bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on Influence of Nonlinear Characteristics of Laminated Rubber Bearings on Reactor Buildings Including SMR
Authors: Kitagawa Keito , Furuya Osamu , Okamura Shigeki 
Published: November 2024

Abstract
The development of small modular reactors (= SMR), which are safer and more rational than current nuclear power generation facilities, is underway. As a countermeasure against earthquakes, it is assumed that a laminated rubber bearing for seismic isolation system will be installed.
It is well known that laminated rubber bearing has nonlinear mechanical properties in a large deformation region.[1] Therefore, searching for conditions where the effects of the nonlinear mechanical properties of laminated rubber bearing are likely to appear through modal analysis. Then, a nonlinear time history response analysis and a floor response spectrum are calculated for the selected conditions. The results will evaluate the effects of the nonlinear mechanical properties of laminated rubber bearing on reactor buildings and internal equipment.
As a result of the evaluation, when the mass ratio of the base plate and the building is equal in the horizontal direction, the seismic isolation period is 2.00 s and the building vibration frequency is 4.00 Hz, and in the vertical direction, when the mass ratio is equal, the isolation frequency 20.0 Hz and the reactor building natural frequency 20.0 Hz, that was confirmed that the influence of nonlinear dynamic characteristics is noticeable.
United States NRC Modernization Efforts to Enable the Safe Use of Digital Instrumentation and Controls in Nuclear Power Plant Safety Systems
Authors: Darbali Samir , Taneja Dinesh , Martinez Rodriguez Erick , Blas Rodriguez Gilberto 
Published: November 2024

Abstract
In the United States (U.S.), licensees for operating nuclear power reactors and applicants for new and advanced reactor designs are pursuing the effective and efficient use of digital technologies for their instrumentation and control (I&C) systems. As such, it is necessary that the U.S. regulatory infrastructure for the safe implementation of digital I&C (DI&C) technology in nuclear power reactors is clear and efficient. Over the past few years, the U.S. Nuclear Regulatory Commission (NRC) staff, in collaboration with nuclear industry stakeholders, has extensively modernized its DI&C regulatory infrastructure to enable the safe and secure use of DI&C technology in nuclear power plant safety systems and components. This paper will describe the NRC’s vision for a modernized DI&C regulatory infrastructure and the significant accomplishments from recent years to improve the clarity and reliability of the DI&C regulatory infrastructure. This paper will also describe ongoing regulatory improvement activities as well as ongoing and planned NRC staff research efforts related to the use of new technologies in nuclear power plant safety systems.
Code-to-Code Verification of Thermal Hydraulic Subchannel Code LINDEN
Authors: Wang Zixuan , Zhu Yuanbing , Wang Yan 
Published: November 2024

Abstract
The subchannel code LINDEN is developed as a dedicated numerical tool for thermal-hydraulic analyses of a pressurized water reactor (PWR) under varying operating conditions. To fully achieve the development goals of LINDEN, rigorous verification and validation (V&V) work should be included in preparation for its intended applications in engineering practice. This research focuses on a code-to-code verification of the code, where a top-down verification strategy has been adopted. The main features and basic models of LINDEN were first introduced. In order to verify the concerned models, a verification matrix for separate-effect test and integral-effect test with different typical thermal-hydraulic scenarios was designed. The test cases involved a typical single channel in a PWR primary circuit and a 5 × 5 rod bundle under both steady-state and transient conditions. The simulation results of several key thermal-hydraulic parameters were then compared with those from the state-of-the-art subchannel code CTF. The calculation results of LINDEN were in good agreement with those of CTF under different test cases, and only the void prediction of LINDEN is on the lower side compared to CTF’s, which could be attributed to different interface mass exchange models taken in the two codes. In summary, the code-to-code based assessment work in this study has confirmed that the basic models are correctly implemented and performed as expected.
Customising a System Code for the Analysis of the Thermal-Hydraulic Behaviour of a Supercritical Pressure Light Water Small Modular Reactor
Authors: Chaaraoui Omar , Pucciarelli Andrea , Ambrosini Walter , Otic Ivan , Schulenberg Thomas , Allison Chris , Fu Zheng 
Published: November 2024

Abstract
The paper reports about the work performed in customizing the RELAP/SCDAPSIM code features for an easier assessment of a conceptual design of a Small Modular Reactor (SMR) operating at supercritical pressure proposed by Schulenberg and Otic [13] in the frame of the EU ECC-SMART Project. First unsuccessful attempts to make use of an NRC version of the RELAP5 code in the analysis of the thermal-hydraulic behaviour of the SCW SMR revealed numerical difficulties at the transition from supercritical to subcritical pressures. This suggested to activate a cooperation for customising an already existing version of the RELAP/SCDAPSIM code, aiming to overcome similar numerical problems and to implement heat transfer correlations in principle more appropriate for supercritical pressure conditions. The improved version of the code, featuring a user defined generalised correlation for heat transfer encompassing several of the presently available literature proposals, was applied in predicting rod-bundle experiments performed by Razumovskiy et al. [16–17] and in analysing the behaviour of the seven core flow passage reactor proposed in [13]. In the paper, the present capabilities of the code are discussed in front of the obtained results and perspectives for future developments are suggested.
Experimental Study on Metal Jet Spreading on Substrate With Roughness
Authors: Xu Yihua , Yokoyama Ryo , Suzuki Shunichi 
Published: November 2024

Abstract
The behavior of molten corium, particularly when it impinges upon the containment substrate and the further spreading, is important in determining the integrity of the containment structures. In nuclear power plants (NPPs), substrates typically display a variety of surface textures; however, the impact of substrate roughness on corium behavior remains inadequately explored. In this study, the influence of substrate roughness on the corium spreading behavior is investigated. A series of scaled experiments were conducted using Wood’s metal as corium simulant impacting substrates of different roughness levels. The heated corium simulant is discharged on the substrate (sandpaper or concrete) and this process is recorded by cameras. Based on the experimental results obtained, a model is proposed to predict spreading area considering both corium flow rate and substrate roughness. It is concluded that increased substrate roughness generally reduces the spreading area of corium. The results of these experiments provide valuable insights into how surface roughness influences the spreading behavior of corium, offering essential data for enhancing reactor safety measures and containment strategies.
Multi-Objective Optimization Design for Nuclear-Coupled Distributed Multi-Energy System in Alpine Region
Authors: Ren Xiaoxiao , Jiang Chao , Wang Jinshi , Yang Sifan , Zhao Quanbin , Yan Junjie 
Published: November 2024

Abstract
The alpine regions are characterized by its remote location, cold climate and limited conventional energy, which presents considerable challenges in ensuring reliable electricity and heating supply. To this end, a nuclear-coupled distributed multi-energy system (N-DMES) is proposed. To attain the optimal configuration and operation of N-DMES, a multi-objective optimization model is developed with net present cost, shortage of electricity and renewable energy power generation as objective functions. Meanwhile, the improved multi-objective grasshopper optimization algorithm is utilized for the model solving. Besides, this system is compared with a diesel-coupled distributed multi-energy system (D-DMES). This paper takes a market town in an alpine region as a case study to assess the effectiveness and feasibility of the proposed N-DMES. Simulation results indicate that the designed N-DMES can satisfy the users’ electricity demands. N-DMES exhibits slightly lower economic performance and renewable energy consumption compared to D-DMES. However, N-DMES surpasses D-DMES in terms of environmental friendliness and reliability. The minimum load factor of MMR and renewable energy capacity have a certain impact on the system performance.
Numerical Simulation on the Thermal Hydraulic Characteristics in Pool-Type Fast Reactor Based on the New-Designed Passive Dynamic Residual Heat Removal Valve
Authors: Qu Yuanshu , Zhang Yuhao , Zhao Haiqi , Lu Daogang 
Published: November 2024

Abstract
The passive features of the Emergency Heat Removal System (EHRS) is a crucial component of the reactor’s safety design, and its passive safety features has a significant impact on its performance during accidents. The coolant in the new sodium-cooled fast reactor enters the core directly through the control valve and heat transfer pipe during accident conditions. Because the decay heat exchanger (DHX) is equipped with a check valve at the outlet position. This valve can be opened automatically in the event of an accident, improving the reactor’s passive safety features. For the new design model of sodium-cooled fast reactor, a calculation model of new residual heat removal system calculation model is established with a new ball check valve by CFD software. The UDF program was used to simulate the opening and closing of the check valve by jointly simulating FLUENT dynamic mesh technology and transient dynamics. This allowed the spool of the new ball valve to move freely with the fluid. Simulation calculations were conducted to determine the thermal and hydraulic characteristics of the ball check valve under normal operation and accidental shutdown conditions. Additionally, the dynamic process of temperature change was analyzed after opening the passive safety system under accidental conditions to evaluate the effect of the control valve. The results indicate that the new control valve is passively opened due to the temperature difference and buoyancy force. Additionally, the fluid in the DHX undergoes natural circulation throughout the entire core, gradually recovering from standby power to full power operation. When the flow rate of the pump gradually decreases, it would take over the task of cooling the core. This not only enhances the passive safety features of the EHRS but also efficiently conducts the heat of the reactor, and effectively discharges the residual heat of the core.
Bamboo-Frame: An Automatic Modeling Tool for the Two-Step Reactor Core Physics Analysis
Authors: Zhou Yuancheng , Li Yunzhao , Li Yisong , Liang Yilin , Zhang Hengrui , Zhou Shilong , Wang Weiguo , Ou Yuxiang 
Published: November 2024

Abstract
Conventional method of reactor core analysis pivots on the two-step methodology requires the users to have expert understanding on the methodology, to provide significant manual input, and to maintain substantial data consistency. Thereby, the calculation process turns out to be complex and error-prone. In light of this, we developed Bamboo-Frame to automate the modeling procedure in Pressurized Water Reactor (PWR) core analysis. This innovative tool requires users to provide only structural and operational parameters, while effectively managing complex tasks such as lattice recognition, thermal expansion, state matrix setting, core nodes division, and input file generation. This culminates in a user-friendly visualization system, thus significantly increasing accessibility and reducing manual workload.
The application of Bamboo-Frame to CNP650, a PWR with a full power output of 1930 MW, showcased its capability to handle the modeling of 52 fuel lattices and numerous reflector lattices. The software’s robustness and precision were evident when compared to the case without Bamboo-Frame, with the input file size dramatically reduced from 38.6MB to 168KB, underscoring its efficiency and the potential to reduce human error. Validation through startup physical test simulations and power history following simulations proved its remarkable accuracy within industrial standards, alongside statistical analysis demonstrating compliance with engineering standards.
Verification of the RMC-SaraGR Nuclear Design Code System Based on the HTTR Benchmark
Authors: Yuan Yuan , Liu Guoming , Zhang Peng 
Published: November 2024

Abstract
In order to perform a detailed nuclear analysis for the small modular prismatic HTGRs, the RMC-SaraGR nuclear design code system has been developed. The verification of the code system using the HTTR benchmark is reported in this paper. The detailed HTTR models have been established with the continuous-energy Monte Carlo code RMC to provide the reference solutions and the 25-group cross sections, which are further corrected by the super homogenization (SPH) factors before being used for the homogeneous core transport calculations by SaraGR or the multi-group RMC (RMC-MG). Various core configurations including fully-loaded core with/without control rods (CRs) insertion and annular cores with different number of fuel columns loaded have been modeled. Compared with the benchmark data, the reference solutions provided by RMC show a 100∼ 500 pcm overestimation of keff for the fully-loaded cores, which may be attributed to the nuclear data uncertainty and the model bias. When the CR block is homogenized as a whole region, the calculation results from RMC-MG and SaraGR agree well but both show a significant keff overestimation of more than 1000 pcm compared with the RMC reference solution for the fully-loaded core with all CRs withdrawn, which is likely to be related to the strong neutron streaming effect through the void holes of the CR blocks. Therefore, a local mesh refinement is imposed on the CR blocks for an explicit modeling of the CR holes, which reduces the keff discrepancy to less than 500 pcm. Most results agree well with the reference solutions and are within an uncertainty of the benchmark values, indicating that the RMC-SaraGR code system is a feasible tool for the nuclear analysis of the prismatic HTGRs.
Human Reliability Analysis for a Passive NPP and Application in Plant Operating Procedure Optimization
Authors: Qiu Yongping , Tan Xiao 
Published: November 2024

Abstract
Human Reliability Analysis (HRA) is an important element in probabilistic safety assessment (PSA) for nuclear power plants (NPPs). Usually three types of human interactions (HIs) are defined in HRA for NPPs, i.e., pre-initiating event HIs, initiating event-related HIs, and post-initiating event HIs. In this paper, a brief introduction of the HRA methodology for a passive NPP is first presented. Then based on the HRA/PSA results and insights, design improvements and PSA applications could be performed. The optimization of plant operating procedures is one of the PSA applications in plant design phase. The risk-important human actions are obtained from the PSA importance analysis. Next, two risk-important human actions related to SGTR accident are studied in detail with the human error probability versus the design of plant operating procedures. Based on the sensitivity analysis, the operator response time for these two events, which varies with the steps of plant operating procedures, has certain influence on the plant core damage frequency (CDF) and rather high impact on the ratio of CDF variation of SGTR accident. Finally some proposals for optimization of the SGTR event analysis and relevant procedure steps are discussed. The plant operating procedures of the passive NPP are still under developing and optimization, and the study provides a useful way for plant operating procedure optimization.
Numerical Simulation of the Difference Between Steam-Seawater Jet and Steam-Freshwater Jet
Authors: Cao An , Niu Shipeng , Li Jianfa , Hua Yongzhen , Qi Xiangjie , Xiao Jiayu , Meng Zhaoming 
Published: November 2024

Abstract
The seawater reserves on the earth are much larger than freshwater, and the use of seawater instead of freshwater as the cooling medium for steam jet condensation can greatly alleviate the shortage of freshwater resources. However, current research on steam-water jet condensation mainly focuses on freshwater jet, with very few studies on seawater jet. In this paper, the software Fluent was used to compare and analyze the differences in the shape of stable jet condensation plumes, plume penetration length, pressure gradient, and velocity distribution between seawater and freshwater. The results show that the shape and pressure gradient of the steam plume in seawater are similar to those in freshwater, while the plume penetration length and axial velocity are slightly smaller than those in freshwater. Seawater has the feasibility of replacing freshwater as the cooling medium for suppression pool.
Study on Bubble Migration Characteristics in a Lead-Cooled Fast Reactor After Steam Generator Tube Rupture Accident
Authors: Feng Zhenyu , Zhang Dalin , Jiao Haoyu , Chen Yutong , Lin Yue , Tian Wenxi , Qiu Suizheng , Su Guanghui 
Published: November 2024

Abstract
In this paper, the bubble migration characteristics of LFR after SGTR accidents are investigated based on the DPM. Firstly, the geometric and mesh models of the primary coolant system of the LFR are established, and the drag coefficient model of bubbles in the DPM is validated based on the bubble terminal velocities. Secondly, steady-state simulation of the primary coolant system of the LFR are performed to determine the temperature and velocity fields. Thirdly, the migration characteristics of the single bubble released from different positions of the SG are investigated, and the results show that bubbles on the horizontal plane at the window position were more easily entrained away from the window side, while those on the horizontal plane below the window were more easily entrained near the window side. Finally, the bubble groups with different diameters and different positions are studied. Bubbles larger than 0.3 mm leaking at any position of the SG cannot be entrained into the core, and bubbles larger than 0.45 mm will escape directly. Additionally, the higher the location of the rupture the lower the probability of bubbles entering the core.
Hydrogen Production by High Temperature Steam Electrolysis Coupled With a Small Modular Reactor: Cross-Comparison Between Various Thermal Architectures
Authors: Tauveron Nicolas , Haubensack David , Dumoulin Pierre , Alpy Nicolas 
Published: November 2024

Abstract
This paper is part of a study on the coupling of a SMR nuclear unit to a hydrogen production process using high-temperature steam electrolysis (HTSE). Different possible coupling architectures are investigated and their energetic evaluation are performed. A numerical design of experiment is established in order to study the separate effects of the various architectural options envisaged in relation to a reference simulation configuration: with or without thermal coupling, direct or indirect steam supply, location of the extraction point in the SMR Rankine cycle, one-way or two-ways thermal coupling, internal heat recovery in the HTSE, and oxygen content.
The results obtained on the reference simulation configuration show that it is possible to obtain a gain of over 13% compared with the solution involving no thermal coupling between the SMR and the HTSE process (i.e. when HTSE heat is provided through electrical heaters), with a deviation of 3% from an idealized configuration with freely available heat and no utilization constraints (“ideal waste heat configuration”). It even outperforms the heat pump solution by more than 4%.
From a purely energetic point of view, and considering only the HTSE operating at atmospheric pressure, it can be seen that each optimized configuration deviates rather slightly from the reference simulation configuration: all energy performances enable gains of around 13% compared with the configuration with no thermal coupling between the SMR and the HTSE process. The solution with two-ways thermal coupling is also more efficient than the one-way coupling.
Enhancing Hydrogen Laminar Burning Velocity Estimation With Artificial Neural Networks and Simulated Data
Authors: Ambrutis Andrius , Povilaitis Mantas 
Published: November 2024

Abstract
Practically relevant containment scale combustion simulations require fast methods applicable to comparatively sparse meshes, e.g. RANS. Combustion rate in this case can be estimated using simplified progress variable equation closures, for example, Turbulent Flame speed Closure model, and correlations for laminar burning velocity (LBV) instead of chemical kinetics. However, existing correlations may lack accuracy and have restricted validity domains. In the previous work, by developing experimental-data-based Artificial Neural Networks (ANN) model (original), we showed that ANN can be used for fast and accurate prediction of LBV in dry hydrogen-air mixtures from temperature, pressure and equivalence ratio. However, due to limited amount of experimental data available in the literature, database consisted of a relatively low number of experimental datapoints (around 2000). This led to model exhibiting some unphysical behaviour at higher temperatures and specific pressures, requiring more data than is available to enhance its reliability. This work presents an attempt to solve the problem of insufficient database by replacing simulated data instead of experimental. The architecture of new ANN was kept similar to original ANN, ensuring similar prediction time. Instead of a few thousand experimental data points, new model was trained using around 1 000 000 data values generated using chemical kinetics simulation. The larger database improved new model predictions compared to original and led to a more physical LBV behaviour estimation. In addition, new model has an expanded range of applicability. In our estimation, any further improvement in our model accuracy is mostly limited by the number of neurons, restricted by the need for fast ANN calculations.