Boiling Water Reactor Research Articles

Experimental investigation of thermal stratification development in boiling water reactor suppression pools during reactor core isolation cooling system operation

The Reactor Core Isolation Cooling (RCIC) system provides make-up water to the reactor pressure vessel during isolation events by passing steam from the vessel through a turbine which drives a pump to return water to the vessel. Exemplary performance of the RCIC system during the Fukushima Daiichi nuclear accident of 2011 has promoted interest in its potential performance.To investigate the development of thermal stratification within suppression pools due to RCIC system exhaust, a facility was constructed to produce the highest-resolution 3-D temperature measurements available in a large water pool during pool mixing by steam injection.Thermal stratification was most strongly influenced by suppression chamber pressure. Increased steam flowrate promoted mixing, which delayed the onset of thermal stratification. Containment response is discussed with respect to thermal stratification, which delays the time of warm water ingestion and cavitation of the RCIC pump, and pool mixing, which reduces the containment pressurization rate.

Comparative assessment of the environmental impact of nuclear power plant technology using life cycle assessment approach: A review

Nuclear power plants (NPP) are being considered as an alternative energy that could reducing greenhouse gas emissions. However, there are various challenges that need to be addressed regarding the variation of current nuclear technology such as environmental impact issues. To ensure the sustainability and effectiveness of nuclear power plants technology, a Life Cycle Assessment (LCA) approach can be used to study the environmental impacts of the nuclear power plant. LCA is a method for evaluating the environmental impact of a product or process from start to finish. By conducting a preliminary LCA study of each Nuclear Power Plant technologies, the environmental impacts of the stages of construction, operation, and management of radioactive waste can be understood. This study can also help compare the environmental impacts of nuclear power plants with other energy technologies. This preliminary study will review the environmental impact from pressurized water reactor, boiling water reactor, molten salt reactor, and compare with conventional coal power plant to provide compact summary of the technology. The results of the LCA study can help identify areas that require more attention in the development of nuclear power plants in Indonesia.

Analysis of main steam line break accident on a BWR test facility using TRACE

PUMA-E is an integral test facility, which is used for extensive experimental investigations to study the behavior of Boiling Water Reactor (BWR) plants under accident conditions. The TRACE code is the best estimate thermal hydraulic system code developed by the U.S. Nuclear Regulatory Commission for performing nuclear power plant safety analysis. The aim of the present work is to conduct a short validation study of the TRACE code applied to the PUMA-E facility. Firstly, a TRACE model of the PUMA-E facility was developed based on the existing RELAP5 input deck. Then, an additional facility TRACE model with reactor pressure vessel modeled by three dimensional “VESSEL” component was also established. After that, a Main Steam Line Break accident test was simulated, and the calculation results were compared with experimental data and RELAP5 simulation results. The TRACE simulation capability to predict the accident phenomenon of BWR test facility was evaluated using both qualitative and quantitative methods. On the basis of the above research, the base case accompanying with one Gravity Driven Cooling System tank failure accident was carried out based on the three facility models. The validity and reliability of the simulation results of the different facility numerical models were mutually verified through a qualitative comparison of the simulation results.

Oxidation of Alloy X-750 with Low Iron Content in Simulated BWR Environment

This paper presents an investigation of the oxidation of Alloy X-750 containing 5 wt% iron in a simulated boiling water reactor (BWR) environment. The specimens were exposed by a water jet (10 m/s) at 286 °C for durations ranging from 2 to 840 h, and the development of the oxide microstructure was mainly studied using electron microscopy. The results showed that the oxide scale consists of blocky crystals of trevorite on top of a porous inner layer rich in Ni and Cr. After the longest exposure time, the trevorite crystals completely covered the specimen surface. The study further revealed that the rate at which the oxide grew and the metal dissolved both decreased with time, and the metal thinning process appeared to be sub-parabolic. Given the significant variation in iron content in the X-750 specification, the influence of this element on the material’s corrosion performance in BWR was examined by comparing the results from this investigation with those from previous work on material containing 8 wt% Fe. The study demonstrates that the oxide growth, metal dissolution and metal thinning were slower in the material with a higher iron content, indicating the importance of this element in limiting the degradation of Alloy X-750 in BWR environments.

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

This paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO2. The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB2 Integral Fuel Burnable Absorber (IFBA) and B4C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd2O3-doped UO2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

3D reconstruction and interconnectivity quantification of the nano-porosity in the oxide layer of corroded Zr alloys

Nano-porosity development in thermally grown oxides plays a significant role in oxidation kinetics as nanopores may provide pathways for oxidizing species. As an example, Zr alloys are the most commonly used cladding materials in pressurized and boiling water nuclear reactors and are known to develop significant oxide nano-porosity. Corrosion results in the formation of an oxide layer containing nanopores while still being protective. It is still under debate whether the nanopores can provide fast diffusion paths for the oxidizing and hydriding species, and possibly accelerate corrosion. In the current study, we precisely quantify the nano-porosity in different regions of the oxide layer using a machine-learning-based method reported in a previous publication. In addition, the entire oxide layer depth and each pore within the layer are imaged, and a TEM-based 3D tomographic reconstruction is obtained. We further investigate the interconnectivity of the pores and the shortest path through the pores as functions of oxide depth, exposure time, and corrosion rate. Results demonstrate that interconnectivity is highest in the proximity of the W/O interface and gradually decreases towards the M/O interface. Specifically, higher temperatures, longer exposure times, and higher corrosion rates are correlated to increased interconnectivity among pores. This work provides essential evidence that pores in the near water/oxide interface region can provide paths for oxidizing species.

Deep generative modeling-based data augmentation with demonstration using the BFBT benchmark void fraction datasets

Deep learning (DL) has achieved remarkable successes in many disciplines such as computer vision and natural language processing due to the availability of “big data”. However, such success cannot be easily replicated in many nuclear engineering problems because of the limited amount of training data, especially when the data comes from high-cost experiments. To overcome such a data scarcity issue, this paper explores the applications of deep generative models (DGMs) that have been widely used for image data generation to scientific data augmentation. DGMs, such as generative adversarial networks (GANs), normalizing flows (NFs), variational autoencoders (VAEs), and conditional VAEs (CVAEs), can be trained to learn the underlying probabilistic distribution of the training dataset. Once trained, they can be used to generate synthetic data that are similar to the training data and significantly expand the dataset size. By employing DGMs to augment TRACE simulated data of the steady-state void fractions based on the NUPEC Boiling Water Reactor Full-size Fine-mesh Bundle Test (BFBT) benchmark, this study demonstrates that VAEs, CVAEs, and GANs have comparable generative performance with similar errors in the synthetic data, with CVAEs achieving the smallest errors. The findings shows that DGMs have a great potential to augment scientific data in nuclear engineering, which proves effective for expanding the training dataset and enabling other DL models to be trained more accurately.

High temperature electrochemical reaction parameters affecting electrochemical corrosion potential of stainless steels in hydrogen peroxide environment

ABSTRACT High-temperature electrochemical reaction parameters for stainless steels have been determined based on data measured in pure water at high temperatures to realize an analytical model that can predict electrochemical corrosion potential (ECP) of stainless steels in a hydrogen peroxide (H2O2) environment without using any adjustment parameters based on ECP data. Behaviors of measured cathodic polarization curves for H2O2 were explained well with the determined parameters. Calculated ECPs of stainless steels agreed with measured values in the cases that it was acceptable for the model to have an error range of ±0.1 V in a range from the deaerated condition to 1000 ppb (= μ g kg−1) H2O2 concentration. It is expected that the ECP model in the H2O2 environment with the determined electrochemical reaction parameters will have greater validity for, and better interpretability of, the ECPs of stainless steels determined in a boiling water reactor environment than previous models.

Linear or non-linear stability monitor in BWRs? Introducing a new non-linear monitor based on the fractal spectrum

The stability phenomenon in nuclear boiling water reactors (BWRs) continues to be of scientific and technological interest contributing to the development of advanced stability monitors. The Mexican scientific community over the course of more than two decades (in the world more than four decades) has developed innovative techniques for stability analysis in BWRs, both online and offline; inspired in the first instance by the Laguna Verde NPP instability event occurred on January 24, 1995. Under this perspective, an extensive discussion of linear and non-linear techniques for stability analysis of BWRs is included in this work as well a novel method of stability analysis in BWR based on a Higuchi Fractal Spectrum for the design of a non-linear monitor. This methodology, based on a generalization of the Higuchi Fractal Dimension (HFD), permits to determine in real-time, the stability of a BWR. Hence, real challenging cases in which the instability of the reactor is not evident, were used to test this proposed stability indicator. In general, this proposal gives a better forecast than that based on the simple HFD.

Minimal requirements to thermal-hydraulics/3D-neutron kinetics clustering for BWR transient analysis

Boiling water reactor (BWR) transients and accidents result in a spatial redistribution of the void content and thus of the power. In addition, for many transients, the reactivity feedback coefficients are varied over a wide range. Therefore, coupled thermal-hydraulics/3D neutron kinetics (TH/NK) methods are required for BWR transient analysis. The mapping or clustering of the system TH channels to the fuel assemblies of the NK module is one of the main modelling assumptions made for such coupled simulations. The simulation of BWR transients requires significant computing time if a detailed one-to-one fuel assembly to flow channel mapping is applied. This paper presents an attempt to find the minimal requirements for TH/3D-NK mapping for BWR transient analysis. In other words, we explore different methods of selecting optimal channel groupings and demonstrate how the number of channels affects the solution accuracy for a selected BWR fast transient. It was found that the optimal mapping strategy strongly depends on the parameters of interest, local, e.g. fuel assembly power, fuel cladding temperature, or integral, e.g. peak dome pressure, reactor peak power or downcomer water level. It was concluded that for many cases one-to-one clustering is an optimal technique for BWR coupled analysis.

Prediction of interfacial shear stress and pressure drop in vertical two-phase annular flow

The interfacial shear stress and pressure drop of an upward vertical annular flow of nitrogen–water, HFC134a–water, and nitrogen–95 % ethanol solution were comprehensively investigated considering the effect of the liquid–gas density ratio and surface tension. A direct link between the disturbance wave height and equivalent sand-grain roughness was noted through the analogy with the famous Moody chart for single-phase turbulent flows. A predictive model of the interfacial friction factor was developed based on this finding. To predict the pressure drop of the annular flow, a new model with good predictive performance for annular flows of various working fluids including steam–water under boiling water reactor operating condition (286 °C and 7 MPa) was proposed.

Effect of diameter and inclination of steam injection pipe on chugging dynamics in direct contact condensation

Low mass flux of steam and high degree of subcooling are the prominent factors for occurrence of unstable direct contact condensation, commonly known as chugging, when steam is injected into a subcooled water pool. Such an arrangement is typically used in the pressure suppression pool of the primary containment vessel in a boiling water reactor. With an attempt to identify the conditions of minimum pressure oscillations during the chugging regime, the present study investigates the effect of internal diameter (0.5–1 in.) and inclination (8.75˚) of steam injection pipe on steam bubble dynamics, pressure transients and heat transfer characteristics at different steam mass flow rates (5–25 kg/hr) and submergence depths of steam pipe (1–13 cm) within the water pool. The study reveals that the condensation cycle time decreases with a decrease in the internal diameter of the pipe. Furthermore, the frequencies of fluctuations in the interfacial area of the steam bubble and pressure transients in the steam pipe exhibit an increasing trend with an increase in steam mass flow rate and a decrease in internal diameter of the steam injection pipe. The chugging frequencies in the inclined and vertical orientation of steam injection pipe lie within the range of 0.5–3 Hz. The average heat transfer coefficient increases with an increase in the internal diameter of pipe and steam mass flow rate. The amplitude of pressure oscillations has decreased significantly whereas heat transfer coefficient has increased marginally upon changing the orientation of steam injection pipe from vertical to inclined. Therefore, smaller diameter and inclined orientation of steam injection pipe are favourable for mitigation of pressure oscillations in applications of direct contact condensation.

Hydrothermal corrosion of PVD and cold spray Cr-coatings on Zircaloy-4 in hydrogenated and oxygenated LWR coolant environments

Accident tolerant fuel (ATF) are currently being developed as part of an international effort to modify or replace zirconium-based cladding materials in light water reactors. The main objective of ATF cladding materials is to improve fuel reliability and safety during accident scenarios. Thus, these materials should perform comparable to or better than the current zirconium alloy. One of the ATF concepts explores modifying standard Zircaloy cladding by coating their outer surface. In this study, surface-modified Zr-4 is produced by depositing a protective coating of chromium through two different coating techniques, Physical Vapor Deposition (PVD) and Cold Spray (CS). The coated specimens were tested at ∼300 °C in out-of-pile autoclaves simulating Boiling Water Reactor (BWR) conditions. Results show that after one month of immersion, Cr-coated Zr-4 evidenced a small mass gain in hydrogenated environments. However, both types of coated specimens showed significant coating degradation and mass loss in oxygenated environments.

Condenser Retrofit in Leibstadt Nuclear Power Plant (BWR) - Far Beyond a Standard Modular Solution

Abstract The Leibstadt nuclear power plant (KKL) in Switzerland was originally commissioned in 1984 with a nominal power output of 1000 MWel. Since then, there have been several uprates of the boiling water reactor (BWR) which have increased the power to 1275 MWel. After the nearly four decades of operation, the original condenser was impacted by advanced erosion corrosion and by progressive performance deterioration. In addition, the thermal power uprates required load reduction during hot summer periods to avoid high condenser back pressure. Consequently, KKL asked GE to carry out a comprehensive retrofit study to identify potential improvements on both lifetime and performance. These included: robust material selection, optimized extraction of non-condensable gases by means of a bespoke air cooler section of the condenser and applying GE Steam Power's latest innovative tube pattern design which is aligned to BWR requirements. However, site constraints prevented a typical modular condenser retrofit. Hence, working in a close cooperation between KKL and GE Steam Power, a customized retrofit concept was developed which used semi-prefabricated skeletal modules and an on-site rebuild of the condenser. In total, approximately 475 tons of old condenser material were dismantled, 24 new semi-prefabricated skeletal modules were installed, a condenser with 51'600 tubes was rebuilt on-site and 8 waterboxes were replaced with an improved design made of stainless steel. More than 350 professionals contributed to the successful retrofit with an overall site effort of 270'000 working hours - increasing the power output to 1285 MWel.

Elemental distribution in a decommissioned high Ni and Mn reactor pressure vessel weld metal from a boiling water reactor

Elemental distribution in a decommissioned high Ni and Mn reactor pressure vessel weld metal from a boiling water reactor

Efficiency analysis of nuclear power plants: A comprehensive review

Nuclear power plants play a significant role in global electricity generation, offering a reliable and low-carbon energy source. Maximizing the efficiency of nuclear power plants is crucial for optimizing energy output and reducing operational costs while ensuring safety and environmental sustainability. This paper presents a comprehensive review of the efficiency analysis conducted on nuclear power plants, covering various aspects such as thermal efficiency, conversion efficiency, fuel utilization, and overall plant performance. The review synthesizes the existing literature, discusses key factors influencing efficiency, and highlights potential areas for future research and improvements. It is revealed from the present review that the efficiency of the Boiling Water Reactor (BWR) of the Nuclear Power Plant is 32% and which can be increased to 33% by using advanced BWR. Also, the efficiency of the Pressurized Water Reactor (PWR) of the Nuclear Power Plant is 33% and which can be increased to 36.5% by using the PWR with Gas Burner and Reheating. On the other hand, the efficiency of the Small Modular Reactor (SMR) of the Nuclear Power Plant is 33.4% and which can be increased to 35.5%, 37.4% and 45% by using SMR with Gas Burner, SMR with Reheating and SMR with CCGT, respectively. Therefore, it is predicted that the Small Modular Reactor (SMR) is the most economic Nuclear Power Plant as it has higher thermal efficiency when it is incorporated with Combined Cycle Gas Turbine (CCGT).

Ultrasonic Velocity Profiler for the Measurement of a Bubbly Flow Velocity Vector in Small Channels

The multi-dimensional velocity distribution of coolant in bubbly flow within the fuel rod bundles of the reactor core in boiling water reactors (BWRs) is elucidated by experimental investigationin this study. Since a measurement technique is required for such an investigation, this paper proposes the development of an ultrasonic velocity profiler (UVP). The combination of special ultrasonic transducers and modified signal processing on the UVP is proposed to obtain a multi-dimensional velocity vector of the bubbles and liquid in bubbly flow. The ability of the proposed technique is demonstrated by performing an experiment in swirling bubbly flowand itsapplicability confirmed by comparing the resultswith another technique. The sound pressure distribution in the narrow channel of the rod bundle isthenmeasuredprior to the verification ofthe ultrasonic wave emitted through a small channel. The echo signal reflected from reflectors dispersed in the liquid,bubble, and tracer particles in the small channel of the rod bundle indicates that the proposed UVP can be applied inthis application with a low level of multi-reflection. Finally, the UVP system is demonstrated to measure the velocity vector of bubbly flow in the narrow flow channel on the rod bundle, andthe velocity vector of the bubble and liquid obtained simultaneously.

Modelling of Tarapur BWR lattice and core to estimate the required number of SB-BE startup neutron sources

During the refuelling and criticality of Boiling Water Reactors (BWR), fission counters are used as Source Range neutron Monitors (SRMs) to get the count rates. The study shows that the neutron counts due to spontaneous fission sources are insufficient to meet the SRM's limitations of 3 counts per second (cps). In this work, TAPS fuel assembly is modelled using the lattice transport code DRAGON5, and lattice parameters are generated as a function of fuel depletion. These parameters have been further used to model the reload-2 of the BWR core. 3D diffusion theory code KINFIN is used to simulate the BWR core with external neutron sources to estimate the neutron flux at three SRM locations. Thermal fluxes are then used to compute the neutron count rates using the sensitivity of SRMs for different combinations of five Sb-Be neutron sources.

CrN–NbN nanolayered coatings for enhanced accident tolerant fuels in BWR

The accident tolerant fuel (ATF) concept has emerged in the years after the 2011 Fukushima accident as part of a renewed effort in research for light water reactors. The primary focus is to further improve safety measures under and beyond design basis accident conditions, and to improve fuel cladding performance in normal operation. The application of a coating on zirconium claddings can achieve both these aims without extensive changes to the reactor design. Metallic chromium coatings have been profusely studied as solution for pressurized water reactors, but the search for an effective ATF coating able to withstand the environment inside boiling water reactors (BWRs) is still ongoing. In this work, two different versions of a novel nitride coating composition were studied. Zirconium claddings coated with 8 µm thick layers of superlattice CrN–NbN and a nanolayered CrN–NbN were tested in autoclave under BWR operating conditions for 60 days. Scanning electron microscopy, transmission electron microscopy, energy dispersive x-ray spectroscopy, electron back-scattered diffraction, x-ray diffraction, and atom probe tomography were employed to characterize as-deposited and autoclaved samples of these two materials. During exposure, both coating versions formed a stable, dense and passivating oxide scale (200–300 nm thick) on the surface, demonstrating improved oxidation protection under operating conditions. Some differences in the oxide growth mechanism were observed between the superlattice and the nanolayered CrN–NbN coatings, which allowed to glimpse at the effect of the layer thickness on the oxidation protection provided by these coatings. The nano-structured morphology of both coatings remained unaffected by the autoclave test, but a 35 nm thick Zr-Cr-N phase was found at the coating-substrate interface of the superlattice CrN–NbN coated cladding.

Construction Materials of Active Zones of New Generation Nuclear Reactors

The materials of the article consider the analysis of construction materials of active zones of new generation nuclear reactors. The analysis reflects general ideas about the development of reactor technologies: in the 1950s and 1960s, the first generation of reactors was created; in the early 1970s, the operation of industrial reactors began - reactors of the second generation: pressurized water reactors (WWER, PWR), boiling water reactors (RBMK, BWR), heavy water reactors (CANDU), as well as gas-cooled reactors (AGR). Further development of some types of reactors made it possible to create reactors of the third generation in the 1980s. Priority when choosing directions of development in the category of revolutionary projects should have proposals capable of bringing a new quality to solving the problems of the nuclear energy industry of the future. Promising reactors have advantages in economy, safety, reliability and non-proliferation of nuclear materials. The effectiveness and reliability of structural materials are determined by the totality of changes in the characteristics of the materials as a result of the entire complex of phenomena occurring in them in the field of irradiation, in connection with the changing parameters and operating conditions. The use of high-purity metals as initial components of new structural materials and the development or optimization of their smelting technologies should ensure the required level of characteristics and properties of products made from them. The implementation of these concepts should be ensured by the development of new structural materials: ferritic-martensitic and austenitic steels, nickel and other new alloys.