When handling spent nuclear fuel, the decay heat must be considered to ensure sufficient heat removal. However, the decay heat measurements for spent RBMK fuel assemblies are missing. In this regard, determining accurate spent fuel characteristics with known uncertainties becomes essential for the safety of RBMK spent fuel management. The decay heat uncertainty over the first 100 years after the fuel discharge for the RBMK-1500 fuel was accomplished in this study. Design and nuclear reaction data uncertainties were propagated stochastically through depletion and decay calculations using the Sampler sequence within the SCALE code package. Nuclear reaction data influenced higher uncertainties in decay heat estimations than design data. Cross-section data related uncertainties overshadow uncertainties caused by fission yields and decay data.

The development of small modular reactors (SMRs) at Pantai Gosong, Indonesia, requires careful attention to personnel health and safety, especially during the construction stage. The geological conditions at Pantai Gosong pose significant challenges, as unstable sandstone and igneous rocks necessitate specialized techniques to avoid increasing occupational health and safety (OHS) risks. To address these challenges, effective OHS risk management is crucial, requiring the creation of a modeling framework to identify, analyze, and control potential risks during construction. This study adopts a mixed-methods approach, following the Hazard Identification Risk Assessment and Determining Control method and international standards. It combines the quantitative analysis of risk probabilities and incident frequencies with qualitative methods, including field observations and expert interviews. Key risk analysis encompasses hazard identification, risk probability assessment, and consequence analysis. The findings identified significant potential hazards, primarily crane accidents caused by structural failure factors and transportation accidents caused by dock structural failure (foundation) or corrosion. The proposed model aims to enhance safety standards and support the sustainable deployment of SMRs in Indonesia.

As part of BWXT’s software quality assurance efforts for RELAP5-3D, a comprehensive suite of software verification and validation test problems was developed, informed by phenomena identification and ranking table analyses. Among these, challenging benchmarks were created to verify transient phenomena in reactor systems, specifically focusing on vessel boil-off and turbomachinery performance. A vessel boil-off study simulates the late reflood phase of a loss-of-coolant accident in a pressurized water reactor, verifying closure relationships associated with reactor power, decay heat, liquid-to-vapor phase change, and heat transfer. A turbomachinery study examines the performance of pumps, compressors, and turbines under transient conditions, confirming consistency between RELAP5-3D’s theoretical models and computational implementations. To achieve independent verification, reduced-order models (ROMs) were developed using the MATLAB math engine. These ROMs provide computationally efficient, simplified frameworks for isolating and verifying specific physical models, offering a transparent means of benchmarking RELAP5-3D. This approach not only facilitates rapid testing and error localization but also increases confidence in the primary code’s fidelity by aligning with regulatory guidance on independent verification. This paper demonstrates solutions replicable for similar system code SQA efforts, highlighting the critical value of ROMs in ensuring accuracy and reliability in nuclear safety applications.

Simulation studies using the TRACE computer code are performed for experiments related to density wave oscillations (DWOs) and Ledinegg instabilities in helically coiled steam generator tubes carried out at the SEIT (Società Informazioni Esperienze Termoidrauliche) facility in Italy. The simulation results show that the predictions of the onset of DWO–induced instabilities are in reasonable agreement with the experimental measurements. The TRACE predictions of the temporal behavior of local pressures and flow rates are also in qualitatively reasonable agreement under conditions that resulted in DWO instabilities. This includes good quantitative agreement on the measured oscillation period. On the other hand, the agreement between the simulation predictions and experimentally measured conditions was not as good whenever Ledinegg instabilities and parallel channel asymmetries become prominent. At high system pressure (80 bar), and intermediate pressure (40 bar) with lower injection rates, both the measured and predicted instabilities are due to pure density wave conditions. At intermediate pressure and higher injection rates, higher heat input, or low pressure, the simulations showed that the system undergoes a superposition of DWOs and Ledinegg instabilities. Ledinegg instabilities that are predicted by TRACE at intermediate pressures were not observed in the experiments. At the lowest pressure of 20 bar, the simulation results showed the onset of pure density wave instability at heat rates generally in agreement with the measurements. On the other hand, simulations at 20 bar are shown to result in Ledinegg instabilities at low power levels. These power levels are significantly below the threshold for DWO initiation, and therefore, similar Ledinegg instabilities are not reported for the experiments. Among the reported experimental data and discussions, Ledinegg instabilities are only illustrated by measurements at 21 bar. TRACE simulations, also carried out at 21 bar for which experimental data have been reported, showed a progression of the instability mechanisms as a function of increasing heat rate that generally agrees with the measurements. The simulations progress from stability to large-amplitude oscillations following flow excursion–induced by Ledinegg instabilities that do not overlap due to large channel asymmetry, to less asymmetry, such that the oscillations partially overlap, to the fully overlapping large-amplitude oscillations that comprise the signature of the pure DWOs and instabilities.

The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. This approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.

The SAM code is under development as a modern system-level modeling and simulation tool for advanced non–light water reactor safety analyses, with recent efforts to add capabilities to evaluate radiological source term risks in these novel reactor concepts. By leveraging the established system-level multiphysics thermal-hydraulic models in SAM, a framework for tightly coupled species transport modeling has been integrated into the code for engineering-scale source term evaluation. This species transport framework was first applied to the simulation of tritium, which is a well-known source term in conventional light water reactors. Tritium poses a unique risk in salt-cooled reactors, especially those with lithium-bearing salts such as the fluoride salt–cooled high-temperature reactor (FHR) concept, as tritium is generated in the salt coolant in significant quantities due to neutron interactions. A compounding factor is the increased mobility of tritium at high temperatures, which is able to permeate through metals while also potentially being retained in graphite pebbles and structures. Engineering-scale models for the tritium transport pathways in a FHR have been developed using the new species transport framework in SAM. The capabilities are assessed through analytical verification problems and validated with data from a graphite retention experiment. The system-level model is demonstrated by performing an initial estimate of baseline tritium generation and flows in a generic reference SAM FHR model, setting a foundation for future studies of source term transient analysis with the potential for further multiscale and multiphysics integration.

The measurement of sedimentary source terms (SSTs) in the primary circuit of a pressurized water reactor is important for radiological protection and collective radiation exposure control. In situ gamma spectrometry technology is the uniquely suitable nondestructive method for SST measurement, which gains strong interest from staff members of nuclear power plants. In this work, we provide a comprehensive introduction and discussion on in situ gamma spectrometry technology, including the basic principles, the development status, and the key technologies, with the aim to provide reference and guidance for future research. Specifically, the key technologies of in situ gamma spectrometry technology, including the detector, collimator, and efficiency calibration method, are analyzed and discussed. The detector is generally based on a high-purity germanium (HPGe) or CdZnTe (CZT) detector with the ability to recognize different nuclides from in situ gamma spectra. Variable apertures are recommended for the collimator adapting to different dose rates. The combined source-less efficiency calibration method, which combines Monte Carlo simulations and numerical integration, is the most suitable for in situ measurement of SSTs. The existing problems and future trends are also analyzed and discussed. CZT detectors with large-size crystals should be developed to increase the detection efficiency for high-energy gamma rays. The in situ measurement system should be operating automatically to reduce the workload and exposure of the plant staff. The accuracy of the calibration remains a problem in practical applications. Attentions should be paid to the accuracy of the field condition modeling. In addition to SSTs surveys during shutdown, in situ gamma spectrometry technology can be used for long-term online monitoring of SSTs, which is very helpful for the study of SST control. The problem is how to distinguish the SSTs from the source terms in the coolant, which is quite challenging and needs more study.

As one of the six proposed designs for Generation IV nuclear reactors, the high-temperature gas reactor (HTGR) is being designed to have various passive safety features. Its system safety performance has been investigated both experimentally and numerically, particularly under depressurization scenarios that may occur during postulated accident conditions. In this study, we consider a pipe break accident in the main loop, in which high-temperature and high-pressure helium is discharged into the reactor cavity, resulting in complex flow phenomena involving helium filling, gas mixing, and natural circulation within the cavity. To investigate the jet discharging behavior near the break and the resulting gas mixing in the reactor cavity, a scaled HTGR reactor cavity test facility has been constructed at the City College of New York in which relevant experimental investigations are being carried out. In parallel, unsteady Reynolds-averaged Navier-Stokes (URANS) models have been developed based on geometry and operating conditions of the experimental setup. Numerical simulations have been conducted to reproduce representative test cases, including a mild-buoyant case and a strong-buoyant case with the injection of 75°C nitrogen and 300°C helium, respectively, into the cavity, which was initially filled with room temperature air. Due to the nature of the flow, which becomes quasi-steady during the long transient, a relatively large Courant-Friedrichs-Lewy number of up to 30 is used to accelerate the simulations, ensuring that the long transient process can be captured at a reasonable computational cost. Overall, the URANS predictions show good agreement with the experimental data in terms of time evolution of local gas temperature and oxygen concentration at various sensor locations within the cavity.

Molten salt reactors (MSRs) offer potential operability and safety advantages when compared to commercial light water reactors (LWRs). However, operating experience with MSRs is sparse in comparison to what exists for LWRs. Further, the chemical and isotopic composition of the fuel and/or coolant salt is dynamic and difficult to characterize continuously, posing potential safety, operability, and safeguards unknowns that need to be addressed. A molten salt sampling system (MSSS) is regarded as a necessary subsystem within first-generation MSRs used to obtain samples of salt for chemical and isotopic analysis in support of the need to monitor and control salt composition during operation. The MSSS is being developed using the Safety-in-Design (SiD) methodology, which incorporates incremental integration of safety analysis into the design process. The MSSS conceptual design emerging from the application of the early stages of the SiD methodology consists of a sample collection system and its housing, a freeze port, and inert gas control and delivery systems. This article describes the prototypes developed to test the functions of these MSSS subsystems, presents the results of testing in both dry and molten salt environments (including reliability data collection performed in accordance with the principles of SiD and the development of a semiquantitative fault tree model), and summarizes the opportunities for future design and testing enhancements based on the results of prototype testing.