ASME 2018 Nuclear Forum最新文献

Research and Practice of Nuclear Equipment Procurement and Supply Management Based on “Internet+” Mode 基于“互联网+”模式的核设备采购供应管理研究与实践

ASME 2018 Nuclear Forum Pub Date : 2018-06-24 DOI: 10.1115/NUCLRF2018-7490

Hui Zhou, L. Ding, Xu Shi, Zhongkui Li

{"title":"Research and Practice of Nuclear Equipment Procurement and Supply Management Based on “Internet+” Mode","authors":"Hui Zhou, L. Ding, Xu Shi, Zhongkui Li","doi":"10.1115/NUCLRF2018-7490","DOIUrl":"https://doi.org/10.1115/NUCLRF2018-7490","url":null,"abstract":"The procurement and supply management of nuclear power equipment are carried out offline in tradition, the mode is mono and inflexible, the organizational process assets is poorly accumulated, the implementation process is not transparent, the tracking control effect is poor, the efficiency is relatively low. Adopting the information technology has acquired certain effect in improving organizational process assets accumulation although, the information is fragmented, efficiency improvement is not obvious. Study on “Internet + Procurement” was carried out since the year of 2015, compared with the traditional model, all the business and supplier information of procurement management of nuclear power equipment were interconnected, the whole process of procurement and supply management are through the internet, it is a innovation of the procurement and supply management mode using information technology. a wealth of data was accumulated for further analysis, an interconnected network was built between buyer and supply, significant change was brought out to the field of nuclear power equipment procurement, the role of this synergistic effect, contributes to the shortening of the construction period of nuclear power projects.","PeriodicalId":315206,"journal":{"name":"ASME 2018 Nuclear Forum","volume":"18 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-06-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131870261","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Thermal Analysis Safety Margins Using ABAQUS for the MP-2 Experiment in the Advanced Test Reactor 先进试验堆MP-2实验用ABAQUS热安全裕度分析

ASME 2018 Nuclear Forum Pub Date : 2018-06-24 DOI: 10.1115/NUCLRF2018-7600

G. Hawkes, D. Crawford, G. Housley

{"title":"Thermal Analysis Safety Margins Using ABAQUS for the MP-2 Experiment in the Advanced Test Reactor","authors":"G. Hawkes, D. Crawford, G. Housley","doi":"10.1115/NUCLRF2018-7600","DOIUrl":"https://doi.org/10.1115/NUCLRF2018-7600","url":null,"abstract":"The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR). MP-2 is considered a non-instrumented drop-in test where small aluminum-clad fuel plate samples are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates were analyzed. A thermal analysis has been performed on the MP-2 experiment to be irradiated in the ATR at Idaho National Laboratory (INL). A new technique for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) using the commercial finite element and heat transfer code ABAQUS is demonstrated. This new technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node. Pressure drop data is fed into the calculations in order to geometrically calculate the water saturation temperature.\u0000 Results from the DNBR and FIR calculations are displayed with the ABAQUS post processor named Viewer. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.","PeriodicalId":315206,"journal":{"name":"ASME 2018 Nuclear Forum","volume":"31 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-06-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115574859","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

The Features of Closed Brayton Cycle and Sub-Critical Combined Cycles Coupled With (Very) High Temperature Gas-Cooled Reactor 封闭布雷顿循环和亚临界联合循环与(极)高温气冷堆耦合的特点

ASME 2018 Nuclear Forum Pub Date : 2018-06-24 DOI: 10.1115/NUCLRF2018-7384

Xinhe Qu, Xiaoyong Yang, Gang Zhao, Jie Wang

{"title":"The Features of Closed Brayton Cycle and Sub-Critical Combined Cycles Coupled With (Very) High Temperature Gas-Cooled Reactor","authors":"Xinhe Qu, Xiaoyong Yang, Gang Zhao, Jie Wang","doi":"10.1115/NUCLRF2018-7384","DOIUrl":"https://doi.org/10.1115/NUCLRF2018-7384","url":null,"abstract":"High Temperature Gas-cooled Reactor (HTR) and Very High Temperature Gas-cooled Reactor (VHTR), are the most promising and achievable fourth-generation nuclear reactor for its inherent safety. In this paper, the performance of Closed Brayton Cycle (CBC) and two sub-critical combined cycles were investigated and compared. The CBC is a recuperated and inter-cooling closed Brayton cycle. Two combined cycles include the sub-critical Rankine cycle without steam reheating (Simple Combined Cycle, SCC) and a sub-critical reheated Rankine cycle (Reheated Combined Cycle, RCC). The topping cycles of SCC and RCC are both a simple Brayton cycle, and connect with the bottoming cycles by a sub-critical heat recovery steam generator (HRSG). Physical and mathematical models of three different thermodynamic cycles were established. Within the temperature range of the HTR and VHTR, the effects and mechanism of key parameters, such as reactor outlet temperature, steam temperature and pressure, on features of three different cycles were investigated. The results showed the elevated reactor outlet temperature could obviously enhance efficiency of three cycles. The results showed that RCC had the highest efficiency while SCC had the lowest efficiency, and the efficiency of CBC is slightly lower than that of RCC. The results could be helpful to understand and develop the power conversion system coupled with (V)HTR in the future.","PeriodicalId":315206,"journal":{"name":"ASME 2018 Nuclear Forum","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-06-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128220026","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Analysis of Loss-of-Flow Accidents in Pre-Cooler and Inter-Cooler of HTR-10GT HTR-10GT预冷器和中间冷器失流事故分析

ASME 2018 Nuclear Forum Pub Date : 2018-06-24 DOI: 10.1115/NUCLRF2018-7381

Xiaoyong Yang, Xiao Li, Jie Wang, Youjie Zhang

{"title":"Analysis of Loss-of-Flow Accidents in Pre-Cooler and Inter-Cooler of HTR-10GT","authors":"Xiaoyong Yang, Xiao Li, Jie Wang, Youjie Zhang","doi":"10.1115/NUCLRF2018-7381","DOIUrl":"https://doi.org/10.1115/NUCLRF2018-7381","url":null,"abstract":"Closed Brayton cycle (CBC) coupled with High Temperature Gas-cooled Reactor (HTGR) has potential application due to its compact configuration, high power generation efficiency and inherent safety. It is also one of the major power conversion methods for Generation IV advanced nuclear power systems. The typical CBC has several helium-water heat exchangers, including pre-cooler and inter-cooler. These helium-water heat exchangers have important influence on the performance of power conversion system, especially in loss-of-flow accidents (LOFAs). A system model including the reactor and the energy conversion system was established in this paper. The 10MW High Temperature Gas-cooled reactor-test Module helium Gas Turbine (HTR-10GT) was taken as the example to show the consequences of LOFAs. The results showed that LOFAs led to the rising of water temperature out of heat exchangers. With the reduction of water flow rate, the maximum water temperature would increase sharply, and the water temperature in pre-cooler was higher than that in inter-cooler. At low water flow rate, the water temperature would exceed the boiling point. LOFAs also made the rising of helium temperature. It had impacts on the performance of helium compressors. The elevated inlet temperature of helium compressors changed the corrected speed and corrected flow rate, therefore caused the deterioration of compressor’s performance. Furthermore, the LOFAs caused the reactor inlet temperature increasing. In low water flow rate, it would make the reactor inlet temperature beyond the temperature limitation of reactor pressure vessel and influence the safety of reactor. And the LOFAs also reduced the output work of cycle. This paper provides insights of features of CBC in LOFAs and will be helpful to the design and safety operation of closed Brayton cycle coupled with HTGR.","PeriodicalId":315206,"journal":{"name":"ASME 2018 Nuclear Forum","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-06-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125715244","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Additive Manufacturing of Spent Fuel Storage Rack Model by Selective Laser Melting 选择性激光熔化增材制造乏燃料贮存架模型

ASME 2018 Nuclear Forum Pub Date : 2018-06-24 DOI: 10.1115/NUCLRF2018-7409

Xiaoming He, Ziqiang Zhu, Changlei Shao, Ran Huang

引用次数: 1