bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Steady States and DLOFC Accident Analysis of HTR-PM in the OTTO Operation Scheme
Authors: Liu Songyang , Li Xuelin , Luo Yong , Liu Wei , Zhou Qin , Gu Cheng , Zhang Tian 
Published: November 2024

Abstract
High Temperature gas-cooled Reactor – pebble bed module (HTR-PM) located in Shidao Bey, Shangdong province, China, which is the first commercial HTR in the world, adopts spherical fuel element with TRISO particle and graphite material as moderator. As the national demonstration project of China, the outlet coolant temperature of HTR-PM can reach 750 degrees and more than 550 degrees of steam, this reactor can be used not only in electrical power generation, but also in industrial heat source supply, hydrogen production and petrochemical industries. Besides, the module design can decrease the financial cost during the construction period, and shorten the time in manufacturing.
For the online reloading operation, the MEDUL scheme was adopted in HTR-PM operation with 15 fuel batches. To shorten the running-in phase and simplify the reloading system, the OTTO scheme can be adopted to replace MEDUL in HTR-PM operation, in this paper, the steady states calculation based on VSOP-THERMIX has demonstrated the neutronic and thermal hydraulics parameters of HTR-PM in the OTTO scheme from initial critical situation to equilibrium states.
With different assumed OTTO operation schemes, the spatial displacement of max power density and max fuel temperature in the core is shown in every loading step. Meanwhile, due to the moderation ability of graphite, the isotope content of uranium and plutonium compared with different loading methods.
Last, based on the coupled calculation of equilibrium states, the maximum fuel temperature analysis of DLOFC in every loading stage by THERMIX demonstrates HTR-PM with OTTO scheme still keeping inherent safety features. The results show the fraction of fuel elements in different loading stages has an effective influence on maximum fuel temperature.
Optimization of Inventory Control Strategy for Part Load Operation in a sCO2 Recuperated Brayton Cycle for a 5 MWth Heat Pipe Micro Modular Reactor
Authors: Peiretti Matthias , Hofer Markus , Buck Michael , Meucci Ruggero , Starflinger Jörg 
Published: November 2024

Abstract
This study focuses on the implementation of an inventory control strategy for part-load operation in a 5 MWth Recuperated supercritical carbon dioxide (sCO2) Brayton Cycle, integrated with a Heat Pipe-Cooled Micro Modular Reactor (MMR). A preliminary investigation emphasises the crucial role of the inventory tank design for efficient control.
However, when relying solely on the inventory tank for control, matching lower power demands is challenging due to the low pressure at the compressor inlet. Without countermeasures, subcritical conditions are reached when the power decreases below 43% of the nominal capacity of the studied system. To overcome this limitation, a combined control strategy which involves increasing the inlet temperature of the compressor is investigated. This allows to reach 8% of the nominal power.
The study further investigates the performance of the system under load-following conditions. An aggressive power rate change of 7.5%/min of the full load until 30% of the nominal power is simulated, with the resulting power output closely tracking the demand.
Optimal Design of Supercritical CO2 Power Cycle for High Temperature Gas-Cooled Reactor
Authors: Zhou Yujia , Zhang Yifan , Li Hongzhi , Yao Mingyu 
Published: November 2024

Abstract
The supercritical carbon dioxide (sCO2) Brayton cycle has been regarded as a promising technique for future energy conversion systems. sCO2 recompression cycle configuration has been widely studied for nuclear reactors with the temperature difference between the reactor inlet and outlet restricted to 200°C. However, this cycle configuration may be not suitable for high temperature gas-cooled reactors (HTGR) with a temperature difference up to 500°C. Herein, we aim to develop an optimal sCO2 cycle configuration for HTGR with a broad temperature span, which has both high cycle efficiency and sufficient mechanical security guaranteed by energy cascade utilization. We propose a recompression cycle combined with reheat and expansion technologies, and perform a parameter optimization in Fortran platform. First, the thermodynamic models of the system are established. Then, the parametric analysis is conducted about the effects of split ratio, temperature difference of the heat exchangers, inlet parameters of turbines and compressors, and isentropic efficiencies on the cycle performance. The sCO2 cycle tailored for HTGR can achieve a cycle efficiency of 44.98%. Our work not only highlights the design criteria of temperature-match between the HTGR and sCO2 cycle but also guides the numerical assessment of the steady-state performance.
Construction and Exploration of a Diagnostic Platform for High-Temperature Gas-Cooled Reactor Units
Authors: Mao Cui , Liu Haisheng , Geng Di , Liu Yuhua , Yin Zaizhe 
Published: November 2024

Abstract
The high-temperature gas-cooled reactor (HTGR) is an advanced fourth-generation nuclear reactor technology known for its advantages in safety, efficiency, economic viability and versatile applications. The secure and stable operation of HTGR is crucial for ensuring sustainable energy supply. The HTGR unit diagnostic platform integrates functions such as data collection, fault recognition and predictive analysis to provide comprehensive support for the operation and maintenance of HTGR. This paper begins by reviewing the background and development of HTGR technology and subsequently discusses the architectural design, key technologies and practical application effect of the diagnostic platform. Through the development and utilization of this platform, operators can monitor the HTGR status in real time, accurately predict potential faults, and thereby ensure its safe and efficient operation.
Validation of Doppler Reactivity Feedback in SPERT-III E-Core With the Best-Estimate Transient Code TRAC Toshiba Version
Authors: Egawa Tohru , Tokashiki Mikio , Miyaji Takamasa , Fukunaga Takanori 
Published: November 2024

Abstract
We have developed a best-estimate transient code TRAC Toshiba version, which is Toshiba’s advanced version of the TRAC code for BWR. This code is expected to simulate thermal hydraulic and neutron kinetics phenomena three-dimensionally with high accuracy for BWR transient events. For the validation of the code, reactivity accident tests in cold-startup of The Special Power Excursion Reactor Test III (SPERT-III) E-core were calculated and compared with experiment results. Two lattice code “NEUPHYS” and “TGBLA” were used for the calculation. “NEUPHYS” developed by Nuclear Fuel Industries, Ltd. is a neutron transport calculation code with Method of Characteristics. “TGBLA” is a current licensing code of Toshiba. For the confirmation of properly simulation of initial core condition of the tests, static nuclear characteristics of the core were calculated. keff, spatial power peaking, ∧/βeff, worths of control rod and transient rod were compared with experiment results, and good agreements were confirmed. In the transient tests, estimated C/E-1 of peak power without simulation error in transient rod worth were −2%±4% for “NEUPHYS”/TRAC and 1%±4% for “TGBLA”/TRAC in high reactivity insertion tests more than 1$ excluding test 20 which is considered outlier case (±4% were evaluated from uncertainty of experiment). In high reactivity insertion transient tests, temperature of UO2 pellet behaves adiabatically by the time of power peak. Therefore, temperatures of UO2 pellet at the power peak are expected to contain small amount simulation error, it is expected that doppler reactivity simulation error are dominant in aforementioned simulation errors.
Study on Fluid-Elastic Instability of Tube Bundles in Cross Flow Based on Spatiotemporal Coherence
Authors: Cheng Yuxuan , Bu Shanshan , Kai Guo , Chen Deqi 
Published: November 2024

Abstract
The flow-induced vibration (FIV) of tube bundles of steam generator is a complex fluid-structure interaction with coupling of multibody. With the impact of the surrounding tubes, the vibration of the tube is excited by forces with different frequencies, and the vibration of tube bundles turns to be nonlinear. At the same time, due to the existence of asymmetry in the support of the tube bundle, resulting in different intrinsic frequencies in the streamwise and transverse directions of the tube bundle, which can result in the difficulty in perdition of fluid elastic instability (FEI). In this work, the fluid elastic instability of tube bundle with different frequency ratios Rf is investigated through water tunnel experiment and numerical simulation. The target tube arrays are rotated triangular arrays with pitch ratio of 1.33 and 1.41. The effects of frequency ratios on the amplitude and frequency response of the tube bundle are analyzed, while the coherence is induced into the analysis of the route to instability. The results show that the critical velocity can be affected by frequency ratios Rf, but the effect of Rf differs in tube arrays with different pitch ratios. The evolution of Lyapunov exponents shows the nonlinearity of the vibration can be influenced by Rf. The method and results of this work can be helpful in the prediction critical velocity of FEI in tube bundles in cross flow.
Dynamic Modeling and Characteristic Analysis of sCO2 Brayton Cycle System
Authors: Yun Shichang , Zhang Dalin , Li Xinyu , Lv Xindi 
Published: November 2024

Abstract
The supercritical carbon dioxide Brayton cycle has advantages such as high thermal efficiency, inherent safety, simple structure, and compact space at medium to high heat source temperatures. Being able to cooperate with nuclear energy, solar energy, fossil energy or other heat sources for power generation has enormous development prospects. The FUSTAR system proposed by Xi'an Jiaotong University, utilizing the high core outlet temperature of the fluoride salt reactor and the supercritical carbon dioxide Brayton cycle system, has the characteristics of high efficiency, high safety, compact structure, and anhydrous cooling. It proposes a new solution for power supply in remote areas. However, the SCO2 Brayton cycle system is a highly integrated system, and its dynamic characteristics are very complex. There is an urgent need to establish a dynamic model of the SCO2 Brayton cycle system to lay the foundation for subsequent control system design. Based on the conservation equations of mass, momentum, and energy and the thermal hydraulic model, heat exchangers, regenerators, and coolers were established. A lumped parameter calculation model for turbine compressors has been established. A SCO2 Brayton cycle system model was built based on the MATLAB/Simulink platform, and the dynamic response of each device under disturbance was calculated. The results indicate that there are significant challenges in the stable operation of the SCO2 Brayton cycle under the most efficient operating strategy. The inlet parameters of compressors and turbines are highly sensitive to the outlet fluid pressure and temperature, and other auxiliary equipment is needed to ensure relatively stable inlet conditions for turbomachinery. The thermal inertia of the heat exchanger is relatively large, and the control response of the SCO2 Brayton cycle system requires a long time. The system responds slowly to changes in turbine and compressor speed and cannot be quickly adjusted. Rapid adjustment devices such as turbine throttle valves are required to respond quickly to changes in system load. The modeling and analysis method of SCO2 system can also be used for other configurations.
An Experimental Study on the Scrubbing Characteristics of Graphite Dust Aerosol in HTGRs
Authors: Wang Yating , Zhang Yiyang , You Zhikai , Zhao Shumiao , Fang Zhu , Sun Libin 
Published: November 2024

Abstract
The potential release of graphite dust of HTGR contributes the most to the source term and has a great impact on the safety of reactors. In water ingress accidents, the graphite dust may be released to the containment during the overpressure discharge, which not only contributes to the source term but also makes trouble for post-accident measures. An important improvement for reducing the radioactive material emissions in the light water reactor (LWR) is the aerosol pool scrubbing. The effectiveness of pool scrubbing with sphere particles has been extensively investigated. whereas, there is little research on the passive removal of dust and the source term for HTGR. This paper explores the possibility of pool scrubbing retaining the graphite particle and fission products in the overpressure discharge. The scrubber system is established and the pool scrubbing experiments with flake graphite particles and spherical silica particles are conducted. The effects of various parameters on graphite scrubbing efficiency are investigated, including water submergence depth, gas flow rate, particle size, and particle mass concentration. The decontamination factor (DF) is a key parameter to describe the scrubbing efficiency. It is shown that the DFs increase significantly with the increase of submergence depth and gas flow rate both for two kinds of particles. Dominated by the inertial deposition mechanism, the decontamination efficiency of graphite is significantly lower than silica particles for particles larger than 1 μm. As for the particles smaller than 1 μm, the difference in decontamination efficiency between graphite and silica is small attributed to the Brownian diffusion mechanism. In general, the total DF of silica is significantly higher than that of graphite particles. To improve the retention efficiency of particles, a metal packing breaking element is placed in the vessel. It benefits the breakup of large globules and the contacting of gas/liquid. In case, the total DF of graphite aerosol increased by an average factor of 2.8 times, meanwhile the pressure drop increased only by 12.4%. It is revealed that almost all particles achieve a DF of 20, corresponding to a retention efficiency exceeding 95%, at the water depth of 110cm. These results are beneficial for the design of scrubber structures that can accommodate low-pressure discharge airflow, which can achieve a high radioactive retention efficiency and maintain a relatively low pressure drop.
Simulation Study of Turbine Trips Without Scram for CPR1000 Nuclear Power Plant Employing MSHIM Control
Authors: Wu Shifa , Chai Yunzhi , Wan Jiashuang 
Published: November 2024

Abstract
In nuclear power plant operations, several signals may trigger the closure of turbine valve, known as turbine trips. It is desirable that these turbine trips would not cause reactor scram to avoid iodine pit and thus save time and work to restart the reactor. This study investigated the turbine trip without scram transients of CPR1000 Nuclear Power Plant (NPP) implementing Mechanical Shim control system. The Nuclear Steam Supply System (NSSS) consisting of the one-dimensional reactor core model, the nonequilibrium three-region pressurizer model, the nonlinear dynamic model of U-tube steam generator was first modeled. The Balance of Plant (BOP) model including the turbine model, the condenser model, the reheater model and the deaerator model was then established. Based on that, the full-scope simulation platform of CPR1000 NPP was constructed with relevant control systems integrated on MATLAB/Simulink. Meanwhile, the protection system of CPR1000 was also considered in which different real-time protection signals will be collected and processed, and once there is one or more signals were triggered, the protection system will be triggered and shut down the reactor for safety reasons. Finally, the most serious turbine trip transient from 100% Full Power (FP) were initiated and simulated. Results showed that this CPR1000 NPP configured MSHIM control can withstand the above transient as well as provide enough safety margins without leading to a reactor scram.
Assessment of Emergency Planning Zone for Small Modular Pressurized Water Reactors Based on the Risk-Informed Methodology
Authors: Wang Mengxi , Wu Nan , Cao Qun , Yan Jin , Gao Jiaxuan , Zhang Jiemin , Xue Na 
Published: November 2024

Abstract
More than 80 Small Modular Reactors (SMRs) designs are under development worldwide. SMRs usually have the characteristics of integrated design, modular manufacturing and installation, high safety, and proximity to users, and have broad prospects for development and application. Due to the lower power levels and passive safety features of SMRs, the radioactive releases, both in magnitude and frequency, would be significantly lower than that of large reactors. Too large emergency planning zone (EPZ) size may limit the flexible deployment of SMRs. How to reasonably determine the EPZ size of small reactors has become a focus in the nuclear industry. Based on the risk-informed methodology recommended by IAEA-TECDOC-1652, considering both the frequency of the entire spectrum of accident scenarios and the dose consequence for each scenario, this paper analyzes and calculates the EPZ size of a certain small modular pressurized water reactor in China using the Level-2 PSA (Probabilistic Safety Analysis) source terms and the site actual meteorological data. Based on the realistic models, the preliminary recommended EPZ size is about 700m. Even in the conservative meteorological conditions, the EPZ size is about 2.7km, which shows a significant reduction than that of large reactors. The risk-informed methodology is an application of Level-3 PSA supporting in the EPZ sizing. The results could further improve the public’s acceptance and recognition of SMRs and promote their safe and reliable deployment worldwide.
Development of Two-Dimensional Heat Conduction Model in Subchannel Code Based on OpenFOAM
Authors: Dong Zhengyang , Liu Kai , Wang Mingjun , Wu Junmei , Tian Wenxi , Qiu Suizheng , Su Guanghui 
Published: November 2024

Abstract
Subchannel codes play an important role in the thermal-hydraulic analysis of reactor cores. Based on the open-source CFD software OpenFOAM, the research team developed CorTAF, a finite volume method-based thermal-hydraulic analysis code for the core subchannel level. All the CorTAF codes are done on OpenFOAM, which has the function of calculating different phenomena in different reactor types. Compared with the traditional subchannel codes, CorTAF utilizes the advanced discrete algorithm of OpenFOAM to distinguish different plane directions on the flow, and implants more advanced models to calculate the phenomena in the reactor core, which has many advantages such as higher computational accuracy, faster computational efficiency, and better developability. Most of the subchannel codes ignore the axial heat transfer when calculating the temperature of the fuel rod, and the corresponding error is introduced here. In this paper, the axial heat conduction of fuel rods is further studied on the basis of CorTAF thermal-hydraulic model. On the basis of the previous model which only considers the heat conduction of radial nodes, the axial heat conduction is considered, the construction and solution of two-dimensional heat conduction matrix is realized, and the two-dimensional unsteady heat conduction model of fuel rod is developed.
The developed two-dimensional unsteady heat conduction model is used to calculate the normal and blocking conditions respectively, and compared with the previous node heat conduction model, it is found that: Under normal operating conditions, there is little difference between them, and the maximum error is less than 0.5K. Under the condition of blocking accident, the overall temperature distribution trends of the two models are basically the same, but the temperature distribution of the two-dimensional heat conduction model is more uniform, and the maximum temperature of the fuel pellet is smaller. In a word, the upgrade of CorTAF fuel rod heat conduction model further improves the accuracy of core thermal-hydraulic calculation, and provides support for its calculation of special working conditions such as blockage accident, which can be used for reference in similar program development.
Development of DLOFC Accident Analysis Function of DAYU
Authors: Dou Haoming , Lang Minggang , She Ding , Zhang Han 
Published: November 2024

Abstract
As a typical design basis accident (DBA) of High Temperature gas-cooled reactor (HTR), the Depressurized Loss Of Forced Cooling (DLOFC) accidents which will result in higher peak temperature receive more and more attention. The dynamics code for accidental system utility (DAYU) for HTR is developed by the Institute of Nuclear and New Energy Technology (INET) to conduct reactor accident analyses. DAYU is designed for thermal design and steady-state performance study, so the change of system pressure caused by the reduction of coolant inventory in the primary circuit can’t be characterized. In this work, the loss mass of coolant is introduced into the system pressure equation to characterize the pressure change process. In effect, the mass loss rate caused by the break will not only lead to the decrease of system pressure but also affect the flow direction of coolant in the primary circuit due to the different locations of the break, which will result in different temperature of the coolant that leaks out. Taking this phenomenon into consideration, the break mass flow rate is covered in the core flow field calculation further, which makes the simulation of the change of flow field in the reactor caused by the break mass flow rate become a reality. With these improvements, the DLOFC accidents analysis function of DAYU has been accomplished.
Time Series Analysis With Combined Learning Approach for Anomaly Detection in Nuclear Power Plants
Authors: Dong Feiyan , Chen Shi , Demachi Kazuyuki , Yoshikawa Masanori , Seki Akiyuki , Takaya Shigeru 
Published: November 2024

Abstract
Timely and accurate anomaly detection is of utmost importance in ensuring nuclear safety to guarantee the normal operation of nuclear facilities and to prevent severe accidents. Due to the complexity of the dataset from the condition-based monitoring, advanced data-driven deep learning algorithms with their state-of-the art performance, are deployed for automatic anomaly detection in nuclear power plants (NPPs). Nevertheless, in the collected dataset, the knowledge about whether some certain periods are in an abnormal or normal state is available, while for a greater portion of the time, only data in unknown operating states can be acquired, rendering the general supervised learning (SL) and unsupervised learning (UL) suboptimal learning methods. Therefore, a combined learning approach is deployed in the proposed time series analysis framework, allowing the utilization of additional domain knowledge within the dataset to train the framework from SL to UL seamlessly. Additionally, due to the intricate inherent relationships among different instruments and the low frequency of anomaly occurrences, the implicit influential weights of each dimension vary when identifying different kind of anomalies. To tackle this challenge, an attention mechanism is integrated into the anomaly detection model for highlighting the prominent abnormal features and hence enable it to address the subtle impact of each dimension to enhance the ability of anomaly detection task. Experimental evaluations on the real-world multivariate benchmark dataset Security Water Treatment (SWaT) and on the high temperature gas-cooled reactor (HTGR) anomaly cases dataset based on the analytical code ‘ACCORD’, demonstrate the feasibility and effectiveness of the proposed approach.