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.

This compilation thesis presents work focused on the design and safety analysis of the small lead-cooled research and demonstration reactor SUNRISE-LFR–the first step toward the construction of a next-generation reactor in Sweden. Two versions of SUNRISE-LFR are introduced—the second necessitated by the lack of access to uranium enriched above 10 wt.% 235U/U. Neutronic characterization is performed using the Monte Carlo code Serpent 2, while the reactors’ behaviours during design extension conditions (DECs) are analysed using the in-house developed code BELLA and an established fast reactor safety code, SAS4A/SASSYS-1. An analytical method for designing a passively safe lead-cooled reactor is derived and used to propose the core configuration of SUNRISE-LFR. This method is subsequently expanded into a semi-analytical framework for designing the Reactor Vessel Auxiliary Cooling System (RVACS), aimed at ensuring fuel cladding survivability during unprotected station blackout (USBO) transients. The model is further extended to evaluate the impact on system temperatures during USBO transients when using nuclear fuel with different actinide compositions. It is shown that actinide compositions with low concentrations of americium and the plutonium isotope 241Pu are beneficial for cladding integrity. Finally, the thesis assesses the impact of coolant circulation on the total neutron activation of the lead coolant over the reactor’s operational lifetime. It is demonstrated that a sufficiently pure lead vector—particularly one with low silver content—could allow the coolant to be exempted from radiological control within a reasonable time frame, thereby avoiding the need for disposal in a final repository. This thesis serves as both a foundation and a stepping stone for the continued development, licensing, and eventual construction of a lead-cooled reactor in Sweden.

Denna sammanfattningsavhandling presenterar ett arbete som fokuserar på design och säkerhetsanalys av den lilla blykylda forsknings- och demonstrationsreaktorn SUNRISE-LFR—det första steget mot att bygga nästa generations reaktor i Sverige. Två versioner av SUNRISE-LFR presenteras, där avsaknaden av tillgång till uran med en anrikning över 10 wt.% ²³⁵U/U var anledningen till att en ny design behövde tas fram. Karakterisering av reaktorernas neutronfysikaliska egenskaper har genomförts med Monte Carlokoden Serpent 2, och deras beteende under utökade konstruktionstillstånd (Design Extension Conditions, DEC) analyseras med den internt utvecklade koden BELLA samt den etablerade snabbreaktorkoden SAS4A/SASSYS-1. En analytisk metod för att designa en passivt säker blykyld reaktor härleds och används för att föreslå härdkonfigurationen för SUNRISE-LFR. Denna metod expanderas för att möjliggöra en semianalytisk dimensionering av det passiva hjälpkylsystem för reaktorkärlet (Reactor Vessel Auxiliary Cooling System, RVACS) i syfte att säkerställa bränslekapslingens integritet under en oskyddad totalförlust av elförsörjning (Unprotected Station Blackout, USBO) transient. Modellen vidareutvecklas för att utvärdera hur systemtemperaturerna påverkas under USBO transienter vid användning av kärnbränsle med olika aktinidsammansättningar. Det visas att aktinidsammansättningar med låga halter av americium och plutoniumisotopen ²⁴¹Pu är gynnsamma för kapslingsintegriteten. Avhandlingen undersöker även hur kylmedelscirkulation påverkar den totala neutronaktiveringen av bly under reaktorns driftstid. Det visas att en tillräckligt ren blyvektor—särskilt med låg silverhalt—kan möjliggöra friklassning av kylmediet inom rimlig tid efter avveckling, utan behov av slutförvaring. Denna avhandling utgör både en grund och ett första steg i det fortsatta arbetet med att utveckla, licensiera och i slutändan bygga en blykyld reaktor i Sverige.

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.

Recently, several countries, including Sweden, have begun investigating the deployment of Small Modular Reactors. Licensing documentation for new reactors must include (preliminary) waste management and decommissioning plans. For light water reactors, the majority of the long-lived radioactive waste from their decommissioning consists of neutron-activated metallic components in the core region, the reactor internals and the reactor pressure vessel. Usually, this waste is planned to be disposed of in geological repositories, and the long-term safety after closure of Sweden's operating and planned repositories is mainly determined by their inventory of the long-lived radioisotopes C-14, Ni-59, and Ni-63. In this study, a three-dimensional model of NuScale's Power ModuleTM, a light-water small modular reactor, was created using Serpent 2, a multi-purpose three-dimensional continuous-energy neutron and photon transport code. The model consists of the nuclear core and surrounding metallic components. Based on burn-up calculations, the neutron flux in the equilibrium core was obtained. From the neutron flux, the cumulative concentrations for a range of relevant short- and long-lived radioisotopes were calculated in the metallic structures in the core region after 60 years of operation. The average specific activity of the selected radionuclides in the different components of the NuScale Power Module was calculated and compared to the activities in existing or anticipated decommissioning waste from Swedish reactors to evaluate the suitability of the Power Module's decommissioning waste for the Swedish waste management system. This study indicates that a single NuScale Power ModuleTM could generate between 1.08 and 2.13 m3 metallic LL-ILW/ GWel-year, i.e., radiologically suitable for disposal in the planned Swedish geological repository for long-lived, intermediated-level radioactive waste. These energy-equivalent volumes are estimated to be somewhat larger than those anticipated for the three existing large Pressurized Water Reactors at the Ringhals site in Sweden. However, the large reactors give rise to additional 4.26 m3 of concrete LL-ILW/GWel-year to be disposed of in the aforementioned repository. In conclusion, if NuScale's Power ModuleTM were to be considered for deployment in Sweden, its decommissioning waste would technically be suitable for the current waste management system. Still, additional studies would be required to determine the optimal disposal route for components situated further away from the nuclear core.

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.

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.

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.