bluuebunny/crossref_metadata_2025 huggingface dataset:
A Statistical Approach to Quantify Taylor Microscale for Turbulent Flow Surrogate Model
Authors: Ross Molly , Matulis John , Bindra Hitesh
Published: November 2022

Non-equilibrium statistical mechanics models can be used to construct reduced order models from the time-dynamics data such as numerical or physical fluid mechanics experiments. One of the well-established statistical projection methods is the Kramers-Moyal expansion (KM) method. The first two terms of the KM expansion result can be used to construct a non-linear Langevin equation, which can serve as the statistically-trained reduced-order model. This non-linear Langevin equation can be approximated to the Fokker-Planck equation, which is similar to Advection-Diffusion equation, thereby preserving some characteristics of fluctuations associated with fluid mechanics. The KM method captures continuous-time dynamics, however, any data obtained through measurement is discrete. In order to accurately capture the time dynamics of the discrete data, the method for calculating the KM coefficients must be carefully chosen and implemented. To better represent the solution from discrete data, the drift and diffusion coefficients can be calculated at multiple time scales and then extrapolated to a time scale of zero, assuming a linear correlation. One challenge in using this method is that the calculated KM coefficients are only accurate for time scales greater than the Taylor microscale. This means that the extrapolation must use only the KM coefficients calculated for time scales greater than the Taylor microscale, however, this value is not always provided from the data nor simple to calculate. This work presents a method of approximating the Taylor microscale from the data through the relationship between the Markov property and the Taylor microscale and implementing this method to find the extrapolated KM coefficients. The KM method implementing the Taylor microscale estimation was applied to existing DNS turbulent channel flow data to model a time series. This generated time series was then compared to the DNS data using a statistical analysis including probability density function, autocorrelation, and power spectral density.
Analysis of Double Heterogeneous Effect Under Double Dispersion of Fuel and Burnable Poison Particles
Authors: Yuan Anmin , Wu Hongchun , Li Yunzhao , Cao Liangzhi , Wang Cong , Jiang Chaofei , Yu Shengzhi , Wang Qianglong , Qiu Jinrong
Published: November 2022

Dispersed particle fuel is an advanced form of fuel. Due to the need for reactivity control, burnable poison particles and fuel particles are often dispersed in the matrix together. In order to prove the double dispersion mechanism of fuel and burnable poison particles with high calculation accuracy, a random dispersion model of burnable poison particles is constructed in this paper under the conditions of fuel homogenization and random dispersion of fuel particles. Compared with the random dispersion model, direct homogenization of fuel or burnable poison can cause large deviation in the calculation of system eigenvalues according to the MCX calculation. Under the same fuel phase volume, fuel enrichment and burnable poison loading, the maximum deviation can reach 1494pcm. The deviation increases with the increase of the loading mass of the burnable poison in the system, subject to fuel particle diameter of the order ∼100 μm. At a relatively high content of the burnable poison, the deviation decreases with the change of the content of the burnable poison. The calculation results show that the double heterogeneity effect of the system composed of fuel and burnable poison particles cannot be ignored, and the traditional homogenization calculation method must be corrected according to the actual situation. This paper is of valuable reference for the calculation and correction of dispersion fuel homogenization under the condition of strong absorber, and the program development and verification of using a new method to deal with the double heterogeneity effect.
Study on Maintenance Method of Design Basis in Operating Nuclear Power Plants
Authors: Wu Guangjun , Xiao Wei
Published: November 2022

How to maintain the design basis to maintain or improve the safety level of nuclear power plants in their lifetime, the major nuclear power operation countries led by the United States and France have made a beneficial attempt. There is no systematic research and Application on the maintenance of design basis in China. With the accumulation of operation time of domestic nuclear power plants, we will soon face the challenge of design basis and the resulting pressure of safe operation. Design basis maintenance concept was raised from international Nuclear Power Plants (NPPs) operating experience, due to the safety and operating performance being challenged by the missing or violating of design bases. Investigation in the international experience of design basis maintenance is carried out and basis and key performance factors are identified base on the research in World Association of Nuclear Operators (WANO) and International Atomic Energy Agency (IAEA) suggestion and criteria. Methodology framework is proposed based on the basis. The framework clarifies the goal and tasks of design basis maintenance, raises a systematic method that carrying out margin management during NPP daily operation and maintenance, controlling the modification and carrying out periodic safety review to maintain the design bases. It can provide a useful reference for how to carry out design basis maintenance in operational nuclear power plants.
Implementation Challenges and Proposed Suggestions for Nuclear Material Accountancy Management in Spent Fuel Reprocessing Plant
Authors: He Lixia , Bai Lei , Miao Qiang , Yang Qun
Published: November 2022

Nuclear material license holder should develop a nuclear material accounting management system and achieve closed balance accountancy. According to regulations requirements, nuclear material accountancy implementation is based on physical inventory and material measurement. Generally, the spent fuel reprocessing plant operates uninterrupted, its annual throughput of nuclear materials is huge, there are different kinds of measurement and analysis techniques configured combing the reprocessing process. So the closed balance approach of nuclear materials accountancy is a great challenge for reprocessing plant holders. In order to improve the accuracy and reliability of nuclear material, enhance the timeliness of abnormal detection, near-real time accounting prospect of spent fuel reprocessing plant was researched. In this paper, the crucial aspects affecting closed balance accountancy was discussed, the factors such as the head-end receiver-shipper inventory difference, the on-line process monitoring accuracy, the applicability of the international target value of nuclear material measurement uncertainty, and the nuclear material balance model of reprocessing plants. As summarized, proposed suggestions and solutions for nuclear material balance in spent fuel reprocessing plant was put forward on the end of the paper.
Experimental Research on Drag Characteristic of Air Flow Path for ACP-100 Passive Containment Air Cooling System
Authors: Yu Mingrui , Wang Hongliang , Liu Zhuo , Feng Yu , Han Xu , Yuan Yidan
Published: November 2022

Passive containment air cooling system (PAS) is an important special safety system for the small modular pressurized water reactor ACP-100. The heat is transferred to the environment by the natural draft air cooling under accident conditions. The flow loss of the air flow paths has an important influence on the heat transfer power of PAS. A one-sixth scale model of PAS air flow path was constructed which was a 90° wedge angle sector of the cylindrical symmetric passages. The pressure distributions at different sections of the model were measured, the drag coefficients were analyzed. The experimental results showed that the total drag coefficients of the four different air inlet paths were almost the same, and they showed a decreasing trend with the increase of Reynolds number. The sleeves had the largest drag coefficient, accounting for about 26%, followed by the air inlet, chimney outlet, the turning and the corridor. The other components had little effect on the pressure loss. Afterward, the aerodynamic modification was carried out, the modification for the turning could reduce the total drag coefficient by 4.5 %. The experimental results of this paper provide necessary data support for the ACP 100 engineering design.
Event Tree Analysis for Material Relocation on Core Catcher in a Sodium-Cooled Fast Reactor
Authors: Yamano Hidemasa , Kubo Shigenobu , Kan Taro , Shibata Akihiro , Hourcade Edouard , Dirat Jean François
Published: November 2022

For an in-vessel retention strategy, severe accident analysis needs to be input the condition of loading on a core catcher based on physically compatible assumptions. This study has conducted event tree analyses emphasizing key phenomena in the lower plenum based on available experimental knowledge as well as design measure effectiveness. Criticality assessment has also been performed for the event tree analysis. In this paper, the approach to event tree development and the scope of the event tree analysis are described with key points on the loading conditions on the core catcher. For the analytical conditions, two loading conditions on core catcher were given as bounding and conservative cases. For the important heading of the event tree, key important phenomena were included: strong back design, fuel-coolant interaction and quench in the sodium plenum design, jet attack, criticality and coolability of molten core materials on the core catcher. In this study, preliminary trial quantification was attempted using a probability ranking table which is based on engineering judgement. This event tree analysis has identified the dominant sequence, and clarified the effect of the loading condition on the core catcher and effectiveness of design measures. This study suggests that the criticality measure is very important for the core catcher study.
High Power Electron Beam Dump at DALS
Authors: Tao Kai , Huang Liming , Ding Yadong , Zou Zhenhai , E Dejun
Published: November 2022

The electron beam dump for Dalian Advanced Light Source (DALS) is designed to absorb 15 kW of electron beam power at beam energy up to 120 MeV. The DALS accelerator produces an electron beam with very small beam size of up to 100 μA average current. The resulting beam power, up to 15 kW at 120 MeV, and the very high beam power density, pose challenging problems for beam dump design.

High power dump with water cooled has been developed for DALS. In the dump, most of the beam power is finally absorbed in water and taken out from the dump.

The core of the high power electron beam dump is designed to be constructed from an aluminum alloy using a cylindrical geometry with fins arranged around to promote the heat transfer. The cooling water is forced by a cooling pump, to cool the core of the dump through the cooling channels.

The beam is stopped in the dump involving a high production of neutron and gamma radiation and activation of its surface. A shield has been designed to attenuate both the radiation produced during accelerator operation and the residual radiation.

Design details for the dump, including radiation shielding calculations, thermal analysis are presented.
Manufacturing Process of Fuel Salt Discharge Tank for 2Mwt Liquid Thorium Based Molten Salt Reactor
Authors: Wang Yan , Shi Hong
Published: November 2022

At present, the research on Thorium Based Molten Salt Reactor (TMSR) mainly focuses on molten salt-molten salt heat exchanger, molten salt-air heat exchanger, high-temperature molten salt pump, gas-liquid separator and Hastelloy structure and performance, and there is no research on the manufacturing process of fuel salt discharge tank. The fuel salt discharge tank mainly performs the function of fuel loading and unloading, that is, loading and unloading the fuel salt through the gas pressure. The difference gas pressure between the core and the fuel salt discharge tank make the fuel flow into the reactor pressure vessel or flow out of the reactor pressure vessel. The fuel salt discharge tank also has the function of storing the molten salt fuel and preventing fuel leakage when the core is overheated. It exists as a safety facility and has a very important impact on the normal and safety operation of the reactor system. Based on the requirements of standards and specifications, the main manufacturing process of fuel salt discharge tank were studied, technological processes such as tank material selection, head forming, barrel rolling, welding and pneumatic test were introduced. The fuel salt discharge tank of liquid TMSR meeting the requirements of standards and specifications was successfully developed.
Research and Application of 3d Visualization Digital Twin System for Ultrasonic Testing of Key Equipment in NPP
Authors: Yang Zheng-Hui , Zhang Yi , Wang Hao , Zhang Ya-Nan
Published: November 2022

The internal conditions and surface defects of key equipment in nuclear power plant (NPP) are mainly detected though non-destructive testing (NDT) technology, which is the key basis for maintenance and diagnosis of nuclear power main equipment. At present, there are many types and huge quantities of NDT data. Due to the lack of unified information management means, it depends more on personal skills and experience. On the other hand, discrete data files and 2D detection data can’t provide the staff with intuitive and stereoscopic visual effects, which affects the prediction and analysis of the staff to a certain extent. In view of the above problems, this paper deeply analyzes the characteristics of nuclear power NDT data, and proposes a 3D visualization method for ultrasonic defects of NPP reactor pressure vessel (RPV) based on digital twin. It also forms a 3D visualization digital twin system, which overlays and maps the 3D digital model of RPV with the reconstructed 3D defects, while providing the necessary 3D interaction. Through the research results of this paper, the digital and intelligent management methods of NDT data in NPP have been improved, and an outstanding 3D basis for prediction and analysis of NDT defects in NPP has been provided.
Verification and Validation of the Fuel Assembly Bypass Flow Analysis Code PEONY V1.0
Authors: Chen Changyi , Yan Jingwen , Xi Yanyan , Zhang Yuxiang , Li Weicai , Wu Change , Lv Qifeng , Ran Shuangye
Published: November 2022

Fuel assembly bypass flow is the coolant flow through the guide thimble and the instrument. The fuel assembly bypass flow should be limited to ensure that there is sufficient coolant in the core channel to remove the heat released by fuel fission, and the bypass flow should be large enough to avoid the boiling of the coolant in the guide thimble and the instrument tube. The calculation of the fuel assembly bypass flow is also important for the mechanical analysis of SCCA (Stationary Core Component Assembly) hold down system and for the analysis of control rod drop. The PEONY V1.0 code developed based on pressure drop balance between different flow channel was used for calculating the flow distribution and then the hydraulic force for SCCA mechanical design verification. The PEONY V1.0 code is verified by STEP fuel assembly bypass flow test. In order to obtain a more accurate measurement results, pressure drop test result is used for flow rate correction. The verification and validation of PEONY V1.0 shows that the code can provide an reasonably conservative calculation for engineering analysis.
Propulsion Control of Marine Nuclear Reactor by Hardware-in-the-Loop Simulation (HILS)
Authors: Zhang Ao , Wang Xiang
Published: November 2022

Compared with offline simulation based on software completely, Hardware-in-the-loop Simulation (HILS) is a simulation method based on software and hardware coupling, which can not only verify the principle of the model studied, but also feedback hardware response for comparison, shorten product test cycle, improve research and development efficiency and accuracy. Taking the marine nuclear reactor as the background, this paper modelled the main circuit of energy power and the part of propulsion power output, simulated the energy power module offline based on Modelica language, and coupled Arduino MCU with LEGO parts to simulate the hardware response of propulsion power output. This study used OpenModelica to construct the steady-state and transient simulation operation of the off-line simulation model, realized the interaction between the hardware signal input and the offline simulation model, processed the mechanical response of the simulation output and communicated with the external hardware. Finally, the feedback action response was compared with the key parameters in the existing literature. The results show that the simulation results of the model and the hardware response feedback results are consistent with the literature parameters, and based on the coupling of Modelica language with Arduino MCU can carry out the HILS for the marine nuclear power system in steady state and transient state.
Design and Comparison of Drainage Technology Scheme for Multi-Purpose Modular Small Reactor Science and Technology Demonstration Project
Authors: Yujia Mao , Liyuan Sheng , Ruobing Yang , Wei Bai , Jing Li
Published: November 2022

Compared with other nuclear power projects, the layout and site conditions of the Multipurpose Modular Small Reactor Technology Demonstration Project (hereinafter referred to as “Small Reactor”) have obvious particularities, which makes the design of the drainage process of the small reactor demonstration project has special considerations compared with that of the dual-reactor nuclear power plant. It mainly includes the following aspects: first, because of the single-stack arrangement, the plant water system cannot use the facilities of other units to drain water during shutdown for maintenance; second, the impact on the already operating Units 1 and 2 needs to be considered; third, consider the interaction with the Units 3 and 4; fourth, the unit cost of small reactors is relatively high, and the construction cost of drainage facilities needs to be controlled. In this paper, two drainage technology schemes are formed according to the particularity of the small reactor and the geological conditions of the site, considering the feasibility of the structure, the results of warm drainage, the construction method and the cost are analyzed respectively, the countermeasures are studied for possible problems and a reasonable arrangement method is finally recommended, which realizing the feasibility, reliability and economy of drainage engineering of Small Reactor.
Study on Performance of New Flexible Shielding Materials and First Demonstration Application in CPR Nuclear Power Plant
Authors: Li Yulong , Liu Feng , Liu Xiajie , Li Li , Kang Zheng , He Ye
Published: November 2022

Excessive gamma rays will be emitted when a nuclear power plant is under the overhauling, leading to a certain number of hotspots. To meet the shielding requirements of these hotspots, a new type of flexible shielding materials with a “sandwich structure” has been developed. New flexible shielding materials consist of two parts, one is silicone rubber with excellent physical and chemical properties by adding reasonable multi-component additives as matrix, and the other is lead powder as shielding body. This kind of material with lead powder dispersed uniformly in silicone rubber was prepared through mixing process. Based on this, the article has carried on the discussion to the performance of functional layer of the new flexible shielding materials. The problem that the process stability of ultrafine powder distributed uniformly in matrix with various densities among different regions has been solved as well. Furthermore, this kind of material has been adopted in CPR nuclear power plant as the first demonstration project. The site shielding effect has been recognized by WANO (world association of nuclear operators). As a result of this research, it is possible that the scheme of radiation shield for a short or long time can be make quickly when nuclear power plant running or overhauling. Meanwhile, this study is of great significance for the popularization and application of shielding materials.
Three-Factor Winding Wire Insulation Aging Evaluation Based on Design of Experiment
Authors: Zhou Daofan , Yan He , Diao Xingzhong
Published: November 2022

Winding wire, as an important part of electrical equipment such as motor coils, is vital to the stable operation of reactor electrical equipment. The insulation aging of winding wire in reactor environment is affected by multiple factors such as temperature, voltage and irradiation. Due to the complex aging mechanism under the effect of multi-factor coupling, it has become a research hotspot to explore the influence of single factor and multi-factor coupling on the insulation lifespan of winding wire, and give the insulation lifespan of winding wire under multiple factors. In this paper, design of experiment (DOE) method was used to test design and results analysis, and the corona resistance tests of winding wire were carried out to get the corona resistance lifespans, which were regarded as the insulation indicator of winding wire under multiples factors. Full factorial experimental design was planned by changing the temperature, irradiation dose, and voltage amplitude at two level. The corona resistance lifespans of the samples were obtained, and the data were processed by the two-parameter Weibull distribution. Finally, the effects of single-factor and multi-factor coupling on insulation lifespan were discussed, and a model for evaluating the insulation lifespan of winding wire under the combined effects of three parameters was developed.
Preliminary Prediction of Young’s Modulus and Thermal Conductivity of Porous U-50 wt.% Zr Alloy: A Molecular Dynamic and Semi-Empirical Study
Authors: Cai Mengke , Cong Tenglong , Gu Hanyang
Published: November 2022

In the upgrades and innovations of nuclear fuel material, the U-50 wt.% Zr alloy is regarded as one of the most promising metallic fuel materials, due to its excellent thermal response, acceptable irradiation performance, and ease of fuel recycling. Under in-pile irradiation, large temperature gradients and dimensional changes contribute to complicated fuel thermal-mechanical behaviors, including the pore effect induced by fission gas production. However, the deficiency of the physical parameters of the porous U-50 wt.% Zr alloy makes it hardly possible to conduct fuel performance prediction under high irradiation conditions. To obtain Young’s modulus and thermal conductivity of porous U-50 wt.% Zr alloy, the molecular dynamics (MD) code LAMMPS and the modified embedded atom method (MEAM) potential for binary U-Zr system were incorporated. In this study, three-dimensional elastic constants were calculated by engineering strain loading method at different ambient temperatures and porosities, and the effective Young’s modulus was computed via Voigt averaging scheme. The phonon thermal conductivity was simulated with the Non-Equilibrium Molecular Dynamics (NEMD) method, and the electron thermal conductivity was predicted by semi-empirical correlations and existing density functional theory (DFT) results. The parallel model, series model and effective medium theory (EMT) were adopted to consider the mixture and pores effect. Finally, porosity factors were proposed to establish new semi-empirical correlations, which could give a preliminary prediction of Young’s modulus and thermal conductivity for porous U-50 wt.% Zr alloy.
Assessment and Control Value Formulation on Radiological Consequence of On-Site Worker due to Radioactive Component Failure for Pressurized Water Reactor Plant
Authors: Jing Zhou , Wen Wang Ran , Quan Gong , Wei Feng Lv
Published: November 2022

All Structures, Systems, and Components (SSCs) in a nuclear power plant need to be classified according to their importance to safety to protect people and the environment. The Specific Safety Guide, No. SSG-30, issued by IAEA, guides how to meet the requirements for identifying safety functions and classification of SSCs and ensuring their appropriate quality and reliability. However, the relevant guidance has not specified the perspectives of low radiological consequences on on-site workers. To refine the SSCs classification and complement the IAEA guidance, a proposed limiting value and corresponding assessment method for low radiological consequence caused by SSCs failure from the on-site workers’ perspective based on international good practices are developed, which include the main process of on-site radiological consequence evaluation, the calculation method for different exposure pathways and the assessment assumptions. Also, typical examples of SSCs failure are also assessed to validate the proposed method’s feasibility. The assessment results show that the activity concentration of failure SSCs, the workers’ locations, the exposure duration of workers, and the local dose rate of each location are the main factors for on-site radiological consequence evaluation, and the proposed method in this paper can assist in the safety classification of SSCs in the third generation of pressurized water reactor nuclear power plants.
Data Reconstruction of Faulty Sensors for the Nuclear Power Plants Control System: A Strong Tracking Filter Approach
Authors: Zhou Hongkuan , Zheng Wei , Tao Mo , Guo Xiaojie , Huang Chonghai , Chai Wenting , Chen Kai , Chen Zhaoxu
Published: November 2022

As an indispensable component of the control system, the sensor usually works in high temperature, high pressure, strong vibration and other failure-prone environment. The complex, changeable, and high intensity working environment makes the sensors work unstable, which eventually leads to the sensor fault. Once the sensor fault occurs, the output signal will deviate from the normal value, which seriously decreases the accuracy, stability and reliability of the control system. To solve this problem, the researchers tried to locate the faulty sensors and recover the missing data by the expertise, which results in low accuracy and long-time consuming. In this paper, a novel data reconstruction method for padding the missing data of the faulty sensor is proposed. Based on the extensive redundancy and complementarity of the information obtained from various types of sensors, the data reconstruction of the target sensor is realized by combining the thermal system model with the strong tracking filter algorithm. Simulation data of the single pressure condenser system show that, under different sensor fault modes, the reconstruction accuracy deviation of the signal reconstructed by the algorithm is less than 5% compared with the normal signal. The application of the proposed algorithm can significantly improve the safety and reliability of the control system.
Source Localization of Accidental Radionuclide Release Using a Data Assimilation Method Based on Forward and Backward Dispersion Simulations
Authors: Xu Yuhan , Fang Sheng , Dong Xinwen , Zhuang Shuhan
Published: November 2022

Source localization of accidental radionuclide release is crucial to nuclear emergency and decision making, especially when no reliable information about possible release regions can be previously obtained. To rapidly and accurately determine the source location with a limited number of measurements, this paper proposes a two-step data assimilation method based on forward and backward dispersion simulations in the RIMPUFF model. In the first step, meteorological variables such as wind speed are reversed and the measurements are set as sources for several backward RIMPUFF runs. By identifying the overlap of backward plumes originating from different measurement sites, potential regions discretized into grids can be roughly determined. In the second step, the grids selected by the first step are set as sources for forward RIMPUFF runs with unit release rate and the source location will be refined by minimizing a correlation-based function of simulated and measured data. The method is verified by SCK-CEN 41Ar field experiment. Using gamma dose rate data from all measurement sites, source location is retrieved with errors of 22.36m and 30m respectively on the two days of the field experiment. Compared with the direct correlation-based method, the proposed method achieves satisfactory solutions in an obviously shorter time. In particular, there is no need for predetermining the source release rate throughout the process, indicating that the method can conveniently combine with other source term inversion approaches whose source location is viewed as a known quantity. Therefore, the presented two-step data assimilation method here is potentially a useful tool with high accuracy and efficiency for further integration in nuclear emergency response system.
Validation of NUIT for Nuclides Composition Calculation With PWR Spent Fuel Isotropic Measurement
Authors: Tan Kai , Li Ruihan , Li Jian , Liang Jingang , She Ding
Published: November 2022

The NUIT (Nuclide Inventory Tool) is a reactor core nuclide inventory calculation code, which has been developed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. NUIT is able to compute instantaneous and integral values of various physical parameters including isotope composition, decay heat, particle spectrum, and radioactivity in fuel burnup, material activation, and decay calculations. The validation of the NUIT is an important process to ensure the correctness and applicability of the newly developed code. Presently, there is a lack of validation of the nuclide density calculation function of NUIT. Hence, this work is to validate this function of the NUIT code based on the PWR spent fuel isotopic compositions benchmark issued by the Oak Ridge National Laboratory in the SCALE program. In this work, the 103 spent fuel samples from 9 PWR plants are used to validate the nuclide composition calculation function of the NUIT, in which the initial enrichment of U-235 ranges from 2.453% to 4.657%, and the burnup ranges from 7.8 GWd/MTU to 60.7 GWd/MTU. The results show that NUIT’s nuclide density calculation results are in good agreement with the experimental values in general and the average error is about 10% for all nuclide density calculation cases. The typical nuclides, such as U-235, U-238, and Pu-239, etc., have an average calculation error of less than 5%, and most of the average errors of other actinide nuclides and fission products, such as Am, Cm, and Pu, etc., are within 10%∼40%, which meets the requirements of the radioactive source term analysis for the reactor core. Meanwhile, the comparison results show that the calculation accuracy of NUIT is comparable to other radioactive source term programs such as SCALE and OpenMC. The validation results further demonstrate NUIT’s capabilities in nuclide composition calculations.
RISMC Analysis Method Used in Operation and Maintenance of Nuclear Power Plant
Authors: Wang Zhao , Liu Weidong , Liu Leilei , Wang Yongxiang , Xu Liangwang , Pei Liang , Li Qiongzhe , Guo Jianbing
Published: November 2022

Risk informed safety margin characteristics (RISMC) analysis Method, is a safety analysis method which combines probability safety analysis and determinism safety analysis of nuclear power plant. General RISMC analysis platform “CARS” (Coupling Platform for Advanced Risk Simulation) can provide parameter distribution and Monte Carlo sampling function, and interactive coupling function with data of the thermal hydraulic safety analysis system program. The paper uses CARS and thermal hydraulic safety analysis system program to analyzes the technical requirements of personnel training and engineering modification of nuclear power plant, by means of accident scenario sampling, physical mechanism calculation and scenario consequence statistics. In the analysis process, various important factors which may affect the event process and consequences are comprehensively considered through the analysis of nuclear power plant operation rules, equipment reliability and the uncertainty distribution characteristics of key factors. And consequences of event scenario development, nuclear power plant response and the time course are analyzed through establish a realistic physical model by using thermal hydraulic safety analysis system program. The calculation results give the target level of personnel training and the effectiveness of engineering modification. The results show that RISMC analysis Method is an effective safty analysis and decision making method for operation and maintenance of nuclear power plants.
Reliability Analysis of Active Reactor Pit Flooding Function of EHR DFT System of a NPP Using an Enhanced Component-Connection Method
Authors: Guo Dingqing , Wang Shaoxuan , Ge Daochuan , Tong Lili , Cao Xuewu
Published: November 2022

Redundancy design is often used to enhance the reliability of complex systems which often run as safety-critical systems of nuclear power plant (NPP). Due to the redundancy design and management strategies, safety-critical systems of a NPP often have complex sequential failure behaviors. For a safety-critical system of a NPP with sequential failure behaviors, it is very important to evaluate its reliability for a given mission time. In this paper, the active reactor pit flooding function of EHR (containment heat removal system) of a nuclear power plant (NPP) is modeled by dynamic fault tree (DFT) and analyzed by BDD. To build the BDD, an enhanced component-connection-based method is proposed. The failure probability of the active reactor pit flooding function is analyzed. Besides, the FV, RAW and RRW importance of the EHR equipment are also evaluated. The results show the cumulative failure probability at mission time 24 hour is 1.81E−05, which is very low. The cumulative failure probability of train A (train B) is 4.25E−03. It can be concluded that the active reactor pit flooding function of EHR has a very high reliability given a mission time 24h.
Effect of Carbon Content on the Resistance to Intergranular Corrosion of Alloy 800H
Authors: Zhong Qiang , Wang Man , Sun Wen-Qiang , Wu Jun-Jie , Qian Jiong , Ning Sen-Sen
Published: November 2022

The effect of carbon content on the intergranular corrosion performance of Alloy 800H was studied by thermodynamic software simulation calculation combined with double-loop electrochemical potentiodynamic reactivation (DL-EPR), sulfuric acid-copper sulfate test, and microscopic morphology observation (optical microscope, scanning electron microscope). The results show that the main precipitation phases of Alloy 800H are α-Cr phase, M(C,N), M23C6, η phase and σ phase. At 500∼720°C, the higher the carbon content and the more M23C6 phase, the higher the susceptibility to intergranular corrosion. In the solid solution state, the Alloy 800H has low sensitivity to intergranular corrosion, and the carbon content has no obvious effect on its resistance to intergranular corrosion. In the sensitized state, carbides are continuously precipitated at the grain boundaries, and the intergranular corrosion susceptibility of the alloy increases rapidly.
The Applicability Analysis of Fusion Reactors and China Nuclear Safety Code HAF102
Authors: Qian Yijie , Li Chong
Published: November 2022

With the development of ITER project, China is preparing to build China Fusion Engineering Experimental Reactor (CFETR), at present; there is no domestic nuclear fusion regulation framework. The nuclear safety regulation framework based on the fission technology is relatively solid and referable, which can provide reference for the establishment of fusion regulation framework. HAF102 “Safety Code on Nuclear Power Plant Design” is one of the most important regulations in Chinese nuclear safety framework. HAF102 specifies the nuclear safety requirements and provides guidance for the design of fission reactors in terms of defense in depth, safety functions and safety classification, these design principles are also crucial for fusion reactors. Therefore, this paper analyzes the general applicability of fusion reactors and HAF102-2016, identifies the non-conformance terms and categories the terms into generally applicable and partially applicable. The application analysis aims to identify the important safety issues that should be considered in the early stage of the development of fusion reactors, in addition, the inapplicable provisions of HAF102 are identified based on the characteristics of fusion reactors. The analysis lays a preliminary foundation for the establishment of a fusion regulation framework on safety issues in China, and on the other hand, it provides recommendations for the design principles of fusion reactors.
Study for the Construction of Gas Leak Detection System in Nuclear Power Plants Using Ultrasound
Authors: Kai Seigo , Moriya Takeshi , Shoji Naruki , Endo Gen , Takahashi Hideharu , Kikura Hiroshige
Published: November 2022

On March 11, 2011, a serious accident occurred at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company Holdings, Inc. due to the Great East Japan Earthquake and tsunami. Currently, nitrogen gas is used to fill the reactor containment vessel, pressure vessel, and plant piping to prevent a hydrogen explosion. Gas leak detection technology is needed to maintain and control the integrity of nitrogen gas containment and to ensure plant safety, as well as to identify the leak location in the event of a leak and understand the nature of the leak before plugging the leak location. However, because of the high radiation environment in the plant, it is not practical for a person to enter the plant to detect gas leakage. Therefore, there is a need for a leak detection method for gases that can be used in a radiation environment, and that is compact, measurable, and can be loaded by robots. In this study, we focused on acoustic gas leak detection as a leak detection technique. The characteristics of ultrasound generated by gas leakage were investigated for the development of gas leak detection technology and source localization system in nuclear reactors.
Comparative Study on Creep-Fatigue Damage Assessment Between R5 Procedures and a Full Inelastic Analysis
Authors: Hurst Antony M. , Chang Louis C. W. , Cho Nak-Kyun
Published: November 2022

This research aims to investigate the differences in structural integrity assessment results between R5 Volume 2/3 procedures and a detailed inelastic analysis. The membrane sidewall/roof bimetallic stub tube subjected to creep-fatigue loading is selected as an example. Based on a linear elastic analysis, the maximum equivalent stress is found to be at the weldment located at the top of the stub tube. The R5 Volume 2/3 procedures enhance the start of a dwell stress and inelastic strain ranges by applying a weld strain enhancement factor and a stress concentration factor, and the assessment predicts a total creep-fatigue damage of 54% for a 45 year life. As a comparison, a detailed inelastic analysis has been conducted for the same bi-metallic stub tube using an incremental, step-by-step analysis. Temperature dependent material properties are used to simulate stress-strain response and a user subroutine is employed for creep deformation. Assessment results for the detailed inelastic analysis confirm the weldment is similarly affected as indicated by the elastic analysis results but the comparable total predicted creep-fatigue damage is less than 35%. Cold eye review of both assessment results concluded that the significant deviation is attributed to the additional factors applied for the assessment of weldments, including weldment geometry, flaw effects and weld micro-cracking, as provided in the R5 Volume 2/3 procedures.
Manufacturing Technology of Mn-Ni-Mo Alloy Steel Heavy Forgings for HPR1000 Nuclear Power Plant
Authors: Zhifeng Zhang , Xiang Li , Xiaofang Wang
Published: November 2022

HPR1000 nuclear power technology is the mainstream technology of commercial pressurized water nuclear power in China, which is adopted by most nuclear power plants under construction or planned. Mn-Ni-Mo alloy steel heavy forgings are widely used in the equipment of pressurized water nuclear reactor. In order to achieve the aim of safe operation for at least 60 years, the technical specifications of Mn-Ni-Mo alloy steel heavy forgings for HPR1000 are high and strict. Due to the high technology and difficult manufacturing, only a few factories have the manufacturing capacity of forgings with stable quality worldwide, which can’t meet the continuous growth demand of Chinese market. Therefore, whether Mn-Ni-Mo alloy steel heavy forgings can be delivered on time has become an important factor restricting the construction cycle of HPR1000 nuclear power plants.

In order to meet the market demand, Shanghai Electric SHMP Casting & Forging Co., Ltd (SCF) has developed the manufacturing technology of heavy forgings for HPR1000. According to the technical specifications of HPR1000, SCF has carried out systematic research in smelting, forging and heat treatment technologies. By 2020, SCF has realized the mass production of heavy forgings for HPR1000 nuclear power plant.

This paper reports new technologies obtained by SCF in the development of Mn-Ni-Mo alloy steel heavy forgings for HPR1000 nuclear power plant.
Sensitivity Analysis on Key Parameters of Severe Accident Source Term of PWR
Authors: Ou Pingwen , Chen Yongzheng , He Dongyu , Chen Peng
Published: November 2022

The complicated severe accident phenomena in typical Pressurized Water Reactor (PWR) Generation III may have a strong influence on source term release into environment and radiological consequence. The study on sensitivity analysis is beneficial to the identification of important factors in severe accident source term analysis and the quantification of their impact.

ASTEC, the integral code of severe accident analysis developed by IRSN, is used to analyze the sensitivity of key parameters of severe accident source term for typical PWR Generation III, with the simulation of safety systems and source term phenomena, in the representative sequence with fast accident progression, Large Break Loss of Coolant Accident (LBLOCA). With the consideration of the design features of typical PWR Generation III and research status of severe accident source term, the key parameters for sensitivity analysis are identified and selected based on the whole process of radionuclides release, including gaseous iodine mass release fraction from primary circuit to containment, silver iodide reaction, dose rate and pH value in sump, washing effect, etc.

The sensitivity is quantified by iodine release mass to containment, which is one of the most dangerous radionuclides due to its threat to environment and human thyroid after inhalation and ingestion. The gaseous iodine mass release fraction from primary circuit to containment, silver iodide reaction and washing effect are presented in results as the major contributors to the variation of severe accident source term evaluation.
Preliminary Study on Three-Dimensional Characteristics of Modular High Temperature Gas-Cooled Reactor
Authors: Zhang Haojie , Zheng Yanhua , She Ding , Shi Lei
Published: November 2022

The two-dimensional (2D) system analysis codes are usually used to design and analyze the modular pebble-bed High Temperature Gas-cooled Reactors (HTR). However, in some cases, three-dimensional (3D) characteristics, for example the 3D power distribution and temperature distribution due to a part of control rods movement, also need to be concerned. In this paper, based on the 200 MWe Pebble-bed Modular High Temperature gas-cooled Reactor (HTR-PM) design, a 3D system analysis code was used to analyze some three-dimensional characteristics in steady state, as well as the transient process caused by the asymmetric movement of control rods in the side reflectors.

Under normal operations of HTR-PM, the control rods are inserted at certain heights and the primary coolant enters the Reactor Pressure Vessel (RPV) from the outer pipe of the coaxial hot gas duct. As a result, the fuel temperature has a relative circumferential deviation of about 1.64% on the outside of the pebble bed, while a very small deviation at the center of the bed. The temperature distribution of RPV is significantly affected by the position of the helium inlet as well. The analysis results showed that, during the transient process caused by the withdrawal of two control rods, the relative circumferential deviation of the fuel particle temperature on the outside of the pebble bed would increase to around 8.47%. However, because of heat conduction in relatively long distance, the movement of the control rods has very little effect on the circumferential temperature deviation of the fuel particles at the center of the bed. These 3D characteristics of HTR under different conditions demonstrate the good applicability of 2D and 3D system codes and will provide support for more accurate safety analyses in the future.
Transient Thermal-Hydraulic Analysis for Thermal Load Fluctuation Test Using HTTR
Authors: Aoki Takeshi , Sato Hiroyuki
Published: November 2022

High temperature gas-cooled reactor (HTGR) has a potential to produce competitive and large amount of carbon-free hydrogen because of its capability to supply high temperature heat of nuclear energy. It is required to establish the control method and system for the HTGR hydrogen production system to maintain its normal operation against the abnormality in the hydrogen production facility. Performances of the control method and system such as readiness and stability have to be demonstrated by transient thermal-hydraulic analysis with a code developed based on RELAP5/MOD3 code.

The reactor response against the disturbance in the reactor inlet coolant temperature is revealed in the HTGR hydrogen production system using the control systems. The analytical results showed that the reactor outlet coolant control system enabled to control the variation of the reactor outlet coolant temperature within 4°C against 30°C of large disturbance in the reactor inlet coolant temperature. A variation of the reactor outlet coolant temperature has to be less than 5°C in the simulation to maintain its normal operation in the HTGR hydrogen production system. Thus, the effectiveness of the control method was confirmed. The simulations code and results will be validated through a HTTR thermal load fluctuation test.
Reliability Analysis of Phased Mission Systems Using GO-FLOW Methodology
Authors: He Zhanyu , Yang Jun , Matsuoka Takeshi , Yang Ming
Published: November 2022

A Phased Mission System (PMS) is defined as a system subject to multiple, consecutive and non-overlapping operation phases of tasks during its mission, which is commonly found in complex technological and engineering practices such as aerospace, nuclear and military operations, etc. Over the past few decades, extensive efforts have been devoted to advance the theories, methods, techniques and tools used for reliability analysis of PMS. The general methods include fault tree based combinatorial approaches and Markov chains. In this paper, a success-oriented GO-FLOW method with a new exact algorithm is presented for reliability analysis of multi-phase mission systems. The feasibility and correctness of the extended GO-FLOW method (GFA) for PMS analysis are proved by comparing the results with fault tree analysis based on Sum of Disjoint Products (SDP) generation algorithm for two example case studies. The comparison results show that the GO-FLOW method can be effectively applied for reliability analysis of PMS in a very compact way. Consistent results can be obtained using extended GO-FLOW algorithm when compared to the fault tree analysis.