Accidents in sodium cooled fast reactors such as Unprotected Loss of Flow (ULOFA), and Total Instantaneous Blockage (TIB) involve coolant boiling and molten material motion inside the voided channel. Accurate modeling of molten clad dynamics is required to calculate transient power evolutions, fuel motion modeling, and transition and disassembly phase calculations. In the present study, a one dimensional molten clad motion model is developed. The model is validated with analytical calculations and benchmark experiment data. Integrating the clad motion model, TIB and ULOFA scenarios in a medium sized SFR is analyzed using the ASTRA code. The TIB analysis revealed that fast voiding resulted in rapid power rise and fuel melting. The coolant channel was blocked by the refrozen clad at the lower axial blanket. The parametric study showed that the heat generation rate affects the time of occurrence of key events such as sodium boiling, clad melting, fuel melting. The ULOFA analysis showed a gradual introduction of steel relocation feedback and a complete channel blockage at the top of the fissile region at the time of fuel melting. The parametric study showed that the heat generation rate is the most influential parameter that affects the time of occurrence of key events.

Understanding the spatio-temporal transport of the aerosol resulting from a sodium fire is essential for mechanistic assessment of the severe accident source term. To develop mechanistic models for evaluating aerosol transport, the Indira Gandhi Center for Atomic Research (IGCAR) has started a series of experiments to understand the spatio-temporal behaviour of the aerosol in a large volume. These experiments involved the measurement of the temperature, humidity and aerosol characteristics at multiple locations, which improved the understanding of aerosol spatial dispersion during typical sodium fires. In the first experiment, a pool fire was created using 2 kg of sodium in the MINA test chamber to obtain the spatial dispersion of the aerosol over time along with the temperature distribution. Before attempting a detailed mechanistic assessment of aerosol transport, lumped codes were developed in-house to assess the sodium pool, average gas and wall temperatures, pressures, and aerosol characteristics. Results from the experiment indicate spatial heterogeneity in the aerosol concentration across different elevations of the MINA test chamber, and there is a need for CFD-based assessment. The lumped analysis agrees well with the average temperatures, median diameters, and mass concentration. The maximum concentration of sodium aerosols in the chamber is approximately similar to 3.2g/m3. The median diameter of the initial aerosol increases from 0.2 mu m to 5.5 mu m after 50 min of sodium fire. The median diameter of the aerosol decreases to 3 mu m after 180 min of the sodium fire. The details of the experiments and significant findings are discussed in the present manuscript.

Nuclear reactors generate high levels of radiation, including neutron radiation, which poses significant risks to human health and structural integrity. Effective neutron shielding materials are essential to ensure safety in nuclear facilities by attenuating neutron flux and reducing radiation exposure. This study presents the development and performance assessment of boron carbide (B4C)-enhanced concrete designed for structural integrity and neutron shielding in nuclear reactor applications. B4C, known for its high neutron absorption cross-section, was used as a partial fine aggregate replacement at 5%, 10%, and 15% by weight. To improve durability and mitigate potential strength reduction, 20% fly ash was used as a cement substitute. The mechanical and durability properties were evaluated through compressive strength test, water absorption, sorptivity, chloride permeability, fire resistance, and shrinkage testing. The optimized mix, which contained 15% B4C and 20% fly ash (CFB), had a 28-day compressive strength of 38.65 MPa, low water absorption rate of 1.58%, and significantly improved resistance to chloride ion ingress. Neutron attenuation experiments were conducted at the KAMINI research reactor, demonstrated a reduction in thermal neutron flux by over two orders of magnitude using 19 cm of B4C concrete, equivalent to more than 35 cm of conventional concrete. Fast neutron flux was reduced by over 35 times at an energy level of approximately 600 keV. Gamma attenuation remained comparable to standard concrete, ensuring multipurpose shielding capacity. These results demonstrate that B4C combined with fly ash provides a sustainable, high-performance concrete solution capable of effective neutron shielding while maintaining mechanical integrity, offering considerable advantages for radiation-prone infrastructure.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300 mu Sv/h and 175-200 mu Sv/h, respectively, when the reactor is operating at 10 kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178 keV. Apart from energy reduction, 10 kW normalised neutron flux of south beam pit is also attenuated by the shield by 25 000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

An investigation of inlet blockage in the central subassembly (SA) of a medium-sized sodium-cooled fast reactor (SFR) is carried out. To analyze the accident scenario, a sodium boiling model is developed based on two-fluid approach. The model is added to the thermal model based on the single-pin assumption developed in the previous study. The developed model is validated with CABRI BI1 loss of flow experimental data. Having established the validity of the model, a detailed analysis of inlet blockage in the central SA of a medium-sized SFR is carried out. The model is able to predict the accident detection time by temperature monitoring thermocouple located at the SA outlet. Boiling front evolution, temperature evolution of the core components, dryout time, and clad melting time are predicted. A parametric study on the effect of growth of blockage at the SA inlet on time of detection and boiling is carried out.

An optimisation methodology is presented using the primary coolant circulation system of a nuclear reactor as its case study, the purpose of which is to find combinations of selected design and maintenance parameters to maximise the reactor safety and minimise monetary expenditure. The parameter space was sampled by a Monte Carlo method and Petri net modelling was used to predict the performance of each of these options. The optimal solutions were then extracted from the data via computation of the Pareto front, with further analysis conducted on parameter sets of interest.

The reactor Point Kinetics Equations (PKE) are simpler zero-dimensional approximation to space dependent dynamical models of nuclear reactor core, that are accurate enough to describe transients in small to medium size fast reactor cores. Even these simplified equations can be solved only by numerical methods, except in a very few restrictive cases, where they are amenable to analytical solution. Symmetry methods using Lie's point symmetry have been shown to be a systematic and powerful tool to solve any given ordinary or partial differential equation. An approximation of the PKE, known as Prompt Jump Approximation (PJA) converts the coupled system of first order ODEs with one delayed-neutron precursor group and power feedback, into a single first order nonlinear ordinary differential equation. In this study, we demonstrate an application of Lie symmetry method for solving the point kinetics equation under PJA. The analytical solution obtained is compared with benchmark numerical solution of PKE with PJA.

With the advent of new processing technologies, electronics have become more susceptible to interference, making fault detection an essential component of devices to ensure their functional integrity. Ensuring functional integrity is a critical requirement for safety in applications like nuclear instrumentation, automotive electronics, aerospace systems, industrial automation, and medical devices. One of the prominent fault detection mechanisms used in processor-based devices is the dual lockstep structure. In this paper, we present the design and implementation of a dual lockstep processor based on the open source SHAKTI C-Class processor. The design is simulated and implemented on a Field Programmable Gate Array (FPGA), and the results demonstrate its effectiveness in fault detection, ensuring functional safety. The synthesized results are compared with those of the base processor for area and performance.

A pseudo-bond graph is presented to model the heat transferred from the fuel rods to the coolant via its cladding in a generic nuclear reactor case study. Simulations performed using this model are used to explore the temperatures of the core components under ordinary and emergency scenarios, considering various conditions of coolant supply and reactor power output. The model is combined with a timed stochastic Petri net to produce a hybrid model, in which the reactor operation and fault status is determined by the Petri net and fed into the bond graph to examine the resulting impact on core temperatures, which in turn are fed back into the Petri net process. The results predict the distribution of the reactor operational durations before a disruption occurs. The model provides the temperature profiles attained by the cladding and fuel components, indicating a low probability of dangerous temperatures.

The article explores a fire plume mitigation strategy within a horizontally vented enclosure, potentially applicable in nuclear plants, ship cabins, and building ventilation systems. The investigation consists of a compartment with dual horizontal ceiling vents operating under both forced and natural ventilation conditions, featuring a centrally located fire source. The study models the turbulent three-dimensional flow and heat transfer characteristics using a Large Eddy Simulation model. Parametric investigations vary the fire size, location, and forced ventilation velocity through the horizontal vent. The findings reveal various flow regimes, from pure natural convection to a mixed convection regime of bidirectional flow phenomena. Results show that temperature reduction of 83% in the overall ceiling gas temperature for the case of 3 m/s ventilation velocity. It is also observed that with rise in the ventilation velocity, the Froude number increases and for higher heat release rates, the plume temperature increases and hence the Froude number is decreased by 42%. The fire sources mounted adjacent to the wall and beneath the naturally vented horizontal vent demonstrate the maximum fire plume velocity of 2 m/s. The heat transfer analysis suggests that the average temperature ratio for the fire source of 0.1 m diameter case is dropped by 30% in the case of maximum ventilation velocity of 3 m/s and whereas for the case of 0.141 m diameter fire source, the 40% drop in the average temperature ratio is observed in this study. The proposed correlation, which links the dimensionless mass flow rate through the vent with the Froude number, provides insights into hybrid ventilation phenomena. This study enhances the understanding of airflow and safety design in buildings, particularly concerning smoke control.

The cover gas space is an inert isolation layer provided for sodium systems in sodium-cooled fast reactors. During normal reactor operation, sodium aerosols are generated continuously in the cover gas space. Understanding the complex dynamics of the evolution and transport of the aerosol is essential from the perspective of the reactor operation. Such assessments provides vital insights into deposition patterns of aerosols to the components mounted on the roof slab. In the present manuscript, the evolution and transport of aerosol in the cover gas space as well as in roof-slab annular gaps are studied in detail with the help of computational fluid dynamics tool. The present model is validated against the experimental data from the literature. There is good agreement between temperature variation, aerosol number and mass concentration across the cover gas height. Post validation, the study of thermal and aerosol transport in the full-scale reactor cover gas for a medium-sized reference reactor is carried out. It is observed that aerosol sizes greater than ∼31 µm are mostly concentrated near either the sodium pool surface, component wall or near the vessel boundary. It is found that the average mass concentration in the cover gas space is uniform (∼ 29 g/m3). However, the annular regions are found to have a non-uniform distribution of aerosols with heavier particles confined to the lower annular regions in wavy like patterns having the same CMD as in the bulk cover gas space. The CMD in the top annular regions is ∼ 2 µm.

The Unprotected loss of flow accident (ULOFA) is a severe accident that may lead to core disruption in a sodium cooled fast reactor (SFR). It is investigated as a part of the defense in depth concept in nuclear safety. A coupled thermal hydraulics neutronics model is required to model the severe accident scenario where reactivity is affected by the changes in the temperature of the core components, coolant boiling, and material distribution during the transient. The coupled code previously developed for the analysis of total instantaneous blockage (ASTRA) is improved for the study of ULOFA by adding various models, viz., (i) a one-dimensional two-fluid sodium boiling model, (ii) a simple primary hydraulics model, (iii) a simple clad motion model, and (iv) an improved point kinetics solver. The boiling model is validated with different loss of flow experiments. The improved code is used to analyze ULOFA in a 500 MWe medium size sodium cooled fast reactor up to the onset of fuel melting. At 21.5s, coolant boiling is initiated in the central channel. The temperature evolution of the core components, reactivity feedbacks, power evolutions, and coolant voiding propagation are calculated. The code predicts early fuel melting at 24.9s due to the power excursion by the early voiding in the fuel channels compared to the study without including the boiling model. Boiling is started in five out of ten representative fuel subassembly (SA) channels at the time of fuel melting.

In this study we propose a simple, yet inherently fault-tolerant and robust controller utilizing a combination of both feedforward and feedback control schemes. Feedback controllers are known to be robust, but feedforward controllers can be reliable in the context of safety as they can be less dependent on measurements. Though, many combined feedforward and feedback controllers have been proposed in the literature, to the best of our knowledge a controller tolerant to the failure of power feedback signal is nowhere demonstrated. In this scheme, major corrections are effected by a model predictive forward control unit, while bounded uncertain part is corrected by a feedback control unit. The feedforward control is implemented using the inverse equations of the nuclear reactor point kinetics model. The bounded error control is effected by a simpler limited Proportional–Integral–Derivative (PID) feedback controller. Stability analysis of the controller is carried out numerically by Lyapunov’s direct method with the help of Particle Swarm Optimization (PSO) technique. The proposed controller (referred to as Inverse Dynamics Corrected Control (IDCC)), is studied for reactor power control with model error plus uncertainty and a principal feedback signal failure case ( i.e. power sensor failure). It is shown that the IDCC has excellent load tracking performance for a challenging demand profile with model error, in comparison to Sliding Mode Controller (SMC) and a manually tuned PID controller. For the loss of principal feedback signal case, there is a trade off between the tracking performance of IDCC and safety aspects. While both PID and SMC drive the reactor power to unsafe levels, IDCC maintains the reactor power within a safer bound.

Proper estimation of neutron source strength from external source subassembly of PFBR is very important. The present work uses the radiation transport simulation toolkit GEANT4 to do the same by modelling full geometric details of the source. Simulations have been carried out to first determine the gamma source strength from activated Antimony isotopes in the subassembly. This is followed by transport of these Antimony decay gamma photons to estimate photo-neutron production yield. The calculated results are in good agreement with earlier calculations, indicating feasibility of GEANT4 code for use in FBR calculations.

Passive safety systems help to improve overall plant safety, reliability, and resilience. However, the real gain in the use of passive systems depends on the robustness of the design utilizing the passive process and the role of active elements, if any. In this paper, we propose a fan-controlled sodium-to-air heat exchanger (AHX) system design for a failsafe and passive decay heat removal (DHR) function in a pool-type sodium-cooled fast reactor. The proposed system uses a fan to control air flow and minimize heat loss during normal operation, and when the fan trips due to loss of power or a trip signal, DHR gets enabled in a failsafe mode. The system is analyzed with the help of a simplified one-dimensional model as well as with detailed computational fluid dynamics software. It is found from analysis that it is possible to control and maintain the air flow to about 4% to 5% of full flow, as in the case of conventional dampers, to minimize heat loss during normal reactor operation. The reliability of the proposed system is also analyzed and shows that the fan-controlled AHX-based decay heat removal system (DHRS) has a much better reliability compared to the conventional passive DHRS with active damper-dependent operation.

In sophisticated and complex system such as nuclear power plant, fault estimation and fault tolerant control always play an important role in maintaining the system stability and assuring satisfactory and safe operation. Thus, in this work a fault estimation and fault tolerant control scheme based on sliding mode theory is proposed for a pressurized water reactor type nuclear power plant considering simultaneous actuator and sensor faults. First, using descriptor sliding mode observer approach, an accurate estimation of the system states and sensor fault vector have been obtained simultaneously. Then, based on the estimated information, an integral type sliding mode control scheme is proposed to stabilize the resulting faulty system. With the help of Lyapunov stability theory, reachabilities of the proposed sliding mode surfaces are shown in both the state estimation space and the error estimation space, simultaneously. Finally, the efficacy of the proposed control scheme is shown by applying it to a nuclear power plant.

In the nuclear power industry, safety and reliability are of the utmost importance. Sensors and actuators are integral components in such systems, and potential faults may adversely impact system performance. It is therefore imperative to design a fault detection and diagnosis (FDD) system that achieves the highest standards of safety. This paper presents a machine learning-based fault detection and diagnosis (FDD) technique for actuators and sensors in a pressurized water reactor (PWR). In the proposed FDD framework, faults are first detected using a shallow neural network. Second, fault diagnosis is performed using 15 different classifiers provided in the MATLAB Classification Learner toolbox, including support vector machine (SVM), K-nearest neighbor (KNN), and ensemble. Several classifiers were found to provide superior classification performance, including medium KNN, cubic KNN, cosine KNN, weighted KNN, fine Gaussian SVM, quadratic SVM, medium Gaussian SVM, coarse Gaussian, bagged trees, and subspace KNN. The accuracy of the FDD approach was demonstrated using a set of simulation results.

This study proposes a feedback linearization-based control using a dynamic neural network to control a pressurized water-type nuclear power plant. The nonlinear plant model adopted in this study is characterized by five inputs, five outputs and, 38 state variables. The model is linearized through dynamic neural network-based system identification and feedback linearization. The proportional-integral-derivative (PID) controller is subsequently applied to the linearized process. The effectiveness of the proposed approach is demonstrated by simulations on different subsystems of a pressurized water reactor nuclear power plant model. Simulation results show that the proposed strategy offers good performance and is capable of effectively tracking the reference under disturbances.

Passive safety systems help to improve overall plant safety, reliability and resilience. In this article, we propose a fan controlled sodium to air heat exchanger system design for fail-safe and passive decay heat removal (DHR). The proposed system uses a fan to control airflow and heat removal during normal operation, and when fan trips due to loss of power or a trip signal, DHR gets enabled in a fail-safe mode. The system is analyzed with the help of a 1D model as well as Computational Fluid Dynamics software. It is found from analysis that it is possible to control the air flow to about 5 to 10\% of full flow as in the case of conventional dampers to minimize heat loss during normal reactor operation. Reliability of the proposed system is also analyzed and shown that the proposed system has a much better reliability compared to the conventional passive DHRS.

A nuclear reactor is expected to function for extensive periods, during which, coolant circulation and core reactivity must always be maintained safely. Understanding the risks associated with the operation of such systems requires proper consideration of ageing components and the effects of preventative maintenance. The traditional methodologies, such as Fault Trees and Event Trees, have limitations in their abilities to model ageing processes and complex maintenance strategies. Petri Nets have been used in this research as a more suitable alternative. A case study reactor is presented to demonstrate this capability. Petri Nets were developed for five key subsystems: primary coolant circulation, shutdown condensation, emergency core coolant injection, emergency shutdown, and control and monitoring, building a representation which considers their failure modes, reaction of the system to faults, and ongoing component maintenance actions. These models reveal statistics for the timing of failure of these subsystems and relative frequencies of outcome categories. (c) 2021 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY license (http:// creativecommons.org/licenses/by/4.0/).