Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation最新文献

{"title":"Tests of the Package for the Transport of Natural Uranium Hexafluoride","authors":"Sun Hongchao, L. Guoqiang, Zhuang Dajie, Sun Shutang, Meng Dongyuan, Lian Yiren, Chen Lei, Zhang Jiangang","doi":"10.1115/ICONE26-82151","DOIUrl":"https://doi.org/10.1115/ICONE26-82151","url":null,"abstract":"Several tests has been conducted to illustrating the safety performance of a type of package for the transport of natural uranium hexafluoride meet the requirements of GB11806-2004 (Regulations for the safe transport of radioactive material). The requirements of GB11806-2004 are same with the requirements of IAEA SSR6 (Regulations for the safe transport of radioactive material). These tests include heat, cold, reduced external pressure, increased external pressure, free drop, thermal test. Certification testing was performed on full-scale model and a test plan was developed that identified the specific free drop necessary to evaluate both GB11806 and SSR-6 requirements. A total of two 0.6-m free drops were performed. The leaks were detected after each free drop test. The accelerations were recorded for use in finite element structural analyses. This paper reviews the test planning and results with a discussion of how the test and finite element structural analyses were combined.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"2 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121111757","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}

{"title":"The Design of Data Structure and Interface for Nuclear Emergency Assessment and Decision Support System","authors":"Yang Yapeng, Wei Hong, Feng Zongyang, Jia Linsheng, X. Xiaoxiao, Zhang Jiangang","doi":"10.1115/ICONE26-81885","DOIUrl":"https://doi.org/10.1115/ICONE26-81885","url":null,"abstract":"Emergency Assessment and Decision Support System has become an important tool to effectively respond to a nuclear emergency in nuclear power plant, the system mainly used to on-line assessment the real-time reactor conditions, calculate the source term released to the environment, and suggest the protection actions taken to protect the public and emergency workers. This paper describe the development of the system, analysis the business flow chart of nuclear assessment, introduce details of several functional modules, such as core damage assessment, source term calculation and radiological consequence assessment. The system’s data structure and data interface also be describe in detail.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"64 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124824892","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}

{"title":"Verification of Shielding Calculation Capability of RMC With H.B.Robinson-2 Pressure Vessel Benchmark","authors":"Junjie Rao, Xiaotong Shang, Kan Wang","doi":"10.1115/ICONE26-81694","DOIUrl":"https://doi.org/10.1115/ICONE26-81694","url":null,"abstract":"RMC is a 3-D continuous energy Monte Carlo code developed by REAL team in Tsinghua University, China. Besides the capability of fuel cycle burnup calculation, hybrid MPI/OpenMP parallelism strategy, sensitivity and uncertainty analysis, N-TH coupling calculation, shielding calculation methods including general source description, regional importance method, weight window method and source biasing method have been also developed for deep penetration problems. H.B.Robinson-2 Pressure Vessel Benchmark (HBR-2 benchmark) is used for the qualification of pressure vessel neutron flux calculation methods and shielding calculations based on this model have been performed by Monte Carlo codes such as SCALE, MCNPX and deterministic transport code DORT. In this work, the verification calculation of shielding calculation capability of RMC is conducted based on HBR-2 benchmark. The total calculation consists of two stages. Criticality calculation is performed first to obtain the fission neutron distribution in the reactor core assemblies. Then the fission neutron distribution is regarded as the initial neutron source in the following fixed source calculation. Variance reduction techniques such as source biasing and regional importance methods are combined together to be able to reduce the variance of the neutron flux in regions within and outside the pressure vessel including the downcomer and cavity regions. The preliminary calculation results show good agreement with MCNP and the shielding calculation of RMC is justified and applicable for deep penetration problems.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"19 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128312129","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}

{"title":"Uncertainty Analysis of Power Distribution for NESTOR Based on the Double Latin Hypercube Sampling Method","authors":"Hongkuan Liao, Qing Li, Yingrui Yu, Yuying Hu, L. Wu, Chenlin Wang, Jinyu Wang, Jinghui Wang, Peng Xiao","doi":"10.1115/ICONE26-81441","DOIUrl":"https://doi.org/10.1115/ICONE26-81441","url":null,"abstract":"Due to the complexity of the reactor system, many approximations are used in the nuclear design and calculations inevitably. The accuracy of the nuclear design software is closely related to the safety of the reactor design and operation. Besides, improving the accuracy is an effective way to excavating the economy of nuclear power plants. To analyze the uncertainty of power distribution for NESTOR nuclear design software, we suggested an uncertainty analysis method based on the double Latin Hypercube Sampling (LHS) method and the random sampling statistical analysis (RSSA) method, and built an uncertainty analysis process based on LHS. The uncertainty of physical model and the uncertainty of the change of parameters were both taken into consideration with the double samplings, and 3481 core states were generated by the double samplings. Therefore, the uncertainty of power distribution could be directly analyzed through modeling computation, and the uncertainty of radial power distribution was achieved as ±3.856% under the condition of 95% confidence coefficient and 95% probability. Meanwhile, according to deviation transmission idea, we obtained the uncertainty of power distribution from physical models and the change of parameters based on the measured method. The result shows that the accuracy using the double sampling method is nearly the same to which achieved by the deviation transmission idea, and more conservative.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"97 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124537312","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}

{"title":"Study on Offsite Emergency Preparedness for the Industry Application of HTR-PM","authors":"Hongchun Ding, J. Tong, Liguo Zhang","doi":"10.1115/ICONE26-81166","DOIUrl":"https://doi.org/10.1115/ICONE26-81166","url":null,"abstract":"The offsite consequence study on High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM) shows that its emergency planning zone (EPZ) for whether plume exposure pathway or ingestion exposure pathway can both be limited within the exclusion area boundary (EAB) of nuclear power plant (NPP), which is about a few hundred meters away from the reactor. This conclusion provides theoretical basis and strong technical support for potential industrial applications of HTR-PM. By this premise, the coupling risk for the industrial application of HTR-PM is quantitatively evaluated in this paper. The results indicate that the risk increments for both HTR-PM and industrial facility due to the colocation on one site could ensure they still meet their corresponding risk criteria. On this basis, the current proper management mode is recommended for the ally of HTR-PM and industrial facilities from technical perspective, which is the onsite of this HTR-PM-industry ally can adopt the mode of separated management, while the offsite co-management. Further, the final suggestions on offsite emergency preparedness for this HTR-PM-industry ally are also given out co-considering the current policy, practice and public acceptability. This work will also support the industrial application of other advanced reactors with inherent safety features.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"37 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130076356","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}

{"title":"Identification of Gas Accumulation Susceptibility in NPP’s Safety Related Systems and Operability Evaluation due to Gas Transportation","authors":"Zhen-Yu Hung, Pei-Hsun Huang, Chaojun Li","doi":"10.1115/ICONE26-81074","DOIUrl":"https://doi.org/10.1115/ICONE26-81074","url":null,"abstract":"Instances of gas accumulation in the subject systems have occurred since the beginning of commercial nuclear power plant operation. NRC Generic Letter 2008-01, “requests that each licensee evaluate its ECCS, DHR system, and Containment Spray system licensing basis, design, testing and corrective actions to ensure that gas accumulation is maintained less than the amount that challenges operability of these systems, and that appropriate action is taken when conditions adverse to quality are identified.” All of the three NPPs have accomplished this evaluation and propose some corrective measures like revision of the operation procedures, installing the venting valves etc. Taipower also committed AEC to establish the acceptable quantity of the gas accumulation and continue to follow the development of gas transport methodologies in the industry. According to the NRC NRR Action Plan TAC.NO. ME3939 GAS MANAGEMENT (March 2011), other safety related systems, in additional to the systems covered by GL 2008-01, also have the gas accumulation issue and therefore must be concerned. This project will develop the numerical evaluation process with two phase flow software to simulate the gas accumulation and the transportation phenomena for the GL 2008-01 systems and validate the results by experimental analysis.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"2 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134111533","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}

{"title":"Structural Redundancy Design and Reliability Analysis of Magnetic Bearing for HTGR Primary Helium Circulator","authors":"Yangbo Zheng, Zhengang Shi, Xingnan Liu, N. Mo, Guojun Yang","doi":"10.1115/ICONE26-81336","DOIUrl":"https://doi.org/10.1115/ICONE26-81336","url":null,"abstract":"As primary pump of High Temperature Gas-cooled Rector (HTGR), the safety and reliability of primary helium circulator has a direct relationship with heat export in the reactor core. This paper proposed two conceptual design methods, which are structural redundancy design in the axial and radial magnetic bearings (coils), and crossing redundancy design between amplifiers and the redundant bearings (coils), combined with the characteristics of magnetic bearing in the primary helium circulator. Then we built Markov model for the magnetic bearing with redundant structure, which was used to reliability analysis and calculation. Finally, we performed simulation of the redundant magnetic bearings in Matlab/Simulink. The results show that the redundant structure design can greatly improve the reliability of the magnetic bearing system. Therefore, these methods would provide a new approach for further improving the safety and reliability of primary helium circulator.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128830767","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}

{"title":"The Development of the Advanced Method for the Source Term Evaluation Applicable to the Dynamic PRA","authors":"Koichi Nakamura, Sunghyon Jang, A. Yamaguchi","doi":"10.1115/ICONE26-82523","DOIUrl":"https://doi.org/10.1115/ICONE26-82523","url":null,"abstract":"Useful insights on nuclear safety are provided by the level 2 probabilistic risk assessment (PRA), which evaluates the risk of fission products (FP) released in the environment during an accident in nuclear power plants (NPPs). The containment event tree method is generally employed for level 2 PRA. In this method, the accident scenarios are expressed by the combination of a number of branch points. The possible accident scenarios are approximately or representatively dealt with in this method. Dynamic PRA can evaluate the accident risk with complex changes or composition of various accident events dynamically. A reasonable source term evaluation method, which can be conducted with a small computational load, is developed for the establishment of dynamic PRA method focused on the risk of the release of FP in the environment. The proposed study aims to develop a reasonable source term evaluation method by applying phenomenological relationship diagram (PRD) method applicable to the dynamic PRA is developed.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130777039","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}

{"title":"A Critical Experimental Study of Bubble Effect in the Process of Spent Fuel Dissolving","authors":"Z. Xiaoping, Liang Shuhong, Xia Zhaodong","doi":"10.1115/ICONE26-81644","DOIUrl":"https://doi.org/10.1115/ICONE26-81644","url":null,"abstract":"The influence of gas introduction on the critical safety of the nuclear fuel system under the condition of cold condition, given reactor material and geometry structure is studied.\u0000 Refer to bubble effect test experiment on nuclear critical safety test device (YSR) and considering solid-liquid two-phase nuclear fuel system with uranyl nitrate solution - uranium dioxide fuel element as the experimental platform, the dynamic process of the real behavior of bubbles in uranyl nitrate solution has been simulated in the quasi-static way by replacing bubble generator with aluminous bubble simulation elements.\u0000 Bubble effect is the reactivity change caused by the change of volume of solution, neutron leakage and absorption property in the nuclear fuel system due to the bubbles generated in the solution. In the dissolving process of spent fuel, oxygen or nitrogen are usually added to accelerate the dissolution of fuel element shear section, and some other bubble production are also caused by the heat released during the dissolution process. Here, the bubble production caused by the heat is omitted and only artificial gas introduction is considered in my study. When there are enough bubbles in the uranium solution system, the volume of the solution will increase, which will inevitably influence the absorption and leakage property of the neutrons, and further influence the reactivity of the nuclear fuel system.\u0000 The corresponding relationship between the bubble-intake rate and the bubble equivalent diameter, arising velocity and bubble share is determined through fluid dynamics modeling to manufacture the aluminous bubble simulation elements. The theoretical calculation by MONK9A and the critical experimental measurements are also compared and analyzed in this paper.\u0000 The results showed that the reactivity caused by bubbles was negative, and the greater the intake rate, the greater the negative effect. Meanwhile the theoretical calculated value was in good agreement with the experimental value and the maximum deviation was 63.4 pcm.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122898079","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}

{"title":"Preliminary Hazard Analysis of Uranium Hexafluoride Accident","authors":"Chen Lei, Zhang Jiangang, L. Guoqiang, Sun Shutang, Meng Dongyuan, W. Ning","doi":"10.1115/ICONE26-81956","DOIUrl":"https://doi.org/10.1115/ICONE26-81956","url":null,"abstract":"Uranium hexafluoride (UF6) accident is a typical accident in the nuclear fuel cycle. It combine radioactive and chemical hazards, so it is necessary to attach great importance to the UF6 accident. This paper analyzed and summarized the possible accident scenarios, causes and consequences, and classified the UF6 accident risk factors, and put forward corresponding preventive and emergency measures. Preliminary hazard analysis of the accident can help us better understand the accident process, so we can takes steps for corresponding risk factors in advance, and prevent it will not be developed into an accident, so than we can obtain the effect of nip in the bud.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"11 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121236585","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}