International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第10页

A movable boundary model for helical coiled once-through steam generator using preconditioned JFNK method 用预调节JFNK方法建立螺旋盘式蒸汽发生器的可动边界模型

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-03-01 DOI: 10.1016/j.jandt.2021.12.001

Yingjie Wu, Baokun Liu, Han Zhang, Jiong Guo, Fu Li

{"title":"A movable boundary model for helical coiled once-through steam generator using preconditioned JFNK method","authors":"Yingjie Wu, Baokun Liu, Han Zhang, Jiong Guo, Fu Li","doi":"10.1016/j.jandt.2021.12.001","DOIUrl":"10.1016/j.jandt.2021.12.001","url":null,"abstract":"<div><p>Helical coiled once-through steam generator (H-OTSG) is the largest and the most complicated heat exchanger in the high-temperature gas-cooled reactor (HTGR), whose performance plays an important role in the nuclear power plant design and dynamic behavior analysis. A mathematical model of H-OTSG is developed based on the movable boundary method and solved by the advanced fully implicit Jacobian-free Newton-Krylov algorithm. The physical-based preconditioners are proposed and analyzed in this work, which achieves high-performance in solving the nonlinear steam generator system. In order to evaluate the performance of the movable boundary model, the fixed fine mesh model is also considered. The comparison is made between these two models from the perspective of the computational efficiency and accuracy. The simulation results are validated by the HTR-10 steam generator design data, showing that both models agree well with the design data, but have their own features. The fixed fine mesh model has its advantages in the accuracy, while the movable boundary model can realize the high computational efficiency.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 1","pages":"Pages 1-8"},"PeriodicalIF":0.0,"publicationDate":"2022-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605021000338/pdfft?md5=5a775afe1ebbf35a2eebc3a8c0f3c0f6&pid=1-s2.0-S2468605021000338-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75827991","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Study on startup characteristics of prototype once-through steam generator for China fast reactor 中国快堆直流蒸汽发生器样机启动特性研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-03-01 DOI: 10.1016/j.jandt.2022.02.003

Xin Min, Dalin Zhang, Rongshuan Xu, Shibao Wang, Youchun Chen, Chenglong Wang, Wenxi Tian, Suizheng Qiu, G.H. Su

{"title":"Study on startup characteristics of prototype once-through steam generator for China fast reactor","authors":"Xin Min, Dalin Zhang, Rongshuan Xu, Shibao Wang, Youchun Chen, Chenglong Wang, Wenxi Tian, Suizheng Qiu, G.H. Su","doi":"10.1016/j.jandt.2022.02.003","DOIUrl":"10.1016/j.jandt.2022.02.003","url":null,"abstract":"<div><p>Once-through steam generator (OTSG) is a vital heat transfer equipment of China Fast Reactor-600 (CFR-600), and its startup characteristics are of great importance to the safe operation of nuclear power plant. In this paper, first of all, the Thermal-hydraulic analysis Code Of Sodium heated once-through Steam generator (TCOSS) has been further optimized, and the code was introduced in the text. Secondly, the Steam Generator Test Facility (SGTF) and India Commercial Fast Breeder Reactor (CFBR) were selected for steady-state verification, and the feedwater flow rate increase condition of the prototype OTSG experiment was selected for transient verification. All the verifications were in good compliance, which proved the reliability of the TCOSS code for subsequent calculation of startup conditions. Finally, taking the prototype OTSG of CFR-600 as the research object, and the transient characteristics of cold startup and hot startup have been predicted and analyzed. The results show that both processes of cold startup and hot startup for prototype OTSG can be realized stably. The simulation results of startup processes showed a good agreement with expectations, which can provide support for the development of CFR-600 steam generator.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 1","pages":"Pages 26-35"},"PeriodicalIF":0.0,"publicationDate":"2022-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000035/pdfft?md5=3ff8c65d4e5a7b2ff849fda68f4cb3f4&pid=1-s2.0-S2468605022000035-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76618682","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 4

Corrigendum to “Heat-transfer characteristics of screw tube in one-side high heat load condition for fusion reactor divertor application”[Int. J. Adv. Nucl. Reactor. Des. Technol. 3 (2021) 213–225] “单侧高热负荷条件下用于聚变反应堆导流器的螺旋管传热特性”的勘误[j]。J. Adv.核反应堆。工业技术，3 (2021)213-225 [j]

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-03-01 DOI: 10.1016/j.jandt.2022.02.002

Ji Hwan Lim , Su Won Lee , Hoongyo Oh , Minkyu Park , Donkoan Hwang , Moo Hwan Kim , HangJin Jo

引用次数: 0

Analysis on containment thermal hydraulic behaviour under passive containment heat removal system operation condition for HPR1000 HPR1000被动安全壳排热系统运行工况下安全壳热工性能分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-03-01 DOI: 10.1016/j.jandt.2022.02.004

Jing Sun, Hui Wang, Jingjing Li

{"title":"Analysis on containment thermal hydraulic behaviour under passive containment heat removal system operation condition for HPR1000","authors":"Jing Sun, Hui Wang, Jingjing Li","doi":"10.1016/j.jandt.2022.02.004","DOIUrl":"10.1016/j.jandt.2022.02.004","url":null,"abstract":"<div><p>HPR1000 NPP is designed with Passive Containment Heat Removal System (PCS) following the technical route called active and passive combined concept. The installed PCS system will have effect on the thermal hydraulic (T-H) behavior in the containment, and the complex T-H environment will influence the heat removal and operating character of PCS system backwards. The main goal of this work is to evaluate the effectiveness of the PCS system of HPR1000 and analyse the complex T-H environment of HPR1000 when the PCS system is operating. The HPR1000 containment is modelled by the containment T-H code and the PCS system is modelled based on experiment results. Then, the severe accident sequence initiated by large break LOCA is calculated by an integrated computer code, which outputs mass and energy release sources from reactor coolant primary system as boundaries for the containment T-H code. The results show that HPR1000 PCS system has enough heat removal capability under selected severe accident, which could ensure the containment meet the design requirement. Also, the nonuniformity of temperature and gases in the containment will disappear when the mass and energy releases turn stable, and the operating of PCS system does not cause a huge effect on containment inhomogeneity.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 1","pages":"Pages 36-42"},"PeriodicalIF":0.0,"publicationDate":"2022-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000047/pdfft?md5=370f58bf57e5db8124c52f234a1e1fe5&pid=1-s2.0-S2468605022000047-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76190681","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 3

Heat transfer characteristics of single-layer and two-layer corium pool in elliptical lower head 椭圆下封头单层和双层堆芯池的换热特性

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-03-01 DOI: 10.1016/j.jandt.2022.02.001

Y.P. Zhang , D.H. Zhu , K. Ge , Y.W. Wu , W.X. Tian , G.H. Su , S.Z. Qiu

{"title":"Heat transfer characteristics of single-layer and two-layer corium pool in elliptical lower head","authors":"Y.P. Zhang , D.H. Zhu , K. Ge , Y.W. Wu , W.X. Tian , G.H. Su , S.Z. Qiu","doi":"10.1016/j.jandt.2022.02.001","DOIUrl":"10.1016/j.jandt.2022.02.001","url":null,"abstract":"<div><p>In this study, we investigated the characteristics of natural convection heat transfer of single-layer and two-layer corium pool in the elliptical lower head by the numerical method. Numerical simulations were performed using FLUENT software with the WMLES (Wall-Modeled Large Eddy Simulation) turbulence model, solid-liquid phase change model and the VOF (Volume of Fluid) model. The heat transfer characteristics such as pool temperature field, flow velocity field and heat flux distribution were obtained. For the single-layer corium pool, different degrees of melting occurred on the wall surface from about 1712 mm arc length inside the inner wall to the liquid level. The transient velocity field of single-layer corium pool showed that the natural convection first occurs near the wall surface and then gradually forms in the center of the corium pool as the temperature rises. When the quasi-steady state is reached, the upper part of the pool forms strong turbulence, while the lower part has lower velocity and forms transverse flow. For the two-layer corium pool, the lower head wall is more severely melted than the single-layer configuration. Whereas, the crust at the wall near the stratified interface is only slightly melted due to the lower flow rate and weaker heat transfer capacity. The research revealed the flow and heat transfer mechanisms of corium pool, and provided reference for the technology design, system optimization and safety evaluation of the IVR (In-Vessel Retention) – ERVC (External Reactor Vessel Cooling) severe accident mitigation strategy for the large-scale advanced pressurized water reactor.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 1","pages":"Pages 13-24"},"PeriodicalIF":0.0,"publicationDate":"2022-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000011/pdfft?md5=8fcd9cb6e8c4c8eafe777db6c9f316f0&pid=1-s2.0-S2468605022000011-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78769317","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 1

Improvement and evaluation of ATWS Protective Signal and mitigation system for ACPR1000 nuclear power plant ACPR1000核电站ATWS保护信号与减缓系统的改进与评价

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-03-01 DOI: 10.1016/j.jandt.2021.12.002

Juanhua Zhang

引用次数: 2

Low temperature overpressurization analysis for CPR1000 CPR1000低温超压分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2021-01-01 DOI: 10.1016/j.jandt.2021.09.001

Yu Liang , Mingjun Wang , Zhen Li , Haoyu Liao , Dalin Zhang , Wenxi Tian , Suizheng Qiu , G.H. Su

{"title":"Low temperature overpressurization analysis for CPR1000","authors":"Yu Liang , Mingjun Wang , Zhen Li , Haoyu Liao , Dalin Zhang , Wenxi Tian , Suizheng Qiu , G.H. Su","doi":"10.1016/j.jandt.2021.09.001","DOIUrl":"10.1016/j.jandt.2021.09.001","url":null,"abstract":"<div><p>An additional low temperature overpressure protection system which relies on the Pressurizer (PRZ) pressure relief valves has been put forward under the situation in which the Residual Heat Removal System (RHRS) is unavailable or isolated. Taking into account the risk of Reactor Pressure Vessel (RPV) brittle fracture and LOCA under cold overpressurization transients, it is urgent to strengthen the low temperature overpressure protection for in-service Pressurized Water Reactors (PWR) in China. In this paper, the simulation and analysis of two types of overpressure transients at low temperature during the shutdown process are carried out through the detailed Relap5 modeling of CPR1000. The results show that the place where the brittle fracture occurs firstly is the lower plenum. The sensitivities of the system temperature and the opening/closing time of the additional protection system's valve are analyzed. The system set point becomes more restrictive at lower coolant temperature. Additional threshold values of the PRZ pressure relief valve are given under different Effective Full Power Years (EFPY). The set point should be less than 5.5 MPa in the case of 5 EFPY and less than 3.7 MPa at 10 EFPY.The results can be applied to engineering practice, which can effectively improve the safety of CPR1000.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"3 ","pages":"Pages 145-153"},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S246860502100017X/pdfft?md5=8add661d101a165bb6b67c928fcbbd76&pid=1-s2.0-S246860502100017X-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85984588","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Determination of the neutron and gamma ray dose rates of the irradiating beam tubes of a pool-type research reactor by using Monte Carlo simulation and experimental detectors regarding radiation protection issues 用蒙特卡罗模拟和实验探测器测定池型研究堆辐照束管的中子和伽马射线剂量率

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2021-01-01 DOI: 10.1016/j.jandt.2021.06.001

Afshin Hedayat, Javad Emami, Ramin Salartash

{"title":"Determination of the neutron and gamma ray dose rates of the irradiating beam tubes of a pool-type research reactor by using Monte Carlo simulation and experimental detectors regarding radiation protection issues","authors":"Afshin Hedayat, Javad Emami, Ramin Salartash","doi":"10.1016/j.jandt.2021.06.001","DOIUrl":"10.1016/j.jandt.2021.06.001","url":null,"abstract":"<div><p>In this paper, the reactor core, irradiating beam tubes, and radiological shields of a 5 MW open pool type Material Testing Reactor (MTR) are simulated in detail by using the MCNPX 2.6 code and its default library (i.e. mostly made of ENDF/B VI and VII). A safety assessment is performed and discussed based on the related health physics issues. Independently, dosimetry parameters are measured by using the Berthold LB 6411 neutron dose rate monitor and Berthold LB 123 Gamma dose rate monitor. Experimental results are used to benchmark the modeling and calculations especially regarding dosimetry, shielding, and health physics problems. Results are fairly appropriate for further calculations to be validated but some aging problems could be raised for a 50 years old research reactor particularly due to secondary gamma rate of the activated components. Then major beam tubes are characterized for high qualified irradiating applications. Moreover, external dose rates are estimated for empty beam tubes whenever radiological shields fail such as large break of beam tubes, hazardous seismic conditions, or any accident that can remove the plugs. Then, results are very important for the safety of the reactor operator to determine and establish emergency zones (i.e. yellow zones) and planning, respectively.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"3 ","pages":"Pages 27-43"},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/j.jandt.2021.06.001","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"106468085","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 3

Reactor dynamic simulation to analyze possible scenarios after a spurious opening of the safety flapper valve of TRR during the normal operation regarding the inherently safety features and Design Extension Conditions (DEC) by using the RELAP5 code 利用RELAP5规范，对TRR安全挡板阀在正常运行过程中假开启后可能出现的情况进行反应堆动态仿真，分析其固有安全特性和设计扩展条件(DEC)

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2021-01-01 DOI: 10.1016/j.jandt.2021.07.005

Afshin Hedayat

{"title":"Reactor dynamic simulation to analyze possible scenarios after a spurious opening of the safety flapper valve of TRR during the normal operation regarding the inherently safety features and Design Extension Conditions (DEC) by using the RELAP5 code","authors":"Afshin Hedayat","doi":"10.1016/j.jandt.2021.07.005","DOIUrl":"10.1016/j.jandt.2021.07.005","url":null,"abstract":"<div><p>LOFA is one of the most important PIE in nuclear reactors. For the pool-type research reactor, the safety flapper valve is usually located below the reactor core and have a very important rule on the reactor safety especially during the loss of the pump accident. In this paper, TRR has been simulated and analyzed against the spurious opening of the safety flapper valve during normal full power operation. First of all, the steady state parameters are successfully benchmarked against the operational data of TRR. Then, the reactor dynamic calculations are successfully benchmarked against a sophisticated experimental dynamic test to evaluate the reactor power changes due to reactivity feedback. Then three different types of possible scenario including both types of DBA and DEC are simulated and analyzed. Furthermore, inherently safety features are evaluated against temperature rises. Results indicate that if the anticipated emergency shutdown signal is triggered successfully, the reactor core remains safe against any physical damage. If the reactor reactivity control system is completely disabled, the inherently safety features of the reactor decrease the reactor power effectively via negative reactivity feedback and the reactor fuel assemblies remain safe again. But if the fault diagnosis system of the spurious opening of the safety flapper valve fails to detect the occurrence or it does not send the emergency shutdown signal, and at the same time, the automatic power regulation system keeps the normal operating conditions, the hot spot may reach to the fuel melting point just a few seconds after the accident occurrence.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"3 ","pages":"Pages 119-133"},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/j.jandt.2021.07.005","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75503728","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Neutrons diffusion variable coefficient advection in nuclear reactors 核反应堆中子扩散变系数平流

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2021-01-01 DOI: 10.1016/j.jandt.2021.06.005

Rami Ahmad El-Nabulsi

引用次数: 1