bluuebunny/crossref_metadata_2025_split huggingface dataset:
Further Study on Real-Time Positioning of Fuel Failures Method in HFETR
Authors: Sun Shouhua , Li Jian 
Published: October 2021

Abstract
Research reactor’s fuel element integrity is essential for nuclear safety. In normal operations, radionuclides could be well contained in fuel element. Unfortunately, under severe conditions, the amount of radionuclides may increase, releasing to the primary coolant and, as a result, crack or damage the fuel element mechanically or chemically. Thus, it is imperative to help operators take measures effectively, locating damaged fuel assembly in reactor cores rapidly. In this paper, a quantitative fuel failure positioning method based on cesium (134Cs and 137Cs) detection was studied. An exact model to describe the relation between cesium radioactivity and fuel burnup based on nuclide conversion chain was put forward. Radioactivity measurement data has approved the validation of this method when it applied to HFETR. Moreover, based on the measurement of 24Na activity, we proposed an approach to identify fuel failures and cladding corrosion. The validation results further indicate the accuracy and good applicability of this method, that provides guarantee for the safe operation of HFETR.
Estimation of Long-Term Ex-Vessel Debris Cooling by Water in Fukushima Daiichi Nuclear Power Plant Unit-3
Authors: Sato Ikken , Yamaji Akifumi , Li Xin , Madokoro Hiroshi 
Published: October 2021

Abstract
Interpretation for the two-week long Unit 3 ex-vessel debris cooling behavior was conducted based on 1F plant and site data such as pressure, temperature, gamma ray level and live camera pictures. It was estimated that the debris relocated to the pedestal was in partial contact with liquid water for about initial two days. With the reduction of the sea water injection flowrate, the debris, both in-vessel and ex-vessel ones, became “dry”, in which the debris was only weakly cooled by vapor and this condition lasted for about four days until the increase of the sea water injection. During this dry period, in-vessel and ex-vessel debris were heat up and it took another four days to re-flood the heated up debris.
Preliminary Core Design of the Solid Moderator Reactor for Investigation of the In-Depth Europa Ice Layer
Authors: Fukizaki Shuta , Yamaji Akifumi , Fukuda Takanari 
Published: October 2021

Abstract
The use of different nuclear reactor systems as power source of deep-space explorers has been studied over the past decades as solar power cannot be expected in deep-space beyond Jupiter. In the preceding study, a cylindrical solid moderator reactor concept of about 500 kg in weight, which consists of 20% enriched UN fuel, YH1.5 moderator and Be reflector without using working fluid, was developed. In this study, we propose to further extend use of the developed core concept for the in-depth ice layer’s investigation of Europa (one of the moons of Jupiter). At the end of the journey, the bare reactor is to be landed on the ice layer of Europa and sink down through the ice-layer as it melts the ice layer with its thermal power. For this purpose, this study aims to flatten the core power distribution to increase the core surface temperature for efficient ice melting while keeping the peak temperature of the core below a design limit. The core neutronics characteristics and power distributions are evaluated with neutron diffusion approximation and ablation and sinking behavior of the reactor through the ice layer is analyzed with Moving Particle Semi-implicit (MPS) method. Among different designs, the Hollow core could reduce the radial power peaking relatively well. The trial analysis by MPS method showed that modeling convective heat loss of the core surface and / or modifications to the core design may be necessary to prevent excess heat-up of the core before it sufficiently melts the ice and sinks down into the ice layer.
Influence of Inlet Turbulent Flow Generated by Periodic Computations on the Pressure Drop and Axial Velocity Distribution Predictions
Authors: Qiu Chufa , Raverdy Bruno , Faucher Vincent , Bergeron André 
Published: October 2021

Abstract
This study presents the CFD modeling of the adiabatic rod bundle test (MANIVEL) with a Simple Support Grid (SSG) at a Reynolds number of 100 000. Calculations are carried out on grid using a one-block fully hexahedral mesh. To reduce the computational power requirements, the computational domain is limited to one span with one SSG. The inlet is a fully developed turbulent flow obtained using periodic computations that are applied on two different domains: an inlet box and a span box. The mean axial velocity distribution at different elevations and pressure drop are compared with measurements. Small discrepancies of the mean velocity distributions and of the pressure drop per span are observed between the two types of periodic inlet conditions. The influence of SSG is not negligible and is still present downstream. The application of periodic conditions generated on span box is closer to experimental data and so is more appropriate.
Affect Analysis of Surface Liquid Film Coverage on the Safety Performance of Containment
Authors: Shi Xingwei , Shaoxin Zhuang , Xinfang Cui , Wei Song , Jiaxu Zuo 
Published: October 2021

Abstract
Under the severeaccident condition, the containment is the last shielding for preventing the accidentalrelease of radioactive fission products to the atmosphere. To figure out the failure margin in accident conditions, it is necessary to research and analyze the effects of surface liquid film coverage on the containment safety. The Passive Containment Cooling System’s thermal-hydraulic model has been built with passive containment analysis code in Large Power Passive Reactor. Based on the reference conditions of Double Ended Guillotine (DEG) on a cold leg and Main Steam Line Break (MSLB), the sensitivity analysis of parameters related to surface liquid film coverage has been performed, the cases of the water distribution tube failure and the spray pipe pluggage added the failure of water distribution tube, and the scram under different power level have been carried out, respectively. Results: The analysis results show that: (1) the containment would fail in 24 hours after DEG accident start as a result of the liquid film coverage lower than 45%; (2)under the MSLB condition, the containment pressure is highest when scram power level is 30% of Fission Power (FP) under keeping the design liquid film coverage, and with the decrease of liquid film coverage, the containment is not failing as a result of overpressure, although the pressure is highest on the scram power level 0% FP.
Implementation and Validation of an Improved Interfacial Area Concentration Model for Two-Phase Flow CFD Simulations
Authors: Zhang Xiang , Peng Minjun , Cong Tenglong , Lu Chuan , Wang Chenyang 
Published: October 2021

Abstract
The interfacial area concentration (IAC) is an important parameter in the calculation of interfacial transfers in two-fluid model, which can affect the accuracy of the boiling simulations. In this paper, an improved IAC model based on drag force and drift velocity is obtained, which can make full use of the experimental data and the models of the drag force and the drift velocity to avoid the shortage of IAC algebraic model in two-phase flow simulations theoretically. The improved model is validated by the DEBORA boiling flow experiment data. The reasonable radial distributions of void fraction, liquid temperature and phase velocity can be obtained, which indicates that the improved IAC model coupled in boiling flow model can be applied in CFD simulation of two-phase boiling flow. The improved model provides a new calculation approach for the IAC in the boiling flow with multi flow regimes.
Study on Main Radionuclides of Liquid Waste in Containment Under Severe Accident
Authors: Wang Shaowei , Chen Haiying , Li Wei 
Published: October 2021

Abstract
After the Fukushima accident, the disposal of radioactive liquid waste has received great attention. The analysis of nuclide activity in the liquid waste of the containment after a severe accident is the prerequisite for the treatment of the liquid waste. Based on the source and reduction mechanism of radionuclides in the liquid waste of the containment after the severe accident, the composition change of nuclides in the liquid waste was studied. The results showed that during the release stage of nuclides from core to containment, the activity of each group nuclides in the liquid waste gradually increased. Then the activity of each group nuclides gradually decreased with time except for the lanthanum group nuclides. The total iodine activity decreased the fastest, and the total activity of the cesium group decreased the slowest. At the beginning of the accident, iodine was the main nuclide in the liquid waste, followed by cesium group. The percentage of cesium group in the total activity decreased first and then increased, while that of other groups first increased and then decreased. 134Cs, 137Cs, 89Sr, 90Sr, 127Tem, 106Ru, 90Y, 144Ce, 241Pu decayed very slowly, which were the important components of the radioactive liquid waste.
Research on Multi-Objective Optimal Design of Plate Heat Exchanger in Nuclear Power Plant Cold Chain System
Authors: Zhao Weiguang , Zhou Xiuan , Shi Jiangwu , Yan Changqi , Wang Jianjun 
Published: October 2021

Abstract
In order to improve the scientificity and rationality of cold chain system design, the plate heat exchanger of a branch of the nuclear power plant cold chain system of nuclear power plants is taken as the research object. Based on the basic principles of optimization design, appropriate optimization calculation algorithms are developed and plate type heat exchangers are developed. The design program of the heat exchanger forms an optimal design analysis tool. Taking a plate heat exchanger as an example, taking its mass, flow resistance on the hot side and flow resistance on the cold side as the optimization objectives, a new hybrid genetic algorithm was used to carry out a multi-objective optimization case analysis. The calculation results show that there is a competitive relationship between the mass of the plate heat exchanger of the nuclear island cold chain system of the nuclear power plant and the flow resistance on the hot and cold sides, and the optimal solution cannot be obtained at the same time. There is a correlation between the flow resistance on the cold side and the flow resistance on the hot side. If the heat exchanger mass is reduced by 6.39%, the flow resistance of the hot side and cold side will increase by 6.55% and 7.71% respectively; if the flow resistance on the hot side and cold side are reduced by 32.22% and respectively 32.01%, its mass will increase by 2.16%; on the premise of keeping the mass of the heat exchanger basically unchanged, the flow resistance on the hot side of the heat exchanger is reduced by 23.84% and at the same time the flow resistance on the cold side is reduced by 19.77%, and the flow resistance on the cold side is reduced by the largest 23.35% and at the same time reduce the flow resistance on the hot side by 12.54%. Therefore, there is a lot of room for optimization in the design of nuclear island cold chain system of nuclear power plant and there are many optimization schemes. The research results can provide reference for the subsequent optimization of nuclear power plant cold chain system design and improve the scientific and economic efficiency of system design.
Sensitivity Analysis of Power Related Parameters in a Reactivity-Initiated Accident of a Molten Salt Reactor
Authors: Wang Chaoqun , Yang Qun , Wang Kai , Jiao Xiaowei , Wu Yanhua , He Zhaozhong 
Published: October 2021

Abstract
Molten salt reactor (MSR) is one of the promising candidate Generation IV reactor technologies for future consideration. The rapid power rise during the accident is usually a critically important threat for the safety of a reactor. In a MSR, the increasing of reactor structural temperature caused by power rise may damage the integrity of reactor pressure boundary. During the accident scenario simulation of a MSR, the trend of reactor power really depends on the initial power, power peaking factors, and fuel salt and moderator temperature reactivity coefficients. This work aims to identify how these parameters influence the consequences of the accident for a MSR. In this study, a reactivity-initiated accident (RIA) is considered as the initiating event, which is caused by the withdrawal of control rod during power operating. This accident will lead to a rapid increase of reactor power and core temperature. A local sensitivity analysis was utilized to analyze the power related parameters. And a wide variety of values were selected for each parameter as the assuming initial conditions of the analysis. Multiple thermal-hydraulic simulation runs through a modified RELAP5 code were performed to acquire the accident analysis results for sensitivity analysis. According to the results of sensitivity analysis: power related parameters have a low impact on the consequence of the reactivity-initiated accident under power condition of the molten salt reactor. And the consequence of the accident is more sensitive to the changes in power peaking factors than the other parameters.
Dynamic Characteristics Analysis of NuScale in Frequency Domain
Authors: Yang Jingrui , Ma Qian , Han Lingtong , Sun Peiwei 
Published: October 2021

Abstract
NuScale is a small nuclear reactor that relies on natural circulation. Its modular production and inherent safety can not only be used to generate electricity in some remote areas, but also provide energy for water desalination and regional heating. However, the dynamic characteristics of the NuScale are different from those of the traditional PWRs because of its passive characteristic. Therefore, it is necessary to study and analyze the system dynamic characteristics of NuScale.
The NuScale PWR model is established based on MATLAB&Simulink. It includes point-reactor kinetics model with six groups delayed neutrons, coolant system and steam generator system. The model is established based on the conservation equations of mass, energy and momentum. And the correctness of the model is verified by the comparison between the steady-state calculation results and the design parameters. Transient calculation results are verified by mechanism analysis.
To evaluate the dynamic characteristics of NuScale, the sine function changes with different frequencies in reactor reactivity and feedwater mass flow rate are introduced. The amplitude and phase responses of reactor power, secondary loop steam pressure, secondary loop mass flow rate, secondary loop steam temperature, and coolant average temperature are recorded. Then the Bode plot can be drawn with amplitude and phase responses in different frequencies. To evaluate the NuScale dynamic characteristics, frequency domain analysis is performed.
Experimental and Analytical Investigation on Local Damage To Reinforced Concrete Panels Subjected to Projectile Impact Part 1: Penetration Damage Mode due to Normal Impact
Authors: Kang Zuoyi , Okuda Yukihiko , Nishida Akemi , Tsubota Haruji , Li Yinsheng 
Published: October 2021

Abstract
Most empirical formulas have been proposed to quantitatively evaluate local damage to reinforced concrete (RC) structures caused by a rigid projectile impact. These formulas have been derived from impact tests performed to the target structure with a normal angle, while only a few impact tests involving soft projectile to the target structure have been studied. Recently, we conducted a series of impact tests to evaluate local damage to RC panels subjected to normal and oblique impact due to rigid and soft projectiles. The final goal of our study is to establish a new formula for evaluating local damage to RC structures caused by oblique impact based on experimental and analytical investigation. This paper summarizes the results of experimental and analytical investigation on penetration damage mode to RC panels subjected to normal projectile impact. Through the comparison between experimental and analytical results, the analytical method is validated.
Experimental Study of Flow Characteristics in Round Jet Flow Using Particle Image Velocimetry (PIV)
Authors: Wu Lei , Wang Jianjun 
Published: October 2021

Abstract
High-resolution particle image velocimetry (PIV) was used to study the flow characteristics in the nearfield of a turbulent round water jet issuing from a circular nozzle. Mean velocity, velocity gradient, turbulence intensity, Reynolds stress and other turbulence statistics are investigated over a range of jet Reynolds number Re (based on jet exit velocity and equivalent nozzle diameter) varying from 531 to 13500. The distribution of parameters of different Reynolds numbers and different height sections was studied. The velocity measurements reveal that all exit jet profiles in the near-field closely approximate a single peak shape, and they show strict self-similarity. Compare turbulence intensity, Reynolds stress with different Reynolds number jet, the self-similarity is shown along jet path when Reynolds number is low, and the self-similarity is shown after the turbulence is fully developed when Reynolds number is high. The shapes of Reynolds stress of high Reynolds number jet change from double peak to single peak along jet path, and the shapes are always single peak at low Reynolds numbers. This critical Reynolds number is about 2000. In addition, the turbulence statistics can be used to correct the coefficients of the turbulence equation and large eddy simulation.
Load Match-Oriented Coordinated Control for Modular High Temperature Gas-Cooled Reactor Based on Dynamic Matrix Control
Authors: Jiang Di , Dong Zhe , Li Bowen , Huang Xiaojin 
Published: October 2021

Abstract
To cope with the flexibility from both load side and supply side, nuclear power generation should provide flexible operation services to improve its economic competitiveness. The prerequisite of flexible operation is the real-time realization of the operation point which is usually achieved by unit coordinated control finding the setpoints of nuclear power, coolant flow and feedwater flow to meet various load demand and keep key parameters within reasonable limits. Modular high temperature gas-cooled reactor (MHTGR) is typically a small reactor and adopt adjustable helium flow, graphite and once through steam generator (OTSG) as coolant, moderator, heat exchanger, respectively. The thermal hydraulic characteristics of MHTGR are of significant difference compared with that of pressurized water reactor (PWR). As a result, the coordinated control design for MHTGR plant is quite different from the PWR. In this paper, the feedwater flow and control rods are solely used to regulate the steam temperature and nuclear power, respectively. Moreover, the so-called dynamic matrix control (DMC) then is utilized to realize the load match for MHTGR thermal power, where the setpoint of helium flow is regarded as manipulated variables and the modular thermal power is regarded as controlled variables to be optimized. The effectiveness of the proposed method is then tested and verified by a hardware-in-loop simulation through a commercial distributed control system (DCS).
PSA Analysis of Switch Port Disabled on DCS Layer 1
Authors: Luo Yang , Liu Deyi , Zou Shengjia , Cao Yong 
Published: October 2021

Abstract
This paper first introduces the background, purpose and implementation process of SDP, and then analyzes the SDP based on the disabled fault of the switch port of DCS layer 1 in nuclear power plant. This DCS failure caused 14 ports between the room level switch and the unit level switch on the first floor of DCS to be disabled, resulting in partial interruption of communication between the room level switch and the unit level switch, resulting in abnormal connection between KIC/BUP and the first floor, and all KIC control functions and BUP in order to prevent the fluctuation of working conditions from introducing new disturbance to the control system, the unit has been maintaining stable operation. During the whole event, the reactor was in a safe state, the reactor protection system was available, the three barriers were complete, and no radioactive material was released. The influence of DCS failure on core safety is analyzed by SDP, and the conclusion is given.
Migration Characteristics of Sub-Source Items in Breach Occurrences
Authors: Peng Xu , Ruifeng Tian 
Published: October 2021

Abstract
In the loss of coolant accident in a PWR nuclear power plant, the loss of coolant will cause the core to melt, resulting in the release of large amounts of radioactive fission products from the core[1]. Therefore, nuclear power plants must have effective protective measures to prevent the release of radioactive fission products to protect workers, the surrounding population, and the environment from radiological hazards. Most of the current scholars have conducted detailed studies using modular software such as MELCOR[2], but the three-dimensional distribution of the source material in the breach space cannot be visualized using this method.
In this paper, a typical pressurized water reactor nuclear power plant is taken as the research object. CFD method is used to analyze the condition of coolant loss accident, study the force of radioactive aerosols matter from different heights away from the wall under the accident condition, and consider the influence of ventilation system on the distribution and migration characteristics of the source matter in space.
The results show that the migration behavior of particles is mainly affected by gravity and drag force. For particles with small size, the combined action of gravity and thermophilic force causes them to deposit at the wall surface, while for particles with large size, gravity causes them to deposit at the wall surface. The particle deposition on the wall surface can be effectively reduced by increasing the flow velocity and decreasing the fluid temperature at the breach.
Application of Finite Difference Jacobian Based Newton-Krylov Method for Coupled Neutronics/Thermal Conduction Problems of Nuclear Reactor
Authors: Liu Baokun , Wu Yingjie , Zhang Han , Guo Jiong , Li Fu 
Published: October 2021

Abstract
The design and safety of nuclear reactor are strongly dependent on the neutronics and thermal conduction. The interaction between these coupled issues forms multi-physics nonlinear system, which has limited the economy of the nuclear power plant because of the complexity. And the accurate prediction of the reactor behavior still remains a challenge for the nuclear engineering. Traditionally, the neutronics and thermal conduction are solved separately by operator-splitting, which may be unstable and introduce significant errors. In recent years, fully coupled Newton-Krylov methods are prefer to be adopted due to these methods are more robust, accurate and efficient. However, each Newton-Krylov method has merits and demerits. In this study, finite difference Jacobian based Newton-Krylov (DJNK) and widely used Jacobian-free Newton-Krylov (JFNK) method are employed to solve the coupled neutronics/thermal conduction problems of a high-temperature gas-cooled reactor (HTGR). A steady state at thermal power of 250 MW and a transient state of supercritical accident are simulated. The results show that, DJNK is more efficient compared with JFNK, because DJNK performs better in preconditioning, which is the key to successful application of Newton-Krylov methods. And different preconditioning methods are utilized to reduce the number of Krylov iterations. A speedup ratio of DJNK over JFNK is observed to be 3. Synthesis results indicate that compared with JFNK, DJNK is a more potential method, contributing to a deeper understanding of the reactor behavior in a more efficient way.