Reactor Power Plant Research Articles

Study on the potential development of nuscale small modular reactor (smr) nuclear power plant to support the energy needs in the capital city of nusantara

Abstract In facing the dependence on unsustainable energy sources such as fossil fuels, diversifying the energy supply becomes crucial. As Indonesia’s new capital city, Nusantara significantly requires new and renewable energy sources as it can enhance energy security by reducing dependence on energy imports while also providing economic stimulus and creating new job opportunities. The potential development of new and renewable energy sources.in Nusantara is a solution to Indonesia’s dependence on fossil energy, which negatively impacts the environment and public health. One of these energy sources is the NuScale SMR type of nuclear power plant. The research method employed is a literature study based on various relevant sources related to the research topic. The results of the literature study indicate that the NuScale SMR is a nuclear reactor that possesses advantages in terms of size, high safety, reliability, resource efficiency, support for renewable energy, as well as affordable modularity and scalability, making it a prime choice to fulfil the energy needs in Nusantara. By utilising new and renewable energy sources from smaller and modular nuclear technology, Indonesia can reduce its dependence on fossil energy and promote a transition towards cleaner and more sustainable energy sources.

Machine Learning-Assisted Prediction of Stress Corrosion Crack Growth Rate in Stainless Steel

Stainless-steel is extensively utilized in the key structural components of the main equipment in the nuclear island of pressurized water reactor nuclear power plants. The operational experience of nuclear power plants demonstrates that stress corrosion is one of the significant factors influencing the long-term safe operation of stainless steel in the high-temperature water of pressurized water reactor nuclear power plants. This study is based on the stress corrosion crack growth rate data of 316SS and 304SS stainless steel in the simulated primary water environment of pressurized water reactor nuclear power plants. Data mining and modeling were conducted using multiple machine learning algorithms, including Random Forest (RF), eXtreme Gradient Boosting (XGBoost), Support Vector Regression (SVR), and Gaussian Process Regression (GPR), and the Sharpley Additive explanation (SHAP) method was employed to analyze the interpretability of the model. The results indicate that the stress corrosion crack growth rate prediction model based on XGBoost outperforms other models in all assessment indicators. Compared with empirical equations, XGBoost exhibits high flexibility and excellent data-driven learning capabilities. In the test set, 90% of the prediction errors are within the range of experimental values, with the maximum error multiple being 2.5, which significantly improves the prediction accuracy. Moreover, the distribution of SHAP values is consistent with the theoretical study of the stress corrosion behavior of stainless-steel, effectively reflecting the impact of cold working, temperature, and stress intensity factor on the stress corrosion crack growth rate, thereby proving the reliability of the model’s prediction results. The achievements of this study hold significant reference value and application prospects for the prediction of the stress corrosion behavior of stainless-steel in a high-temperature and high-pressure water environment of pressurized water reactor nuclear power plants.

Analytic network process (ANP) based decision support tool for nuclear power plant location and reactor type selection

The nuclear energy industry has seen a resurgence due to interest in sustainable energy. Regulatory agencies are evaluating innovative technology and facility designs for the first time in a long time. The nuclear energy business needs a workable instrument for decision-making in light of the shift in public and regulatory opinion. This study has two main motivations. The first is to determine the set of criteria to be used in determining suitable locations for a nuclear power plant (NPP) to be established in Turkey and to determine suitable regions with Analytic Network Process (ANP) based analyzes in the light of these criteria. The second is to select the most suitable nuclear reactor type for the specified location. Plant efficiency and lifetime vary depending on the reactor type used. For this reason, in the study, both the location of the nuclear power plant and the selection of the appropriate reactor type for the plant were evaluated together. In this context, a criterion set consisting of 3 main criteria and 20 sub-criteria to be used in NPP installation location and reactor selection was determined. In the study, site selection was made by focusing on economic, social and technical factors, and reactor selection was made by focusing on technical factors. Giresun, Trabzon, Rize and Artvin provinces located in the Eastern Black Sea region were evaluated for NPP installation. For reactor selection, we focused on reactor types according to their coolants and in this context, PWR, PHWR, GCR and BWR reactors were examined. NPP location and reactor selection in the study; It was realized as Giresun (PWR) > Rize (PWR) > Trabzon (BWR) > Artvin (BWR).

Impact and tangential composite fretting wear of Zr-4 alloy tubes under random loading conditions

The fuel rod casing tube is a crucial component in pressurized water reactor (PWR) nuclear power plants. Flow-induced vibrations in the circulating water can result in complex alternating fretting wear between the casing tube and the positioning lattice frame. Existing research on fretting wear has primarily focused on unidirectional wear, with limited investigation into composite fretting wear under complex working conditions. This study conducted impact and tangential composite fretting wear tests on Zr-4 alloy tubes under varying constant and random impact loads to examine fretting wear behavior. The findings show that the peak friction coefficients in the impact and tangential composite fretting wear tests were consistent across the three constant load conditions, with higher peak friction coefficients observed under random load conditions and a longer time required to reach these peaks during micromotion tests. Comparative analysis revealed that random impact and tangential composite fretting wear caused the most severe damage. Under constant load conditions, wear damage became more severe with increasing load, transitioning from oxidized and adhesive wear to a peeling layer as the load intensified in the impact and tangential composite fretting wear tests.

Blistering Behavior of Beryllium and Beryllium Alloy under High-Dose Helium Ion Irradiation.

Beryllium (Be) has been selected as the solid neutron multiplier material for a tritium breeding blanket module in ITER, which is also the primary option of the Chinese TBM program. But the irradiation swelling of beryllium is severe under high temperature, high irradiation damage and high doses of transmutation-induced helium. Advanced neutron multipliers with high stability at high temperature are desired for the demonstration power plant (DEMO) reactors and the China Fusion Engineering Test Reactor (CFETR). Beryllium alloys mainly composed of Be12M (M is W or Ti) phase were fabricated by HIP, which has a high melting point and high beryllium content. Beryllium and beryllide (Be12Ti and Be12W) samples were irradiated by helium ion with 30 keV and 1 × 1018 cm-2 at RT. The microstructures of Be, Be12Ti and Be12W samples were analyzed by SEM and TEM before and after ion irradiation. The average size of the first blistering on the surface of Be-W alloy is about 0.8 μm, and that of secondary blistering is about 79 nm. The surface blistering rates of the beryllium and beryllide samples were also compared. These results may provide a preliminary experimental basis for evaluating the irradiation swelling resistance of beryllium alloy.

Adsorption mechanism of octadecylamine on inner surface of steam generator feed water pipe of pressurized water reactor nuclear power plant

Adsorption mechanism of octadecylamine on inner surface of steam generator feed water pipe of pressurized water reactor nuclear power plant

Development of advanced power reactor nuclear power plants containment pressure and temperature analysis methodology using CAP computer code

Development of advanced power reactor nuclear power plants containment pressure and temperature analysis methodology using CAP computer code

Coupling of the best-estimate system code and containment analysis code and its application to TMLB’ accident

With the development of advanced pressurized water reactor technology, the thermal-hydraulic coupling effect between the containment and the primary system becomes increasingly tight. In order to meet the demand for integrated safety analysis between the containment and the primary system, this paper investigates a direct coupling method between the best-estimate system code Advanced Reactor Safety Analysis Code and the containment analysis program ATHROC (Analysis of Thermal Hydraulic Response Of Containment). The feasibility of this direct coupling method and the applicability of the coupled program for overall safety analysis are demonstrated using Marviken two-phase flow release experiments. The ATHROC/ARSAC coupled program is employed to analyze the impact of the pressure relief function of the CPR1000 nuclear power plant pressurizer on the behavior of the primary system and containment during the TMLB’ accident. The calculation results indicate that these measures can reduce the pressure of the primary system to the level acceptable by the low-pressure injection system, but at the same time, they cause the pressure in the containment to rise to nearly 0.4 MPa. Therefore, to ensure the structural integrity of the containment, it is necessary for the non-passive hydrogen recombiner to effectively reduce the hydrogen concentration, thereby avoiding additional pressure increase in the containment due to hydrogen deflagration, which could lead to overpressure failure. The findings of this study are of significant reference value for improving the safety performance of thermal-hydraulic systems in operational Gen-II and advanced Gen-III pressurized water reactor nuclear power plants.

Optimal design of circulating water system in pressurized water reactor nuclear power plant

The optimization of the circulating water system of pressurized water reactor nuclear power plants can effectively improve the profit without affecting the primary circuit operation, which is of great significance in enhancing the market competitiveness of nuclear power plants. The circulating water system of the nuclear power plant is modeled, and the circulating water system optimization scheme is formed after the nuclear power unit is reformed from three aspects: condenser heat transfer area, condenser connection mode, and circulating water pump operation mode. The circulating water system optimization calculation is carried out for each scheme by using the minimum annual cost method. Based on the sensitivity analysis of the factors that affect the profits of nuclear power plants, the best circulating water system optimization scheme is finally given for new nuclear power plants and nuclear power plants in operation, which provides a reference for the construction of related nuclear power projects.

LIBS as diagnostics of analytical chemistry for surface mapping of complex mixed samples simulating debris inside the TEPCO's Fukushima Daiichi Nuclear Power Plant (FDNPS) reactor cores

The Nuclear Energy Agency (NEA) launched the Nuclear Education, Skills and Technology (NEST) Framework to pursue careers in the nuclear field, by exposing researcher working on these topics to international challenging project of real-world issue and by transferring the knowledge and expertise accumulated in the current generation to them through hands-on training.In this framework the 2022 edition of the NEST project offered a training educational period at the Collaborative Laboratories for Advanced Decommissioning (CLADS) located at Tomioka, Futaba District, in the Fukushima prefecture (Japan).Among the research sectors active at CLADS there is the application and development of the LIBS technique as diagnostics of analytical chemistry aiming at characterizing the debris inside the Tokyo Electric Power Company (TEPCO)’s FDNPS reactor cores after the tsunami of March 2011, which destroyed three of the six reactors of the plant. These debris need to be chemically characterized with techniques suitable to be implemented in compact, remote and radiation resistant devices, due to the residual radioactivity of the cores. The present study shows the results in realizing chemical bi-dimensional maps of samples in the form of compressed tablets of mixed oxides, with a complex distribution and concentration of chemicals simulating these debris. The results allowed to create detailed maps of the samples, with spatial resolution down to 0.5 mm and an excellent correspondence between the real spatial distribution of the materials and that reconstructed by LIBS.Moreover, it was found a good correspondence between the nominal concentration of the chemicals and the concentration estimated by using LIBS.This study shows the potentialities of LIBS in the realization of chemical maps on samples of interest for FDNPS, providing multi-elemental images of the samples under analysis.

Controversy or Consensus: Launching Small Modular Nuclear Reactors in Romania

Abstract The transition towards net zero by 2050 is underway in the European Union, but its country members are divided over their energy strategies. France, Poland, Hungary, Bulgaria, Slovakia, and Romania support nuclear plants as contributing to carbon neutrality, while Germany, Spain, Denmark, Portugal, and Luxembourg oppose (Reuters, 2023). The study “Between Stability and Sustainability: Nuclear Energy in Romania's Energy Mix” concludes that Romania can benefit from its nuclear facilities and domestic uranium resources, to consolidate its sustainable energy policy. However, new nuclear capacities may pose considerable challenges, especially in terms of costs (Dumitrescu, 2023). Unfortunately, rising costs per megawatt hour are the very reason why NuScale, the operator chosen by Romania to deploy six small nuclear reactors by 2027, canceled their US-based flagship Carbon Free Power Project (Fortune, 2023). Nevertheless, the setback in the home country did not prevent Romanian authorities from licensing the NuScale small modular reactor power plant in Doicești (World Nuclear News, 2023). The watchdog Greenpeace warned about the project’s high costs and the long construction timeframe. It was also emphasized that Romania adopted a technology that had never been tested before. In addition, the location of the new plant had been decided without prior consultation from the local community, claimed the NGO (Greenpeace, 2022). According to the journalists, the local community has been divided between those who want new jobs and people afraid of radiations. Reporters made a parallel between the current situation and a formerly failed Chevron shale gas production investment, due to the strong opposition to fracking (Hotnews, 2024). This paper attempts to determine how a newcomer to the nuclear energy market, with a significant approval setback in its home country, using untested technology, can complete the construction of six small modular reactors in Romania, integrate those into the national grid and win over the public acceptance by 2030.

The Effects of Cl− and Selected Deoxidizers on the High-Temperature Corrosion Electrochemistry of Alloy 690 in Nuclear Steam Generator Water

The heat transfer tube in a steam generator serves as a critical heat exchange component in the primary and secondary loops of pressurized water reactor (PWR) nuclear power plants. The corrosion resistance of the heat transfer tube material directly influences the longevity of PWR nuclear power plants. This study investigated the electrochemical corrosion properties of 690 alloy (UNS N06690) in a simulated secondary water environment of PWR, focusing on different chloride ion concentrations and combinations of deoxidizers. The findings reveal a gradual decrease in the corrosion potential of 690 alloy, accompanied by an increase in self-corrosion current and a progressive reduction in the passivation range, ultimately leading to its disappearance as chloride ion concentration rises from 0 µg/L to 500 µg/L. Moreover, the impedance value of the inner film exhibits a declining trend with an increase in chloride ion concentration. Conversely, the resistance value of the outer film remains relatively stable while the size and spacing of oxide particles on the surface of the 690 alloy continuously increase. This observation suggests that chloride ions primarily influence the formation of the inner passivation film, which in turn determines the corrosion resistance of the 690 alloy. Notably, the performance of the 690 alloy is similar when the deoxidizer combination is ammonia (NH3) + erythorbic acid (ERA) or NH3 + hydrazine (N2H4), demonstrating the ability to form a relatively complete passivation film and exhibit improved corrosion resistance compared to NH3+N-isopropyl hydroxylamine, additionally, when the deoxidizer combination is NH3+N2H4, the 690 alloy exhibits lower self-corrosion current density across different chloride ion concentrations, indicating enhanced corrosion resistance.

Coolant Hydrodynamics at the Inlet to the FA of the RITM-Type Reactor of a Small Nuclear Power Plant

Coolant Hydrodynamics at the Inlet to the FA of the RITM-Type Reactor of a Small Nuclear Power Plant

Experimental study on the retention of aerosol particles through concrete cracks under high Reynolds number flow

In the event of severe accidents in pressurized water reactor (PWR) nuclear power plants, the potential leakage of radioactive aerosols through containment cracks poses a considerable radioactive hazard to the public. Understanding aerosol transport and retention in cracks helps reduce the conservatism and uncertainty of radioactive hazard assessment. Concrete cracks are recognized as a pivotal pathway for the leakage of radioactive aerosols, and several researchers have undertaken experimental investigations concerning the aerosol transport and retention in concrete cracks. However, the majority of these studies have rather low gas flow Reynolds numbers. In this work, an experimental setup is built to study aerosol transport and retention in concrete cracks under high Reynolds number flow. The TiO2 aerosol with a mass median diameter of 1 μm and two concrete crack specimens are used in experiments. The results of gas flow experiments indicate that the Reynolds number is capable of reaching 10547. Combining the flow experimental data and Suzuki's formula, the equivalent heights of these two crack specimens are approximated as 303.67 μm and 231.48 μm. The experimental results indicate a notably high retention rate of aerosols, exceeding 0.8. Furthermore, under high Reynolds number flow, the retention rate varies over a relatively narrow range, with the larger the equivalent height of the crack resulting in a lower retention rate. The experimental results match well with the mechanistic analysis based on inertial deposition theory, demonstrating the rationality of the inertial deposition theory.

Validation of EPG/SAG Rev. 4 of the Kuosheng Nuclear Power Plant Using the MAAP5 Code

The Boiling Water Reactor Owner’s Group released Emergency Procedure Guidelines and Severe Accident Guidelines Revision 4 (EPG/SAG Rev. 4) in 2018. The major improvement to EPG/SAG Rev. 4 was Contingency 1 (Alternate Level/ Pressure Control). Contingency 1 coordinates the reactor pressure vessel (RPV) water level and RPV pressure control action to prolong the availability of steam-driven injections and optimize the transfer to motor-driven systems. In this study, the effectiveness of the EPG/SAG Rev. 4 Contingency 1 strategy was compared with those of EPG/SAG Revision 2 Contingency 1 and Specific Major Incident Guidelines (SMI) using the Modular Accident Analysis Program, Version 5 (MAAP5). SMI was developed by the Taiwan Power Company to mitigate a Fukushima-like accident. The surrogate plant that analyzed is the Kuosheng Nuclear Power Plant (NPP). The Kuosheng NPP is BWR-6 Mark-III containment. MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic risk assessments and for verification and validation of mitigation actions specified in severe accident management guidelines. The simulation scenarios were extended loss of alternative-current power and loss of ultimate heat sink. The low-capacity, motor-driven portable pump was the only available system for RPV injection in the first hour of the accident. In this time period, the RPV water level and pressure were controlled by reactor core isolation cooling and safety relief valves. After this study, the strategy of EPG/SAG Rev. 4 Contingency 1 was successfully validated, and the effectiveness of minimum pre-depressurization RPV water level and the low-capacity, motor-driven portable pump were also demonstrated in this study.

An integrated assessment method of real-time source term for high temperature gas-cooled reactor

To address high temperature gas-cooled reactor (HTR) nuclear power plant (NPP) safety analysis and emergency response, this paper proposes an integrated source term assessment (ISTA) method, guided by the principles of reality, rapidity and rationality. The fundamental concept of ISTA is to establish a connection between the theoretical model of source term (ST) calculation and actual conditions of HTR NPPs. According to the ISTA method, STs can be evaluated by mapping key parameters through monitoring variables during any abnormal operational conditions or accident conditions, resulting in a more realistic representation. Furthermore, the output of STs in various typical spaces within an HTR NPP further expands the application scope beyond the development of nuclear emergency protection strategies. In addition, a corresponding tool (ISTAT) for engineering application is developed and the case is given to verify the feasibility. This research highlights the technological advancement affecting radiological release scenarios and avoiding unnecessary conservatism, which also can give reference for other advanced reactor NPPs.

A Multisource Uncertainty Fusion Reliability Evaluation Method for the Control Rod Drive Mechanism of Nuclear Power Plants

The reliability of a pressurized water reactor power plant’s control rod drive mechanism (CRDM) is affected by many factors, such as operation states, unit performance, and dynamic environments. Multiple sources of uncertainties, including random, interval, and fuzzy, exist when analyzing the reliability of CRDMs. Modeling reliability without considering the fusion of multisource uncertainties may result in model distortion and may not provide realistic assessment results. This paper proposes a multisource uncertainty fusion method powered by the margin-based variable transformation and the Bayesian network propagation. The interval and fuzzy variables are transformed into random variables to obtain the corresponding generalized probability density function of the CRDM’s feature parameters. After that, the CRDM’s reliability can be evaluated via probability quantization of the Bayesian network inference result through sampling algorithms. A magnetic-jack type CRDM case is presented to verify the proposed method, and the results show that this method can fuse three types of uncertainty variable in a unified way and effectively obtain reliability evaluation results.

Insights into thermomechanical fatigue behavior of nitrogen alloyed 316LN stainless steel under various temperature cycling ranges

Thermomechanical fatigue (TMF) failure has become one of the most important life-limiting factors of components exposure to the high temperature in the fourth-generation sodium-cooled fast reactor (SFR) nuclear power plant. Therefore, TMF tests at different temperature cycling ranges are performed to study the cyclic deformation behavior and damage mechanism of a nitrogen alloyed 316LN stainless steel. Moreover, the isothermal fatigue (IF) tests at the maximum temperature of TMF cycling are conducted for comparison. The cyclic hardening and softening response are analyzed, taking into account the phenomenon of dynamic strain aging which occurs at the temperature range of 350 °C–650 °C. Furthermore, the critical strain corresponding to the occurrence of serrated stress flow in engineering stress-strain curves keeps decreasing from 28.94% at 350 °C to 0.09% at 650 °C. The distribution of plastic deformation and the degree of strain localization are evaluated by the characterization results of kernel average misorientation map, high/low angle grain boundary map and dislocation configuration. The dominant damage mechanism and its corresponding cracking behavior in both IF and TMF tests are analyzed based on the observation of fracture surface morphology. In addition, the sequence order of fatigue life is determined as IF < IP-TMF < OP-TMF. Moreover, the maximum difference of fatigue life between IF test and TMF test occurs at the temperature range of 400 °C–600 °C, in which the IP and OP life is about 2.91 and 3.65 times larger than the IF life. The results in this work provide reliable data support and theoretical guidance for the fatigue design of high-temperature components in SFR.

Proposal and Validation of a Diagnosis Method of Fuel Failures in Pressurized Water Reactor Nuclear Power Plants

Based on research into the diagnosis methods for fuel failures in pressurized water reactor nuclear power plants and an analysis of existing problems, this paper proposes an enhanced fuel failure diagnosis method, addressing three key improvements: identification of the occurrence of fuel failure, diagnosis of the number of failed fuel rods, and the size of fuel rod failure. Other factors that may influence the diagnosis results are also identified and discussed. The proposed method is then verified using operational data from in‐service nuclear power plants in China. The verification results demonstrate that the improved diagnosis method accurately identifies fuel failures and exhibits broader applicability.

Probing Evolution of the Flux-Pinning Landscape in REBCO Coated Conductors Caused by Gamma Irradiation Using DC and AC Magnetometry: A Novel Approach to Tokamak Magnet Material Development

Optimisation of REBCO coated conductor tapes specifically for use in nuclear fusion will help improve the magnet component lifetimes in future tokamak reactor power plants. The focus of this work was exploration of a novel approach to irradiation studies on REBCO tapes, utilising multiple magnetic measurements to probe evolution of the REBCO flux-pinning landscape more deeply than reported in other studies, for the purpose of identifying primary limiting factors affecting performance. Gamma irradiation experiments were conducted, and pre-/post-irradiation results from DC and AC magnetic measurements using a Physical Property Measurement System (PPMS) are discussed. Magnetisation critical current density (Jc) decreased in all samples with increasing dose, except for the silver overlayer-only samples which did not contain artificial pinning centres (APCs), where Jc increased with dose. Removal of the copper stabiliser coupled with the presence of APCs allowed gamma irradiation to induce pinning force maximum peak shifts, from above 14 T before irradiation to below 9 T afterwards. Flux creep rate varied with the evolving pinning landscape, and the degree of Jc degradation directly correlated with creep rate fluctuations post-irradiation. Changes in critical temperature and diamagnetic saturation also corresponded with changes in Jc and flux creep rate. The major conclusion from this study was that minimisation of flux creep rate is the key to maintenance of performance under fusion-relevant operating conditions. Flux creep manifests as problematic AC losses in all high-temperature superconducting machines; therefore, future work will focus on reduction/prevention of the phenomenon to enhance longevity of performance in any application.