bluuebunny/crossref_metadata_2025_split huggingface dataset:
Design and Analysis of a Nuclear Hydrogen Production System by Methanol Reforming Using Light Water Reactor (LWR)
Authors: Liu Ruiyang , Peng Zhiyi , Yin Huaqiang , Zhang Huang 
Published: November 2024

Abstract
Hydrogen is a multiple energy carrier, which can help tackle various critical energy challenges and transition towards green and zero-carbon industries. There are numerous hydrogen production methods, including methane steam reforming, electrolysis of water, and the iodine-sulfur thermochemical cycle. These methods typically require a significant amount of electricity and heat, aligning well with the benefits of nuclear power. Nuclear energy is a reliable, low-carbon power source that generates high energy output. Nuclear reactors can produce a continuous supply of heat and electricity. It is necessary for the production of clean hydrogen, which reduces both infrastructure and hydrogen costs. Light-water nuclear reactors (LWRs) currently comprise over 80 percent of all commercial nuclear reactors worldwide. Improving their overall utilization capacity is crucial for enhancing the economics of commercial nuclear reactors. This study proposed utilizing a methanol steam reforming process technology for hydrogen production, which can be applied to LWR. The technology utilizes steam that produced from the light-water nuclear reactor to produce hydrogen. The steam is also a heat source for the methanol steam reforming hydrogen production reaction. It combines the steam with either pressure swing adsorption (PSA) or membrane separation purification technology to purify the hydrogen in the product gas, resulting in a reliable and high purity supply of hydrogen that could be 99.99%. This system for producing hydrogen using nuclear energy comprises a LWR combined with methanol hydrogen reforming. The study investigated outlet parameters of LWR and determined optimal reaction conditions for producing hydrogen through methanol reforming using nuclear energy. The process operates at temperatures between 200°C and 300°C and at pressures ranging from 1.0 MPa to 1.8MPa. The findings demonstrate the feasibility of hydrogen production from LWR. Additionally, the purification unit separates and purifies the hydrogen product gas to enhance hydrogen purity and minimize impurity effects on hydrogen equipment, thereby supporting the comprehensive use of nuclear energy and expanding the availability of hydrogen. This study compares the economic viability of two systems: a light water reactor combined with methanol reforming for hydrogen production and one that uses electrolysis of water for hydrogen production. Our findings indicate that methanol steam reforming using the light water reactor system is more cost-effective than water electrolysis using the same system. This study aims to design a hydrogen production system that is appropriate for use in a light water nuclear reactor and contributes to the effective utilization of nuclear energy.
A Single Channel Thermal-Hydraulic Calculation Module for PWR Pin-by-Pin Wise Coupled Calculation System
Authors: Li Zhigang , Pan Juejie , Xia Bangyang , Lu Wei , Zhao Wenbo , Qiang Shenglong 
Published: November 2024

Abstract
Due to the strong feedback effect between neutronics and thermal-hydraulics in the core of pressurized water reactors (PWR), neutronics and thermal-hydraulics coupling calculations are often used in the design and safety evaluation of PWR to provide more accurate results. The coupling calculation of neutronics and thermal-hydraulics at the full core pin-by-pin wise can provide more precise and precise coupling parameters, which is a hot research topic in the industry. How to advance the efficiency and accuracy of neutronics and thermal-hydraulics coupling calculations at the entire core pin by pin wise is a challenge in this research. Multiple digital reactor plans have been proposed worldwide, and one important task is to attempt to use the most advanced numerical simulation methods, high-precision calculation models, and ultra large-scale high-performance computing clusters to develop the new generation of digital reactor technology. In order to improve the efficiency of the pin-by-pin wise neutronics and thermal-hydraulics coupling calculation of the entire reactor core, this paper developed a two-phase thermal-hydraulic calculation module TH1D based on a single channel model, which can achieve thermal conductivity calculation of flat fuel and rod fuel, as well as two-phase thermal-hydraulic calculation of a single channel, and improve parameters such as fuel effective temperature, coolant temperature, moderator density, and void fraction for thermal feedback. And coupled it with the core fine mesh transportation calculation program CORCA-SPn to develop a PWR pin-by-pin wise neutronics and thermal-hydraulics coupling calculation program. The overall performance of TH1D was evaluated by conducting pin-by-pin wish neutronics and thermal-hydraulics coupling calculation on two typical PWR core: Hua-long Pressurized Reactor (HPR1000) and the Coupling benchmark of PWR with High Parameters (COPHP). Compared with the coupling program using the traditional sub-channel/multi-channel thermal-hydraulic calculation program CORTH/CORTH-R, the differences in thermal-hydraulic parameters between calculated by TH1D and the results of sub-channel/multi-channel program, as well as the differences in coupling calculation results, were compared. The following conclusions were drawn: 1) Through thermal-hydraulic simulation of a single channel containing a single plate or rod, TH1D showed good load performance compared to the reference results, with a relative deviation of less than 0.5% for all parameters; 2) the key result parameter deviation of the pin-by-pin wise coupling calculation using CORCA-SPn/TH1D is within an acceptable range, with the maximum keff deviation ≤ 250pcm and the maximum power weight deviation ≤ 1.88%; 2) The average efficiency of pin-by-pin wish coupling calculation is increased by 16 times (compared to CORTH) and 0.86 times (compared to CORTH-R), respectively.
Two-Phase Flow Distribution in 5 × 5 Rod Bundle During Boil-Off Process
Authors: Ueda Shota , Arai Takahiro , Ui Atsushi , Furuya Masahiro , Okawa Riichiro , Shirakawa Kenetsu , Yodo Tadakatsu 
Published: November 2024

Abstract
Boil-off may occur in light-water reactors with a stagnant core flow when a long period passes after the initiation of an accident. Improving the predictability of the two-phase mixture water level and void behavior during the boil-off process can effectively evaluate the long-term core coolability in accident progression analyses. Thus, this study investigated the void behavior transition during the boil-off process in a 5 × 5 heated rod bundle, which partially simulates the actual fuel assembly, at atmospheric pressure. Heat fluxes of 3, 6, and 9 kW/m2 were imposed to simulate the decay heat in long-term core cooling during an accident. The influence of collapsed water level on the distributions of void fraction, bubble chord length, and phase velocity was quantified using subchannel void sensors mounted within the rod bundle. The void fraction increased as the collapsed water level decreased because of the decrease in hydrostatic head. The measured three-dimensional distribution of phase velocity indicates the occurrence of inverse flow and the presence of global flow in the bundle during the boil-off process. The obtained results can contribute to the understanding of the multi-dimensional behavior of boil-off phenomena.
Effect of Pressure Acting on the Flaw Surface on Plastic Collapse Strength for Pipes With the Circumferential Surface Flaw
Authors: Machida Hideo 
Published: November 2024

Abstract
The Rules on Fitness-for-Service of Japan Society of Mechanical Engineers (the rules on FFS) stipulates the limit load evaluation method based on the plastic collapse strength as one of the fracture evaluation methods for pipes with circumferential flaws. However, since this evaluation method treats flaws as a decrease in thickness regardless of their location, the internal pressure acting on the flaw surface is not taken into consideration even when the flaws are on the inner surface of the pipe. The limit load evaluation methods that take into account the effect of the internal pressure acting on the flaw surface are derived, and compared the plastic collapse loads with and without considering the internal pressure acting on the flaw surface. When the flaw is shallow and its angle is small, the both plastic collapse loads are almost the same. However, when the flaw is deep and its angle is large, the plastic collapse load decreases due to the internal pressure acting on the flaw surface. Especially when the internal pressure is high, the impact is large. In the rules on FFS, the safety factor is taken into consideration in the evaluation of allowable stress, and the flaw depth is limited by a regulatory requirement when the flaw angle exceeds 60 degrees. Due to these, the impact of internal pressure acting on the flaw surface on collapse strength is negligible. If the flaw depth limitations are eliminated when the flaw angle exceeds 60 degrees or when accurate fracture strength has to be evaluated such as in reliability evaluation, the plastic collapse evaluation methods that take into account the effect of pressure acting on the flaw surface should be employed.
Evaluation of Component Cooling System Based on Optimization Algorithm in HPR1000
Authors: Yu Pei , Ting Hou , Yaguang Liu 
Published: November 2024

Abstract
Aiming at large design margin and low water temperature in winter of Component Cooling System in NPP, an innovative design method is provided, which can be used to determine the system configuration and operation scheme in the overall design stage. Firstly, based on the basic principle of system heat balance, the thermal evaluation model is established. Based on the guiding ideology of improving economy as much as possible under safety and operation conditions, an economic evaluation model is established. Then, based on the principle of optimization design, the multi-objective optimization algorithm and analysis program are developed; Finally, the above program is applied to deal with the multi-objective optimization problem, the scheme evaluation is realized. The optimization evaluation results of Component Cooling System HPR1000, which improves the economic efficiency of the power plant under construction. This method breaks through the traditional single line design process which is too dependent on design experience and insufficient quantitative analysis. In the adjustment of multi-disciplinary and multi variable design scheme and parameter optimization, the method of software decision-making and manual decision-making is adopted to shorten the design time and improve the accuracy.
Analysis on the Leakage Behavior of Cable Facility Using Gothic and Gothic 3D Models
Authors: Jia Ningxi , Liu Zhuo , Yuan Yidan 
Published: November 2024

Abstract
In this paper, the laminar leakage model in the leakage module of GOTHIC8.3 code is used to simulate the leakage behavior of Containment thermalAlhydraulic and aerosol Behavior in Light water rEactor (CABLE) test facility caused by penetration. First of all, the leakage rate factor is measured through the leakage rate test performed on the CABLE test facility under reference pressure and temperature. Then, a lumped parameter model and a distributed parameter model are established respectively for CABLE using GOTHIC8.3 code. Both models are modeled with the leakage module. Then, the leakage rate factor measured from the leakage rate test was used to analyze the leakage behavior in the GOTHIC8.3 distributed parameter model. Finally, two models are used to simulate the leakage rate test of CABLE facility, and the results are compared. Results indicate that the two models with the same leakage rate factor have the same total system leaking mass flow. Settings of different leakage surface affect the leaking mass flow position and the local leaking mass flow of cells in leakage surface, but the temperature and flow field distributions of containment vessel have little difference among the cases. With the advantages of the distributed parameter model, it can simulate the leakage behavior more accurately.
Oxidation Behavior and Tensile Property of Inconel 617 in the Simulation Environment of the Very-High-Temperature Reactor
Authors: Du Bin , Liu Zhiyong , Yin Huaqiang , Xiao Penghui , Zhang Huang , He Xuedong , Ma Tao 
Published: November 2024

Abstract
In this work, Inconel 617 was corroded at 950 °C up to 100 h in the impure helium and air atmospheres. The results show that the alloy oxidizes significantly more in air than impure helium, especially internally. There are two reasons for the difference in the oxidation of the alloy in the two atmospheres, one is the difference in the oxidizing gas content of the two, with air containing 21% O2 and only a small amount of H2O in the impure helium. The second is the occurrence of microclimatic reactions, which prompts the oxides that have been formed to occur reduction reaction. Differences in oxidation inevitably lead to differences in the tensile properties of the alloy. The results of the tensile tests shows that the yield strength (YS), ultimate tensile strength (UTS) and elongation (Elong) of Inconel 617 decreased from 309 MPa, 674 MPa, and 43.5% to 291 MPa, 649 MPa, and 39%, respectively, in the impure helium. In the air, the YS, UTS, and Elong of Inconel 617 were 299 MPa, 535 MPa, and 30%, respectively. One of the most important factors affecting degradation is oxidation, as elongated oxides (Al2O3) precipitating along the alloy grain boundaries can preferentially become the core of crack propagation and accelerate alloy failure. Therefore, to ensure the long-term safe operation of the reactor intermediate heat exchanger, the precipitation of excessive internal oxides from the alloy should be avoided as much as possible. controlling the CO content in helium is a simple and useful way to inhibit the oxidation of the alloy. A simple way to reduce the degree of oxidation of the alloy is to control CO in addition to minimizing the amount of oxygen and water in the helium.
Enabling Passive Scalar Wall Modelling in Large Eddy Simulation for Turbulent Flows at High Schmidt or Prandtl Numbers
Authors: Wong Kin Wing , Mickus Ignas , Grishchenko Dmitry , Kudinov Pavel 
Published: November 2024

Abstract
This study investigates near-wall diffusive flux modeling for passive scalar transport in turbulent flows with high Schmidt (Sc) or Prandtl (Pr) numbers. Under these conditions, the diffusion boundary layer becomes significantly thinner than the velocity boundary layer. Capturing the concentration boundary layer presents challenges due to additional scaling in the viscous-diffusive regime. For DNS, mesh resolution requirements to capture passive scalar behavior near the wall are more stringent than those for Kolmogorov scales in pure hydrodynamics investigations. Consequently, wall-resolved approaches in both RANS and WMLES demand excessive wall refinement, limiting their practicality for high Reynolds numbers and industrial applications. In this work, we focus on turbulent flow without an adverse pressure gradient.
Existing wall models fail to provide accurate estimates of wall diffusive flux for passive scalar transport at high Sc. This failure arises from the breakdown of the assumption of eddy diffusivity asymptotic behavior. Using such models for simulating surface processes (e.g., flow-accelerated corrosion) in RANS and WMLES can lead to non-negligible errors. Our study introduces a two-layer scalar diffusivity model to enhance wall modeling capabilities in passive scalar transport at high Sc or Pr numbers.
Design and Research of Radioactive Waste Gas Treatment System for Medical Isotope Test Reactor
Authors: Yongkang Zhang , Jinsong Zhang , Huiqing Yang , Li Chen , Yunming Chen 
Published: November 2024

Abstract
The radioactive activity concentration and gas composition of hydrogen-containing waste gas of medical isotope test reactor are very different from those of pressurized water reactor. The inert gas has higher radioactive activity concentration and contains a lot of nitrogen oxides. The difference of exhaust gas source term between test reactor and other reactor types and its cause and influence on design are analyzed. The reasonable emission control value is determined by reference to GB6249 and the site status. Based on the difference of the half-life of inert gas nuclides and the dynamic adsorption coefficient of activated carbon adsorption, the storage decay treatment of equal inert gas nuclides with short half-life (87Kr, 88Kr) and the retention adsorption treatment of 133Xe with relatively long half-life are adopted, giving full play to the advantages of the two processes, and greatly reducing the size and cost of equipment. The theoretical analysis and experimental demonstration of the effect of nitrogen oxides on the inert gas retention of activated carbon were carried out. The results showed that nitrogen oxides had no effect on the inert gas retention of activated carbon. Based on the analysis and demonstration of radioactive waste safety and system economy, the main equipment design and item classification are completed. At present, the design scheme has successfully passed the design review and safety analysis review.
Research on the Radiation Monitoring Methods in the Reactor Coolant Pressure Boundary Leakage of Nuclear Power Plant
Authors: Yang Zhenlei , Liu Chengmin , Jiang Tianzhi , Li Jin , Zhuo Xianglin , Zhu Hongliang , He Liang , He Zhengxi , Peng Jinqiu 
Published: November 2024

Abstract
Reactor coolant pressure boundary is one of the important protection barriers in nuclear power plant. The weak parts of reactor coolant pressure boundary, such as crack or fissure of the process pipeline, can present hidden troubles, and it can cause coolant leak and can produce a great threat to the safe operation for nuclear power plant, so it is necessary to pay attention to monitor the reactor coolant pressure boundary leakage.
With the increasing requirements on the quantitative monitoring level of reactor coolant pressure boundary coolant leakage, many kinds of radiation monitor methods (including the particulate iodine and noble gas radiation methods, 13N monitoring method and 18F monitoring method), have been conducted for the researchers. In this paper, the traditional particulate, iodine and noble gas radiation methods (short for PING monitoring methods) will be introduced in brief, including the basal principle, system design flow and so on. Because the monitoring objects of PING monitoring methods are corrosion products and fission products, it can only achieve qualitative monitor. The nuclear power plant operators pay more attention to the quantitative leakage condition, so 13N and 18F has attracted the attention of many researchers due to their existence form and their characteristic of positron-electron annihilation. Based on the advantages of 13N and 18F, the key technologies research has been carried out in Nuclear Power Institute of China. A full-range measurement method is proposed. More specifically, the radioactive activity measurement has been studied when nuclear power is greater than or equal to the set value (for example, 20%FP). In order to improve the accuracy of measurement, the coincidence detection and digital coincidence measurement technique has been adopted. By measuring the charge and time of the detectors’ output signals, the effective coincidence events based on time-to-digital converter (TDC) and high-speed ADC technology can be measured. In addition, the total gamma measurement method has been studied when nuclear power is less than the set value (for example, 20%FP). The digital multichannel pulse height analysis method has been studied. The energy spectra and linear fitting parameters are tested, the test results indicate that the measurement system has a good energy linearity. Through the above research, 13N and 18F monitoring methods have been proven to have advantages of quantitative monitoring compared with PING monitoring methods. At last, the recommended radiation monitoring methods of the reactor coolant pressure boundary leakage in nuclear power plant is given considering the advantages and disadvantages of different monitoring method.
Risk Assessment of the Integration of a Large-Scale Hydrogen Cogeneration Facility Into a VVER Power Plant
Authors: Diaz-Pescador Eduard , Viebach Marco , Gamaleja Florian , Hurtado Antonio 
Published: November 2024

Abstract
This manuscript presents the outcomes from the quantitative risk assessment (QRA) of the integration of a generic hydrogen cogeneration facility into the protected area of a reference nuclear power plant (NPP). The latter is Rivne NPP in Ukraine, and the hydrogen production plant (HPP) is assumed to use electricity from the NPP to operate in base-load mode.
The presented study applies HAZID technique to identify those postulated initiating events (PIEs) that lead to the worst-case scenario in the HPP and therefore major external hazard to the NPP, which is regarded as an hydrogen explosion. The study quantifies likelihood of occurrence of the two identified PIEs and potential accident scenarios. It evaluates as well the consequences of the worst-case scenario by means of empirical and analytical expressions in the process safety literature, together with conservative conditions.
The major external hazard is the storage facility due to the large hydrogen inventory. Nevertheless, storage is assumed outside of the NPP perimeter, and safety distances can be adjusted accordingly. In the electrolyser facility, a detonation of a hydrogen vapor cloud in the gas processing unit would only violate the 10 kPa criterion over the nearby standby diesel generators, under upper-bound conditions. An explosion of the buffer tank would have more severe consequences, requiring mitigation measures such as blast walls to ensure an adequate safety level under such conservative accidental conditions.
Exergy Analysis of a Small Modular Reactor Nuclear Power Plant Under Constant and Pure Sliding Pressure Operation
Authors: Wang Xin , Zhao Gang , Qu Xinhe , Yang Xiaoyong , Wang Jie 
Published: November 2024

Abstract
This study proposes a pure sliding pressure operation (SPO) scheme for a small modular reactor (SMR) nuclear power plant (NPP). The operation range considered is from 30 % to 100 % full power (FP). Subsequently, exergy analysis was conducted under constant pressure operation (CPO) as well as SPO. For this, a thermodynamic model of the SMR NPP had been established using EBSILON software, version 16.0. Exergy destruction and efficiency were estimated for the entire system and variable components. The results reveal that control valve throttling under CPO contributes significantly to turbine exergy destruction, accounting for 31.11 % at 30 %FP condition. This issue can be effectively avoided by implementing SPO. Among all components, the reactors have the highest exergy destruction, followed by the steam generators (SGs) and the turbine. Switching the secondary circuit to SPO results in exergy efficiency improvements of 0.16 %, 0.56 %, and 0.97 % at 90 %, 60 %, and 30 %FP conditions, respectively. Additionally, feed water pumps (FWPs) power consumption decreases by 10.29 %, 40.52 %, and 68.92 % respectively at 90 %, 60 % and 30 %FP conditions. However, there is little change in the exergy destruction of the primary circuit components. The research can serve as a valuable reference for optimizing the thermal performance of the NPP.