Pressurized Water Reactor Research Articles

Atypical Attack of the Dissimilar Metal Welds of a Water–Water Energetic Reactor Nuclear Power Plant Secondary Circuit: Possible Mechanisms of Failure

ABSTRACTDissimilar metal welds are utilized in the energy industry to connect two materials with different material characteristics. In the case of nuclear power plants, the connected materials tend to be low‐alloyed steel and high‐alloyed material. Despite different material and corrosion properties, under the proper environmental conditions, the used construction materials and weld metals are protected either by a passive layer or by a high‐temperature oxide. Although dissimilar metal welds are used in both primary and secondary circuits, the most frequently documented damage is in the secondary circuit, where, in addition to material heterogeneities, local environmental heterogeneities may form. For dissimilar metal welds in WWER nuclear power plants, a water–water energetic reactor, a subtype of pressure water reactor, we note two main types of attack near the dissimilar fusion boundary: on the carbon steel side or on the high‐alloyed weld metal side (X10CrNiMoN16‐25‐6). The possible causes of the latter “atypical” corrosion attack are debated and can be generalized as a consequence of change of the grain boundary condition.

System-level Prognostics and Health Management for Complex Industrial Systems

Prognostics and health management (PHM) has become essential to guarantee aware and safe system operation and to inform economic decision-making. However, due to the nature of detection, diagnostics, and prognostic methods, applications have mainly been limited to the component level. In practice, most industrial systems consist of multiple interacting components whose partial degradation could lead to the system's failure (or subsystems).This research addresses the limitations of traditional component-level PHM techniques by proposing a novel system-level framework. By implementing a hierarchical structure of components and subsystems, we will select an optimal method for each subsystem to aggregate its component health assessments. The overall system health can then be estimated by further combining the obtained estimates. The research considers simplified and holistic modeling techniques, margin-based methods, and hybrid graphical models. This approach aims to provide reliable system health predictions and online components' sensitivity measures to enhance maintenance decision-making. We consider an application in the context of the nuclear industry, characterized by strict safety and economic requirements. Using a SIMULINK model to approximate a Pressurized Water Reactor (PWR) with real industrial inputs, we plan to add component degradation modules and use simulated sensor data and reliability information to test the proposed framework. Initial results on artificial case studies show the feasibility of integrating component-level health predictions.

Research on Reactivity-Equivalent Physical Transformation Method for Double Heterogeneity in Pressurized Water Reactors Based on Machine Learning

Traditional computational methods for pressurized water reactors are unable to handle dispersed fuel particles as the double heterogeneity and the direct volumetric homogenization can result in significant errors. In contrast, reactivity-equivalent physical transformation techniques offer high precision for addressing the double heterogeneity introduced by dispersed fuel particles. This approach converts the double heterogeneity problem into a single heterogeneity problem, which is then subsequently investigated by using the conventional pressurized water reactor computational procedure. However, it is currently empirical and takes a lot of time to obtain the right k∞. In this paper, we train the RPT model by using the existing dataset of plate-dispersed fuel and rod-dispersed fuel by a machine learning method based on a linear regression model, and we then use the new data to make predictions and derive the corresponding similarity ratios. The burnup verification, density verification, fission rate verification, and neutron energy spectrum analysis are calculated through the OpenMC program. For plate-type fuel elements, the method maintains an accuracy within 200 pcm during depletion, with deviations in the 235U density and 235U fission rate within 0.1% and neutron energy spectrum errors within 6%. For rod-type fuel elements, the method maintains an accuracy within 100 pcm during depletion, with deviations in 235U and 239Pu density within 1.5% and neutron energy spectrum errors within 1%. The numerical validation indicates that the reactivity-equivalent physical transformation method based on the linear regression model not only greatly improves the computational efficiency, but also ensures a very high accuracy to deal with double heterogeneity in nuclear reactors.

Matching Analysis of Technical Parameters and Safety Standards for Nuclear Replacement of Coal-Fired Units

In the context of the growing share of renewable energy and the impending decommission of a large number of coal-fired units, nuclear energy is the only green energy that can replace coal power for a stable, clean, efficient, and large-scale power supply. This article compares the differences between coal power and nuclear power in terms of thermal system, thermal cycle, turbine parameters, and safety. It discusses the possibility of replacing the boiler of a coal power plant with nuclear power, that is, replacing the boiler of a coal power plant with a nuclear reactor for generation/heating/cogeneration. For coal-fired units with similar capacity that do not use a reheat cycle (at or below high pressure) and nuclear power units (such as a high-temperature gas-cooled reactor), as well as coal-fired units with a reheat cycle (ultra-high pressure and above) and nuclear power units (such as pressurized water reactors), there are great differences in steam parameters. In terms of steam turbines, the size of nuclear power units is relatively larger, requiring additional dehumidification measures. In addition, the safety factors and management methods considered in the site selection, construction, and operation of nuclear power plants are more stringent and complex, and comprehensive analysis and evaluation are needed in aspects such as waste treatment and accident emergency response. Except for the relevant provisions of the American Society of Mechanical Engineers code for pressure vessels, nuclear power units are not compatible with coal-fired units in terms of safety standards. Therefore, considering comprehensively, it is believed that the scheme of nuclear power replacing coal-fired units for power generation/heating/cogeneration program is not feasible at present.

Data-Driven Remaining Useful Life Estimation

This paper explores the development and application of data-driven prognostic models for estimating the Remaining Useful Life (RUL) of Nuclear Power Plant (NPP) condensers experiencing tube fouling. Due to the unavailability of run-to-failure industry sensor data, we utilized simulated data generated by the Asherah Nuclear Power Plant Simulator (ANS), initially designed by the International Atomic Energy Agency (IAEA) and programmed in Simulink for cyber security simulations. ANS's adaptability allows it to simulate Pressurized Water Reactor (PWR) behaviors given a time series of operating conditions and to introduce degradation modules to mimic fouling effects. Our study compares two primary approaches applied to data generated by ANS: inference-based and direct prediction methods. The selected inference-based approach estimates the health state of the condenser using a pipeline formed by an Auto Associative Kernel Regressor and a Hidden Markov Model (HMM), which subsequently combines the state estimates with its parameters to predict the RUL. The direct prediction method employs a Gradient Boosting Regressor Decision Tree (GBRDT) to map input variables directly to RUL. Our findings demonstrate the efficacy and limitations of each method through the case study, providing valuable insights for the adoption of data-driven RUL estimation techniques in industrial and energy applications.

Online and Offline Fault Detection and Diagnostics in a Nuclear Power Plant Condenser

Nuclear power plants (NPPs) face significant financial pressures due to operational and maintenance costs. This research investigates Fault Detection and Diagnostics (FDD) techniques to optimize maintenance schedules and reduce expenses. The NPP condenser plays a critical role in converting turbine exhaust steam back into water for reuse. Condenser tube fouling, a prevalent fault mode, impedes heat transfer efficiency and can lead to decreased plant efficiency and safety risks. This study proposes an FDD framework that leverages raw signal analyses from temperature and pressure monitoring to detect and diagnose condenser tube fouling in both online and offline settings. The online approach facilitates close-to-real-time predictions, enabling proactive maintenance strategies. Additionally, the framework explores incorporating a condenser’s maintenance history for enhanced diagnostics. We employ a dataset obtained from a simulated nuclear power plant condenser using the Asherah Nuclear Power Plant Simulator (ANS). ANS replicates the operational dynamics of a pressurized water reactor (PWR) type NPP. The proposed methodology utilizes an encoder-decoder (E-D) structured 1DCNN model to analyze the raw signals. The research demonstrates consistent and accurate fault detection and diagnostics for condenser tube fouling in both online and offline scenarios. A high potential for generalization to unseen conditions was observed. However, online detection using small data windows necessitates caution due to potential false alarms around the transition points. Our findings pave the way for further exploration of robust diagnostics by accommodating a wider spectrum of fouling rates within degradation classes using ANS. This combined online and offline FDD approach offers a promising solution for promoting operational safety, efficiency, and cost-effectiveness in NPP condensers.

Analysis of the influence of bottom flow holes in control rod guide tubes on flow field and control rod displacement

During the operation of pressurized water reactors, flow-induced vibration issues are commonly encountered, where control rod components vibrate due to the impact of coolant flow, leading to wear, thereby posing a threat to the safe operation of the reactor. This paper conducts a detailed numerical simulation of the flow field and force distribution inside a full-scale lower control rod guide tube based on the CFD method. The fluid–structure interaction analysis is performed to reveal the influence of cross-flow under various operating conditions on the vibration and displacement of control rods. The research subject consists of 24 control rods, one continuous guide section, and six control rod guide cards. The lower part of the guide tube has two flow holes on each surface, with one assumed to be a cross-flow inlet and the others as outlets. The results indicate that when considering the cross-flow effect at the flow holes, 90 % of the coolant flows out from other flow holes at the bottom of the guide tube, resulting in increased transverse velocity at the bottom and the presence of numerous vortices. The magnitude and location of the force exerted on the control rod by coolant impact are correlated. The control rod closest to the inlet of the flow hole experiences the greatest force, approximately 5.35 times the average at other positions. At a cross-flow velocity of 0.4 m/s and a low frequency of 10 Hz, the maximum displacement of the control rod is 0.3355 mm.

Enhancing Thermal–Hydraulic Performance in Nuclear Reactor Subchannels with Al2O3 Nanofluids: A CFD Analysis

In this paper, the performance of aluminum-based nanofluids with a possible application in pressurized water reactors is numerically investigated. A 605 mm long 4-rod array square (2 × 2) subchannel geometry with a uniform heat flux of 50 kW/m2 has been used in CFD simulation. This analysis has been carried out using the RNG k-epsilon turbulence model with standard wall function in ANSYS FLUENT 2022R1. The impact of various flow conditions and nanofluid concentrations has been examined. The effects of variable velocities on nanofluid performance have been studied using different Reynolds numbers of 20,000, 40,000, 60,000, and 80,000. The analysis was conducted with Al2O3/water nanofluid concentrations of 1%, 2%, 3%, and 4%. A comparison of the Nusselt number based on five different correlations was conducted, and deviations from each correlation were then presented. The homogeneous single-phase mixer approach has been adopted to model nanofluid characteristics. The result shows a gradual enhancement in the heat transfer coefficient with increasing volume concentrations and Reynolds numbers. A maximum heat transfer coefficient has been calculated for nanofluid at maximum volume concentrations (ϕ = 4%) and highest velocities (Re = 80,000). Compared to the base fluid, heat transfer was enhanced by a factor of 1.09 using 4% Al2O3. The Nusselt number was calculated with a minimal error of 3.62% when compared to the Presser correlation and the maximum deviation has been found from the Dittus–Boelter correlation (13.77%). Overall, the findings suggest that aluminum-based nanofluids could offer enhanced heat transfer capabilities in pressurized water reactors.

Evaluation of material accountancy techniques for 233Pa from thorium nuclear fuels

Thorium is a promising alternative to uranium as nuclear fuel with advantages such as higher abundance, lower production of long-lived transuranic elements, and potentially better proliferation resistance. However, thorium presents a potential pathway for proliferation where produced 233Pa can be diverted for the clandestine production of safeguarded 233U. To prevent this, the ability to detect and measure 233Pa must be assessed. This paper reviews several nuclear material accountancy techniques to determine their suitability for detecting 233Pa extracted from irradiated thorium fuel. Hybrid K-edge densitometry and passive gamma spectroscopy have been found to be the best options based on technology maturity, cost, accuracy, and acquisition time. Thorium can be used in various reactor designs such as pressurized water reactors (PWRs), Canada deuterium uranium (CANDU) reactors, and molten salt reactors (MSRs). Therefore, thorium-uranium oxide fueling was modeled for three representative reactors (PWR, CANDU, MSR), burning the fuel to 47 GWd/MTHM for PWR, 19 GWd/MTHM for CANDU, and at a steady power of 52.711 MW/MTHM for MSR. Within each model, the protactinium element in the used fuel was extracted and its isotopic content analyzed. Simulated results indicated that 233Pa can be detected using passive gamma spectroscopy in each fuel type at all decay times (0–300 days) following separation.

Discerning the effect of various irradiation modes on the corrosion of Zircaloy-4

Using proton irradiation, this study investigates the individual influence of several factors on the corrosion kinetics of Zircaloy-4 in a hydrogenated water environment simulating a Pressurized Water Reactor (PWR). Using both simultaneous irradiation-corrosion and autoclave corrosion, we separately examine (i) the effect of pre-irradiation on modifying the structure of the material, (ii) the impact of irradiation on creating defects in the growing oxide layer during corrosion, and (iii) the influence of irradiation on increasing the corrosion potential through radiolysis during corrosion. To replicate the neutron-irradiated microstructure, two proton pre-irradiation schedules were employed: Schedule 1 (isothermal irradiation at 350°C to 5 dpa) to simulate high-temperature PWR conditions, and Schedule 2 (two-step process: irradiation to 2.5 dpa at -10°C followed by 2.5 dpa at 350°C) to simulate lower temperature PWR and Boiling Water Reactor (BWR) conditions. Long-term autoclave corrosion testing for 360 days at 320°C revealed no significant difference between unirradiated samples and those pre-irradiated according to either schedule, with all samples exhibiting sub-cubic kinetics within the pre-transition regime. Pre-irradiated samples underwent Simultaneous Irradiation Corrosion (SIC) tests, corroding in 320°C water while being irradiated with protons. Corrosion was found to accelerate in all SIC-tested samples relative to autoclave conditions, with the greatest increase observed in non-pre-irradiated regions of the samples. Pre-irradiation with either schedule resulted in a slower corrosion rate compared to non-pre-irradiated regions under SIC conditions. The degree of radiolysis observed in the SIC tests surpassed typical PWR conditions, approaching levels found in BWRs. Radiolysis products were identified as a primary contributors to accelerated corrosion, corroborated by radiolysis bar tests. These findings underscore the intricate interactions between irradiation, corrosion, and water chemistry in determining Zircaloy-4 corrosion kinetics within nuclear reactor environments.

Influences of Fluorine on Corrosion Behavior of FeCrAl Alloy in Lithiated/borated Water at 360 °C/18.5 MPa

Abstract As an accident-tolerant fuel cladding material, Iron-Chromium-Aluminium (FeCrAl) alloy shows a big potential in replacing zirconium alloy as the cladding tube for pressurized water reactors. The presence of fluorine impurity in the primary coolant of pressurized water reactors will accelerate the corrosion of fuel cladding tubes. However, studies on the influences of fluorine impurity on the service performance of FeCrAl alloy are relatively limited. In this work, corrosion behavior of FeCrAl alloy in fluorine-containing lithiated/borated water at 360 °C and 18.5 MPa was characterized by Scanning Electron Microscope (SEM), Energy Dispersive Spectrometer (EDS), X-ray diffractometer (XRD), and Raman spectroscopy. The results indicate that FeCrAl alloy exhibited relatively good resistance to fluorine attack. Fluorine ions changed the diffusion behavior of alloying element Cr, so facilitating the formation and dissolution of FeCr2O4. Deposition of Ni and B was observed on the alloy surface. The maximum content of B reached 24.20 at.% after fluorine-containing lithiated/borated water exposure while that was 12.50 at.% in the absence of fluorine ions. Despite this, no specific compounds of B were detected in the corrosion products, but Ni existed in the form of NiFe2O4.

Review of Electrical Parameters Influence on Characteristics of Plasma Electrolytic Oxide Coating on Zircaloy

<p>Zircaloy-4 (Zr-4) is used as a fuel cladding material in Pressurized Water Reactor (PWR). Zr-4 as a cladding material works in extreme conditions in pressurized water up to 150 atm at 325 ˚C. In addition, the refuelling process in the reactor requires surface protection of the clay material to minimize corrosion and wear. One of the raising methods to enhance the corrosion resistance of the Zr-4 is by plasma electrolytic oxidation (PEO). Characteristics of the ceramic oxide layer produced by PEO are influenced by current density, type and composition of the electrolyte, and voltage mode. One of the challenges in the PEO development on the Zr-4 substrate is a high porosity with a range of 5%-20% and the low number (below 6%) of t-ZrO<sub>2</sub> phases in the inner and outer layers. Optimizing the electrical parameters is necessary to overcome this problem. The results of the literature study show that the cathodic current at the AC voltage plays an important role in determining the resulting plasma characteristics. Low duty cycle (cathodic current> 50%) produce plasma with high density, resulting in a low porosity layer. Oppositely, high duty cycle (cathodic current <50%) produced high content of t-ZrO<sub>2</sub> increase the mechanical resistance. Two-step PEO is beneficial in combining the low and high duty cycle to obtain the benefit of each step. </p>

An application of the DEM-FFT method to predict the thermal conductivity of high burnup fragmented fuel

Understanding Fuel Fragmentation, Relocation and Dispersal (FFRD) phenomena is a major item to predict the behavior of fuel rods in the event of a Pressurized Water Reactor loss-of-coolant hypothetical scenario. In order to introduce an effective medium for fragmented fuel that could be used in fuel performance codes, the heat transfer properties of UO2 granular beds in He-Kr-Xe mixes was studied numerically. The Discrete Element Method was first used to compute the spatial arrangement of powder only submitted to gravity with a representative granulometric distribution. These generated geometries were then transformed into a 3D grid (voxelized) to perform numerical homogenization and compute their effective properties using a Fast Fourier Transform solver for heat transfer. Different hypotheses were made in the voxelation process about how the treatment of gas-solid and solid-solid interfaces, which in turn provided different estimates of the effective properties. Due to limits on the discretization and extended particle size distributions, a two-scale scheme was introduced and numerically tested against direct computations. For several beds, computed thermal conductivities were compared to analytical models for granular materials. Some heat transfer properties of fragmented fuel at several burnups were then computed, taking into account approximations for Knudsen and radiation size effects.

Preliminary Implementation of High-Resolution Multi-Scale coupling calculations for the entire pressure vessel based on OpenFOAM

Significant coupling effects exist among system components in nuclear pressure vessel. Due to the complex geometric structures, the nuclear industry primarily relies on system codes or sub-channel methods for core safety analysis. However, these methods suffer from low model accuracy and insufficient coupling capabilities. Additionally, the differences in model scales impede direct coupling analysis with the CFD calculations of the plenum system. To address these issues, this paper proposes a multi −scale coupling calculation method for the entire pressure vessel: For the plenum system, detailed CFD modeling is employed, while the core calculations are conducted using CorTAF, a high-resolution core Multiphysics-coupling analysis method developed by our team. A cross-resolution coupling model is utilized to integrate the two, achieving cross-resolution coupling simulations for the entire pressure vessel, encompassing both the plenum and core system. The above method was applied to the coupling calculations of a typical pressurized water reactor’s lower plenum and core, revealing detailed thermal–hydraulic phenomena under precise core flow inlet distribution conditions. The lateral flow at the core inlet exceeds 1 m/s, with the maximum and minimum fluid velocities in the subchannels deviating by up to 70 % from the average velocity of 2.42 m/s. The flow distribution only begins to stabilize after a height of 1.2 m in the core. The paper also includes inlet asymmetric flow reduction calculations. Overall, the method enables multi-scale and multi-physics coupling calculations, which provide significant reference value for improving the accuracy of current core safety analyses.

Influence of composition and temperature on oxide film formation for Fe-Cr-Ni alloys in simulated PWR primary water

ABSTRACT The growth of oxide films on Fe-Cr-Ni alloys in simulated pressurized water reactor (PWR) primary system environments was investigated, with focus on the influence of alloy composition and temperature. Oxide film thicknesses were measured for 90 specimens prepared from compact tension (CT) test pieces. Cross-sectional observations were also conducted on coupon specimens using scanning transmission electron microscopy (STEM). The behavior of Ni-based alloys, specifically Alloy 600, and Fe-based alloys, specifically Type 316 stainless steel, were compared. Ni-based alloys exhibit superior corrosion resistance due to slower Fe diffusion kinetics compared to Fe-based alloys, leading to the formation of a nickel-enriched protective layer at the metal-oxide interface. The oxidation rates of Ni-based alloys follow an Arrhenius-type temperature dependence. In contrast, Fe-based alloys show a more complex temperature-dependent behavior, with reduced oxidation rates at higher temperatures that are likely due to changes in oxide dissolution. Higher chromium concentrations in both alloy types improve the protectiveness of the oxide film towards corrosion by forming a more compact inner oxide layer. The study identifies three primary factors influencing oxide film growth: reactions at the metal-oxide interface, diffusion within the oxide layer, and processes at the oxide-solution interface. The relative significance of these processes varies depending on the alloy composition and temperature.

Prediction of Decay Heat from PWR Spent Nuclear Fuel Using Fuel Parameters

In the context of a geological repository for nuclear waste, fast and accurate predictions of decay heat are needed for different applications ranging from canister loading optimization to comparing decay heat predictions from state-of-the-art codes with experimental measurements. This work uses a large database of simulated pressurized water reactor (PWR) spent nuclear fuel (SNF) with an extensive range of fuel parameters to demonstrate that by using only the burnup, initial enrichment, and cooling time of the SNF, it is possible to predict the decay heat of a PWR SNF. A linear interpolation model has been developed using the simulated data and tested on data from decay heat measurements using a calorimeter. The model code was also made publicly available [V. Solans, “Python Script for the Prediction of Decay Heat from PWR Spent Nuclear Fuel Using Fuel Parameters,” Zenodo (2024)]. The results show that the decay heat can be well predicted, with the relative error between measurements and predictions ranging between 4% and 8%. After correcting for a systematic deviation between predictions and experimental results using the limited set of experimental measurement data available, the relative error can be further reduced to 2% to 3%.

Generating optimal reloading patterns for CNPGS Unit-3 core using multi-objective elitist teaching-learning-based optimizer

Pressurized Water Reactors (PWRs) management presents the core reloading pattern optimization as a significant problem that contributes to the improvement of reactor productivity and optimal fuel utilization with considering safety conditions. In present study, Multi-Objective Elitist Teaching-Learning-Based Optimization (MO-ETLBO) technique is suggested to cope up the multi-objective reloading optimization problem of the Chashma Nuclear Power Generating Station (CNPGS) unit-3 core. A multivariable objective function is designed to evaluate the quality of each loading pattern while maximizing critical boron concentration (CBC), minimizing the power peaking factor (PPF), and optimally enhancing the cycle length while ensuring adequate safety margins and design limits. It has been found that the equilibrium cycle can be extended to 16.07 extended full power days (EFPDs) while maintaining the PPF and CBC within design limits. To validate the effectiveness of TLBO, the optimized loading pattern of the equilibrium core was evaluated using the deterministic computer code DONJON5, for neutronic parameter analysis. The results show that the algorithm proposed in this study is a promising approach for reloading pattern optimization in CNPGS unit-3, offering potential improvements in reactor cycle length while ensuring safety and enhancing overall performance.

Detection of Axial Locations of Control Rods Using Movable In-core Detectors

Background Reactivity control in Pressurized Water Reactors (PWRs) relies on Rod Cluster Control Assemblies (RCCAs), or control rods (CRs), and boric acid diluted in primary circuit water. The control rods’ axial location inside the reactor core is typically provided by means of two redundant systems: the Rod Positioning Indicator (RPI) and the Rod Control System (RCS). Methods The current work focuses on the development of an additional method for the detection of control rod axial location inside the reactor core of a Pressurized Water Reactor (PWR) using the Advanced Flux Mapping System (AFMS) and its Movable In-core Detectors (MIDS). The paper presents the methodology employed and the results of the first validation campaigns. Results Two validation campaigns were carried out, targeting respectively equilibrium and non-equilibrium xenon axial distribution. The routine proved successful in identifying control rods’ axial insertion in different operating conditions including Flux Maps (FMs) acquired at 2%, 50% and 75% power. Conclusions The routine developed at Beznau Nuclear Power Plant will serve as an additional core monitoring tool to be employed in case of suspicious or faulty indication coming from the main control rod location methods. The developed method proved able to locate control rods’ axial insertion with few steps error, equivalent to maximum 1.6% of total fuel height (i.e. 5 cm as maximum error). There is room for potential improvements with regards to the uncertainty model currently implemented.

Design a robust intelligent power controller for pressurized water reactor using particle swarm optimization algorithm

Abstract Proportional–Integral–Derivative (PID) controllers have been optimized and used to overcome many types of problems in nuclear reactor systems. The high performance of PID controllers depend on optimizing their gains. In this research, an optimized robust PID controller is proposed to control power perturbations in a pressurized water reactor (PWR). The optimization process of robust PID using particle swarm optimization (PSO) algorithm aims to adapt PID gains then after that, H-infinity controller is used. The results show a good performance when that suggested hybrid controller is applied to the nuclear power system since the suggested design makes the system robust due to applying H-infinity method, in addition to get the benefits of the optimized PID controller.

An Experimental Study of an Autonomous Heat Removal System Based on an Organic Rankine Cycle for an Advanced Nuclear Power Plant

The present study focuses on the recovery of waste heat in an autonomous safety system designed for advanced nuclear reactors. The system primarily relies on passive safety condensers, which are increasingly integrated into the design of advanced Pressurized Water Reactors (PWRs). These condensers are typically immersed in large water tanks that serve as heat sinks and are placed at sufficient heights to ensure natural circulation. Such a heat removal system can operate for an extended period, depending on the size of the tank. This research is driven by the potential to recover part of the energy stored in the boiling water volume, using it as a heat source for an Organic Rankine Cycle (ORC) system via an immersed heat exchanger. The electricity generated by the ORC engine can be used to power the system components, thereby making it self-sufficient. In particular, a pump replenishes the water tank, ensuring core cooling for a duration no longer limited by the water volume in the tank. An experimental test setup, including a boiling water pool and an ORC engine with an electrical output of approximately several hundred watts, along with an immersed evaporator, was constructed at CEA (Grenoble, France). Several test campaigns were conducted on the experimental test bench, exploring different configurations: two distinct ORC working fluids, cold source temperature variation effects, and relative positioning of the submerged evaporator and heat source within the water tank impact. These tests demonstrated the reliability of the system. The results were also used to validate both the ORC condenser and evaporator models. This article presents this innovative system, which has recently been patented. Moreover, to the best of our knowledge, the investigated configuration of an ORC that includes an immersed evaporator is original.