by adequate dimensioning of coolant channels, the primary system and the vessel air cooling circuit. These dimensions are determined using algebraic equations and postulated values for a minimum/maximum permissible Reynolds number, dimensionless parameters for the fuel cladding tube geometry and heat sink elevation, a guard vessel height, the nominal core power, permitted temperature gradients in the vessel air cooling system and the air cooling system chimney height. The model suggests that the required coolant volume is a rapidly growing function of core power, and that this volume needs to be 40% larger in a sodium-cooled reactor than in a lead-cooled reactor.

The dynamic response to an Unprotected Station Blackout (USBO) has been evaluated for a small, lead-cooled reactor when fuelled with two different actinide composition: one sourced from spent light water reactor (LWR) fuel and the other from UN fuel discharged from a small LFR. We demonstrate that a reduction in the delayed neutron fraction, primarily due to the addition of americium, leads to lower peak temperatures during phase one of the USBO. This reduction could help with ensuring cladding integrity despite an increased internal gas pressure resulting from helium production during the decay of ²⁴²Cm. It is also shown that the coolant volume required to buffer decay heat until vessel air cooling becomes effective must be increased to ensure the integrity of the fuel cladding. We conclude by demonstrating that (U,Pu)N fuel, with negligible ²⁴¹Pu content, offers the best properties to ensure cladding integrity during the USBO.

China initiative Accelerator Driven System (CiADS) is a 10 MW lead–bismuth eutectic (LBE) cooled subcritical reactor, which is designed to demonstrate the engineering feasibility of the ADS concept. An elaborated methodology was developed to deal with the nonlinear mechanical behaviors of fuels under transients. Based on this methodology, the fuel mechanical module was accomplished and coupled with the neutron dynamics module and thermo-hydraulics module in the extended BELLA code. The cladding mechanical simulation under frequent beam trips for CiADS was carried out using BELLA, and fatigue analysis based on the experimental data was discussed. Combined with the CiADS beam-trip transients, it is concluded that the tolerance of the CiADS cladding to beam trips is between 1000 and 11,000 cycles.

We present the conceptual core design of a small lead-cooled fast reactor, for which a critical configuration has been achieved with a uranium enrichment of 9.9 wt%. This is a novelty for fast-neutron reactors without incorporating mixed uranium/plutonium fuel. It is shown how a reduction in uranium enrichment by two percentage points from a previously designed small lead-cooled reactor leads to an increase in conversion ratio of 20% and a significantly larger reactivity swing. The lowered enrichment gives a stronger Doppler feedback, which leads to lower temperatures during an overpower transient, despite remaining feedback coefficients being less negative. The new reactor geometry is presented along with a detailed neutronic characterisation, where whole-core reactivity feedback coefficients are derived, and depletion calculations are performed. Thereafter, we use the safety analysis code SAS4A/SASSYS-1 to demonstrate that the proposed design remains safe during enveloping unprotected transients, corresponding to Beyond Design Basis Accidents. We show how the reactor has a >2000 degrees C margin to fuel melting during an Unprotected Overpower transient and that thermally induced creep rupture of the fuel cladding tubes is a non-issue despite conservatively assuming 100% fission gas release.

As a new generation of reactor type, lead-cooled fast reactors, have better safety behaviors, higher reliability, and better economic performance, aiming at island power supply through nuclear energy, seawater desalination, optimization of nuclear submarines, etc. Its evolution of nuclear waste and the advantages of nuclear non-proliferation provide a good prospect for development. This paper studied the safety performances of an offshore stationary lead-cooled reactor (OSLR) proposed in the National Key Research and Development Program of China. The transient analysis code SAS4A/SASSYS-1 was used to perform simulations of unprotected over-power accidents (UTOP) and unprotected loss of heat sink (ULOHS) accidents. The results indicated that offshore stationary lead-cooled reactors can withstand a maximum positive reactivity insertion of 0.5$ within 1 s during UTOP accidents without exceeding the working limits of the core. In ULOHS accidents, the inherent safety characteristics of OSLR allowed it to withstand 75 % heat removal capability of IHX. The simulation results were used to analyze the response of this stationary offshore reactor to transient accident conditions and the limits of its ability to withstand accidents in order to provide reference data for subsequent design and ideas for possible development of natural cycle lead-cooled reactors in the future.

A lumped, zero-dimensional, mass transport model is combined with a depletion matrix solver to study the influence of coolant circulation on radionuclide build-up in a small lead-cooled fast reactor. It is shown that the addition of coolant circulation results in a lower activity for a minority of studied nuclides, and it is thus recommended to consider stagnant coolant when licensing a reactor. Activation analysis of three different lead qualities potentially used in SUNRISE-LFR is performed, and the result shows that a low silver content is desirable to simplify maintenance and decommissioning.

The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The subcritical reactor in CiADS is a liquid Lead Bismuth Eutectic (LBE) cooled fast reactor. When the reactor core is in operation, the LBE coolant will directly contact and corrode the inner surface of reactor vessel. Due to the high temperature, the corrosion will be more severe. If the stress on the reactor vessel exceeds the limit, the plastic deformation will occur, leading to the generation and expansion of defects and cracks, and the safety of the reactor will be affected. Therefore, evaluating the stress field of the reactor vessel under different operating conditions is a very important research project. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS, and the ASME Code was used as stress assessment standards. We can preliminarily prove that the stress assessments of the vessel during the postulated Unprotected Loss of Flow (ULOF) accidents satisfy the requirements of ASME Code. The limit reactivity insertion to protect the vessel from plastic deformation is 0.58$ in the postulated Unprotected Transient over Power (UTOP) accidents based on our current results. Therefore, we can preliminarily conclude that the current material selection and structural design of the reactor vessel in CiADS could survive most of the postulated transient accidents considering the stress effect.

UN-UO₂ composites are considered an accident tolerant fuel (ATF) option for light water reactors (LWRs). However, the interactions between UN and UO2 and the low oxidation resistance of UN limit the application of such ATF composite concept in LWRs. A potential alternative to overcome these issues is encapsulating the UN fuel before sintering. Based on our recent studies, molybdenum and tungsten are selected to encapsulate UN spheres. In this article, different coating techniques, such as powder coating, chemical vapour deposition (CVD), and physical vapour deposition (PVD), were developed and applied to encapsulate surrogates and UN spheres. Encapsulated UN-UO₂ pellets fabricated by the spark plasma sintering (SPS) method (1773 K, 80 MPa) were characterised by complementary techniques and evaluated against their oxidation resistance in air up to 973 K. The results show inert, dense, and non-uniform Mo and W layers of about 28 μm and 32 μm, respectively, obtained by the powder coating method. PVD provided uniform and dense layers of Mo and W of approximately 1.0 μm and 4.0 μm, respectively, but with cracks at the interface with the surrogate spheres. PVD-Mo onto UN spheres shows a dense and well-adhered layer of about 0.5 μm but with W contamination from the previous coating. The PVD-W and CVD-W results and the oxidation experiments will be in the final version of this manuscript.

The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The reactor in CiADS is a subcritical fast neutron reactor cooled by a liquid lead-bismuth eutectic. The reactor operates at high temperature and bears high thermal stress. In addition to the heavy weight of the whole reactor, the vessel will bear large effective stress. If the effective stress exceeds the limit of the material, defects and cracks may occur on the main reactor vessel, which will affect the safety performances of the reactor. Therefore, it is very important to analyze the effective stress field of the reactor vessel. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS. We can preliminarily prove that the maximum effective stress that the vessel will bear during the postulated Unprotected Loss of Flow (ULOF) and Unprotected Transient over Power (UTOP) accidents is less than the yield strength of 316L stainless steel. Therefore, we can preliminarily conclude that the current ma-terial selection and structural design of the CiADS vessel could survive the postulated transient accidents considering the effective stress effect.