This article critically examines the UK’s approach to the international principle of justification amidst the ongoing review of the System of Radiological Protection conducted by the International Commission on Radiological Protection (ICRP) and the newly announced UK Nuclear Regulatory Taskforce reviewing the UK’s nuclear regime. It recounts the principle’s journey to legal recognition in the UK, particularly through court cases surrounding activities at the Thermal Oxide Reprocessing Plant (THORP), and its codification into a regulatory mechanism in 2004. By analysing past applications submitted by the Nuclear Industry Association for nuclear new builds, three key challenges emerge: 1) outdated definitions that isolate radiation effects and optimise for the reduction of radiation detriment rather than considering holistic individual and societal well-being, 2) ambiguity regarding what constitutes a ‘class or type of practice’ and the necessary information to achieve higher-level categorisation, and 3) resource constraints causing delays in securing justification decisions. This article explores potential strategies to resolve these issues and advocates for a programmatic approach to regulatory justification for new nuclear reactor designs, complemented by updated guidance documents on the categorisation of nuclear reactor technology, and the option of ‘conditional’ justification. Such an approach could align regulatory justification more closely with ICRP recommendations, de-risk the regulatory decision-making process, and enhance the compatibility of the current mechanism with the UK’s ambitious plans for nuclear expansion and innovation.

In this work, we revisit the use of the virtual density method to model uniform geometrical perturbations. We propose a general algorithm in order to estimate explicitly the effect of geometrical perturbations in continuous-energy Monte Carlo power iteration simulations. We apply the intrusive generalized polynomial chaos method in order to estimate the coefficients of a reduced model giving the multiplication factor as a function of the amplitude of the geometrical perturbation. Our method accurately estimates the reactivity change induced by uniform expansion or swelling deformations of arbitrary geometries, for a large range of deformations within a single Monte Carlo simulation. The reduced model converges rapidly in polynomial order, does not require knowledge of the adjoint flux, and is free from indirect effects.

The global impetus for increased nuclear power is accompanied by the recognition that design standardisation and increased cross-border regulatory harmonisation are essential to achieving such ambitions. This includes establishing greater clarity in how international safety principles are operationalised within national regulatory regimes and aligned with domestic requirements. A persistent source of friction lies in the interpretative ambiguity surrounding the As Low As Reasonably Achievable (ALARA) principle. This study examines the issue through the analytical lens of the UK's non-prescriptive nuclear licensing regime. Within this setting, ALARA is implemented alongside closely related national and international concepts (Best Available Techniques (BAT), Optimisation, So Far As Is Reasonably Practicable (SFAIRP), and As Low As Reasonably Practicable (ALARP)), which are frequently treated as equivalent in practice. The article presents a comparative analysis of these five near-synonyms, examining their regulatory genealogies, underlying justifications, and associated compliance expectations. Drawing on boundary theory, the analysis shows that these concepts function as boundary objects whose apparent equivalence masks divergent regulatory origins and disparate underlying rationales. The analysis demonstrates how false synonymy obscures important differences in how proportionality and reasonableness are embedded at international versus national levels, with consequences for compliance expectations and reactor safety design. Collapsing these into synonymy leaves duty holders navigating unresolved conceptual tensions within a non-prescriptive regime. Such an equivalence carries consequences for reactor safety design and, at a systems level, for the predictability and effectiveness of the licensing regime. The results emphasise the need for earlier and more structured engagement to tackle false presumptions and apparent consensus, and call for sustained international dialogue aimed at developing a verified, shared understanding of proportionality and reasonableness across regulatory contexts.

This paper presents a novel simulation method for transient neutral particle transport. The proposed method builds on the mathematical method of characteristics to solve the time-dependent transport equation by tracing rays that simultaneously traverse the space and time domains. This new approach offers the ability to fully represent the angular component of particle flux in both space and time in a computationally efficient manner. This capability is an advancement over current state-of-the-art methods, which either abandon the angular time dependence and approximate the respective time derivative using isotropic scalar fluxes, or otherwise retain the full angular time dependence at the cost of being computationally impractical for realistic applications. The proposed method adopts the random ray method, which leverages recent advances in stochastic transport theory that enable minimal-bias truncation strategies of the space-time tracks that are not possible with traditional deterministic methods. The resulting truncated space-time characteristics random ray method is implemented in the SCONE neutron transport code and tested on the C5G7-TD community benchmark as well as a simplified radiation streaming problem. The results demonstrate high accuracy without any significant additional time or memory requirements as compared to lower-fidelity state-of-the-art methods and confirm that the new approach remains reliable for problems with strong directional dependence, for which the isotropic approximation fails.

Water-cooled Small Modular Reactors (SMRs) have gained significant attention as they could be modularised, standardised, and factory-fabricated to reduce capital cost and improve passive safety. However, SMRs often come with lower power density, which makes them less competitive with commercial LWRs. To enhance SMRs' operation margin, deployment of the Advanced Technology Fuels (ATFs) can be beneficial. Therefore, this study examines the application of short-term ATF solutions in a soluble boron-free (SBF) SMR design inspired by Rolls Royce's SMR. The analysis integrates neutronic and thermo-mechanical performance evaluations by means of advanced nuclear modelling tools: WIMS, PANTHER, and TRANSURANUS. The study focuses on core design parameters, including reactivity control, power distribution, shutdown margin and other design limits. In addition, fuel performance was investigated in terms of fission gas release (FGR), fuel temperature and cladding strain. The neutronic analyses suggest that the maximum peaking factors of SBF cores are about 15% higher than those of soluble boron-controlled cores. The deployment of FeCrAl cladding, Cr coating of conventional Zricaloy cladding and Cr doping of UO2 fuel have a minor impact on both total power peaking and axial offset. On the other hand, the thermo-mechanical results show that Cr-doped fuel and Cr-coated cladding significantly improve fuel performance, reducing FGR and fuel temperature, while FeCrAl cladding offers limited benefits under the evaluated conditions. Comparative analyses indicate that SBF cores, while operationally viable, present unique design challenges due to a higher power peaking and associated thermal fluctuations, necessitating optimisation of control rod strategies and fuel assembly design. The findings demonstrate the feasibility of incorporating ATFs in SBF SMRs, aligning with European green taxonomy criteria for sustainable nuclear development. This study highlights the potential of combined Cr-doped fuel and Cr-coated cladding as a promising strategy to enhance performance in advanced nuclear reactor designs.

Fluoride-salt-cooled High-temperature Reactors (FHRs) offer high power densities and operating temperatures. The current paper continues to explore the possibility to leverage the knowledge gathered in the operation of British Advanced Gas-cooled Reactor (AGR) fleet to expedite the FHR deployment. The paper looks into the neutronic performance of the potential fuel assembly inside the thermal-hydraulics optimised space. Neutronic simulations were performed at Beginning of Life conditions to identify design streams that demonstrated potential for the most favourable fuel cycle economic performance and other strategic advantages such as tritiumfree NaF-ZrF4 coolant. The study identified several possible assembly configurations with favourable neutronic performance (i.e., negative coolant temperature coefficient). The designs were considered best performing where no further modification of the design parameters led to simultaneous improvement in both k infinity and CTC. The best performing configuration contained FLiBe, uranium-carbide fuel and a small volume fraction of graphite.

Reliable tools to simulate transients are an essential prerequisite for designing new nuclear reactors and assessing their safety. Despite advancements in computer technology and existing neutron transport codes, modeling transients accurately remains resource intensive. This leaves room to explore new transport approaches in the hope they might eventually prove more efficient.This work builds on The Random Ray Method (TRRM), a stochastic variation of the conventional method of characteristics, which has shown great efficiency in terms of run time and precision for steady-state problems. We propose two modified versions of TRRM for time-dependent applications. The first is a time-implicit (TI) method that converges the spatial distribution at each time step before proceeding to the next. The second employs an unconventional approach by converging the entire space-time configuration simultaneously with time-continuous rays (TCR). Both methods were improved with new features for greater precision and dynamic convergence that were not included in previous versions.Tested on C5G7-TD benchmark cases 1-2 and 3-4, the proposed methods reproduced the reference results accurately, with mean errors of less than 0.4%. While TI achieved shorter run times and used less memory than TCR, it resulted in slightly higher errors. Overall, the time-dependent TRRM methods in this work required significantly longer computational times than the reference solutions obtained with the code MPACT, which couples high- and low-order methods. Adopting a similar approach for time-dependent TRRM is a potential topic for future research.

This paper presents an efficient computational approach for modeling the propagation of uncertainties in input variables to output variables in fuel rod thermal-mechanical simulations. Our primary goal was to develop a methodology to identify a reduced sample size capable of providing information on uncertainties and sensitivities while remaining cost effective for computation-intensive high-fidelity three-dimensional simulations or full-core calculations.Our method uses the best-estimate code TRANSURANUS (TU), which is equipped with a built-in Monte Carlo engine. This framework allows for the introduction of uncertainties into the selected input parameters through minor modifications in the input file used for the reference case. We applied this methodology to analyze a representative fuel rod proposed for use in the conceptual molten-salt fluoride-cooled high-temperature reactor (FHR), adapted to the geometry of the advanced gas-cooled reactor (AGR).The computational efficiency of our approach lies in the reduced number of input/output operations. Consequently, we can execute numerous TU runs, enabling a comprehensive comparison of the results generated with a smaller number of statistical runs. To support statistical postprocessing, we developed the TUPython tool. With this tool, we can quantitatively assess both temporal and spatial variations as well as the sensitivity of fuel behavior model responses. The study showed that the sample size of 153, defined by the fourth-order Wilks' method, can be used to economically model uncertainty propagation and perform sensitivity analyses in this specific case.

The XGBoost machine learning algorithm for regression was used to predict (n,2n) microscopic cross sections by training models on physical parameters describing various target nuclei and their corresponding evaluated (n,2n) cross sections sourced from ENDF/B-VIII. Research was concentrated on nuclides with nucleon numbers 30 $$ \le $$<= A $$ \le $$<= 208. Machine learning predictions were compared to library evaluations from ENDF/B-VIII, JENDL-5, JEFF-3.3, CENDL-3.2, and TENDL-2021. Predictions for many nuclides were found to be in agreement with existing evaluated cross sections, with $${r<^>2} \ge $$r2 >= 0.95, with respect to at least one library evaluation found for 73.5 +/- 1.0% of nuclides in ENDF/B-VIII. Predictions were subsequently made on a wide range of exotic nuclides and compared to evaluations from the TENDL-2021 and JENDL-5 libraries.

The Emergency Core Cooling System (ECCS) is one of the critical safety systems in the Gas-cooled Fast Reactor (GFR) demonstrator ALLEGRO. The ECCS design must be improved to enhance fundamental safety functions, particularly in removing heat from the core under postulated plant conditions. In this paper, we focused on exploring the effect of coolant injection point on the performance of ECCS, for more efficient core cooling under Loss Of Coolant Accident (LOCA) scenarios. We initially performed a 1D primary system loop thermal hydraulic simulation to determine appropriate boundary conditions in case of a LOCA. These conditions were then applied to a CFD model of the ALLEGRO reactor pressure vessel, including injection nozzles at two positions. Transient 3D CFD simulations of LOCA in ALLEGRO were carried out using Fluent to determine location of the most effective emergency coolant injection point. We simulated the flow distribution through the core up to around 30 s after both small- and large-break LOCA events. The results suggest that bottom injection of the emergency coolant is preferred in an updated ECCS design. It provides a more uniform flow distribution across the core and therefore more advantageous than the previously suggested location in the downcomer.

Spent nuclear fuel contains fissile isotopes of uranium and plutonium that can present a proliferation concern. Dry cask storage is a popular method for the interim storage of spent fuel assemblies, but once a dry cask is sealed, its contents are no longer directly accessible for verification. Consequently, there is a need for nondestructive analysis techniques to verify dry cask content for nonproliferation security. Since spent fuel assemblies are sources of radiation and heat, any malicious diversion of cask contents should result in observable deviations from the radiation dose and temperature profiles predicted by nominal case calculations. Such deviations between measurements and calculations, therefore, can inform when there may be malicious diversion of the cask content.To examine the potential for a neutronics-based test (based on two-dimensional neutron diffusion) and a thermal-based test (based on one-dimensional heat transfer) for verifying dry cask content, we consider two different methods of applying "single-physics" measurements and predictions in complementary manners, and also develop a multiphysics test. Although neither approach provides a high level of confidence, we find that the latter method gives better results and can be recommended as the preferred approach for future development. We close with a sketch of a potential validation pathway for the multiphysics testing methodology.

A novel transport operator, called “HELL”, has been developed and integrated into the SCONE Monte Carlo code. This alternative approach combines two commonly used tracking methods: surface-tracking and deltatracking. The unique aspect of this operator is its flexibility, allowing users to assign these tracking methods to different geometry universes according to their specific requirements. The methodology of HELL and its implementation in SCONE is presented and detailed. The performance and efficiency of the HELL have been evaluated through its application in two different cases. The first case involved a core model featuring involute surfaces, while the second case focused on a Pressurized Water Reactor (PWR) application. The results demonstrated that HELL effectively eliminates the limitations typically associated with surface-tracking in complex geometries and delta-tracking in materials with localized heavy absorbers. This makes it a valuable tool in nuclear physics simulations and computations. The HELL method can significantly reduce the calculation time required for such scenarios, and can be widely applied to Monte Carlo modelling in reactor physics.