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

{"title":"Facilities and Experience on Impact Test of Packages for Radioactive Materials Transport","authors":"L. Guoqiang, Zhuang Dajie, Wang Xuexin, Meng Dongyuan, Zhang Jiangang, Sun Hongchao, Sun Shutang, Ma Anping","doi":"10.1115/ICONE26-81451","DOIUrl":"https://doi.org/10.1115/ICONE26-81451","url":null,"abstract":"According to the requirements of Chinese GB11806 and IAEA’s SSR-6, the packages for radioactive materials transport shall be tested to demonstrate the ability to withstand transport accidents. A small facility and a large one for package impact test have been designed and constructed to conduct the free drop test I experiment required in the GB11806 and SSR-6. The target of the smaller facility is a reinforced concrete component with volume of 3800mm × 3550mm × 3500mm. It is riveted by steel plate marked Q460 with volume of 3500mm × 3250mm × 100mm at surface. Total mass of the target is about 130t. The hoisting facility is mobile crane. The facility is used for the impact test of package within 13t. The target of the larger facility is a reinforced concrete component with volume of 11000mm × 7350mm × 7000mm. Steel plate marked Q460 with a volume of 11000mm × 5380mm × 100mm was anchored at concrete component top. Total mass of the target is about 1400t. The maximum height of the lifting hook in tower is 17.1m. The larger facility is used for the impact test of package within 130t. The two facilities for package impact test are both equipped with acceleration and stress measuring system, image measuring instrument, and geometrical measuring system. These two facilities have been used to conduct a lot of impact test of small packages weighing less than 12t. Some experience in construction and experiment has been concluded.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"49 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":"127740501","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":"Validation of Film Evaporation Model in GASFLOW-MPI","authors":"Liu Yabing, Z. Han, Xia Jianjun","doi":"10.1115/ICONE26-81249","DOIUrl":"https://doi.org/10.1115/ICONE26-81249","url":null,"abstract":"A dynamic film model is developed in the parallel CFD code GASFLOW-MPI for passive containment cooling system (PCCS) utilized in nuclear power plant like AP1000 and CAP1400. GASFLOW-MPI is a widely validated parallel CDF code and has been applied to containment thermal hydraulics safety analysis for different types of reactors. The essential issue for PCCS is the heat removal capability. Research shows that film evaporation contributes most to the heat removal capability for PCCS. In this study, the film evaporation model is validated with separate effect test conducted on the EFFE facility by Pisa University. The test region is a rectangle gap with 0.1m width, 2m length, and 0.6m depth. The water film flowing from the top of the gap is heated by a heating plate with constant temperature and cooled by countercurrent air flow at the same time. The test region model is built and analyzed, through which the total thermal power and evaporation rate are obtained to compare with experimental data. Numerical result shows good agreement with the experimental data. Besides, the influence of air velocity, wall temperature and gap widths are discussed in our study. Result shows that, the film evaporation has a positive correlation with air velocity, wall temperature and gap width. This study can be fundamental for our further numerical study on PCCS.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"33 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":"128131855","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":"QUENCH-06 Experiment Post-Test Calculations and Integrated Uncertainty Analysis With RELAP/SCDAPSIM/MOD3.4 and MOD3.5","authors":"C. Allison, B. Le, G. Gerova, Ivan G. Spasov, M. Pérez-Ferragut, J. Hohorst","doi":"10.1115/ICONE26-81912","DOIUrl":"https://doi.org/10.1115/ICONE26-81912","url":null,"abstract":"The Quench-06 experiment (ISP-45) has been used as a benchmark and training aid for Innovative Systems Software (ISS) and our users/students since it was completed in the early 2000s. The experiment was first analyzed by several international organizations using RELAP/SCDAPSIM/MOD3.2. These results were submitted to the “blind” and “open” phases of the ISP. The experiment was subsequently used for basic user training for experimental analysis by our RELAP/SCDAPSIM/MOD3.4 and MOD3.5 users. It is also used extensively in our university support and training internships.\u0000 This paper describes an integrated uncertainty analysis of the QUENCH-06 electrically heated experiment, looking at the influence of uncertainties in experimental conditions and important models/correlations. The QUENCH calculations demonstrated the use of the new IUA, “Integrated Uncertainty Analysis”, option introduced into RELAP/SCDAPSIM/MOD3.4 in the summer of 2017 and MOD3.5 in the fall of 2017. The input models and results from both versions are discussed in the paper. The MOD3.4 results are based on the original input model developed for MOD3.2 and refined in the open phase of the ISP. The MOD3.5 results are based upon two base input models. The first was developed specifically to test the impact of MOD3.5 modeling improvements for the Quench electrically heated fuel rod simulator and was used in an early paper presented at this meeting in 2014. The second input model has been refined as part of the university support and training internship program and was used originally in 2016 to look at the influence of different approaches in modeling the insulated shroud used to minimize radial heat losses.\u0000 The uncertainty analysis provided in this paper looks at the influence of uncertainties in (a) the parabolic equations for Zircaloy oxidation, (b) the tungsten heater element resistances, (c) the convective heat transfer coefficients, (d) the contact resistance of the heater elements, and (e) the thermal conductivity of the porous zirconia used in the shroud. The uncertainty analysis demonstrated very clearly that a bias was introduced into the 2014 MOD3.5 input model. This bias was subsequently determined to be primarily associated to the modeling of the shroud and associated radial heat losses. This bias was reduced in the 2016 version of the MOD3.5 input model and will be further refined as our training activities on the modeling of experiments continue.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"33 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":"134038240","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":"Analysis Contour Plots in RELAP/SCDAPSIM/MOD 3.4 and MOD 4.1","authors":"H. Sánchez-Mora, C. Chávez-Mercado, C. Allison, J. Hohorst","doi":"10.1115/ICONE26-81991","DOIUrl":"https://doi.org/10.1115/ICONE26-81991","url":null,"abstract":"RELAP/SCDAPSIM is a nuclear reactor simulator and accident analysis code that has been used in the nuclear energy industry for many years. Currently, Innovative Systems Software is developing a new tool that will show the behavior of the core components during a simulation of an accident. The addition of contour plots for the SCDAP components showing different properties: temperature, hydrogen production, etc. will allow a better understanding of core behavior during a severe accident in a nuclear reactor. The SCDAP components are fuel rods, electrically heated simulator rods, such as those used in the CORA experiments, control rods, a shroud and a BWR blade/box. This paper describes the progress in the development of the contour plot tool based on the OpenGL and FORTRAN90 libraries. The purpose of this tool is help to the user analyze the simulation of an accident and to debug an input file.","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":"132222659","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":"Research on Optimization of Ingestion Emergency Planning Zone Sizing","authors":"Mengxi Wang, Na Xue, Xinjian Liu","doi":"10.1115/ICONE26-81285","DOIUrl":"https://doi.org/10.1115/ICONE26-81285","url":null,"abstract":"Food contamination has aroused public concern since Fukushima accident. As emergency preparedness is often viewed as an important approach to protect staff working on site and public around the site, ingestion emergency planning zone (EPZ) is applied to protect public from the exposure of contaminated food. Ingestion EPZ is one of the technical foundations for nuclear emergency preparedness, which will be influenced by design features of plant and characteristics of the site. This paper is devoted to the research on the optimization of ingestion EPZ sizing from the view of the atmospheric dispersion model and the food chain model, which are crucial points for the sizing of ingestion EPZ. Compared to the traditional straight-line Gaussian plume model with a quite conservative assumption that plume segments always transport in the downwind direction, the Lagrangian Gaussian puff model considers the swing of wind direction over time, which makes the simulation more realistic. With the results of radionuclide concentrations evaluated by the dispersion model, the transportation of the radionuclides in food is simulated by the food chain model. The traditional food chain model is essentially a static model with no consideration that food contamination level has a strong dependence on the accident date, which may overstate the risk from nuclear plant accidents and result in unfounded fear of public. The dynamic food chain model, which takes daily changes of plant biomass, or livestock feeding periods in consideration, has been developed to estimate radionuclide concentrations in different foodstuffs. On basis of the study of the dispersion models and food chain models above, we evaluate the ingestion EPZ size of Tianwan NPP by choosing the comparatively realistic ones from them. In the scenario considered in this paper, the simulation domain of Tianwan NPP within 80km-range and hourly time-step is applied, and meteorological conditions are carefully set according to observation data in recent years. Results show that there is significant margin and conservatism in the traditional ingestion EPZ sizing. Radionuclide concentrations predicted by the Lagrangian Gaussian puff model is almost an order of magnitude lower than the Gaussian plume model. Moreover, the dynamic food chain model considers the seasonal effect that simulation results of radionuclide concentrations in foodstuffs are significantly higher in summer than in winter, which helps to make a more realistic consideration of ingestion pathway. This research gives an example of the application of new models for the optimization of ingestion EPZ sizing, which may contribute to strengthen public confidence in nuclear safety and emergency preparedness.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"42 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":"123335908","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":"Numerical Simulation of Debris Bed Relocation Behavior in Sodium-Cooled Fast Reactor","authors":"Chunming Teng, Bin Zhang, J. Shan","doi":"10.1115/ICONE26-82493","DOIUrl":"https://doi.org/10.1115/ICONE26-82493","url":null,"abstract":"For the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel or steel is solidified into debris particles which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes debris relocated and flattened, which called debris relocation. The thickness of debris bed has great influence to the cooling ability of fuel debris in low plenum. To ensure the effective implementation of the in-vessel retention (IVR), it’s very necessary to evaluate the transient changes of shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a debris relocation model based on COMMEN code was developed in this paper. The debris relocation model was established based on the extrapolation of the shear strength mechanism, which was originally proposed and widely applied in soil mechanics filed. Shear strength is a function of the particles’ density and position. Debris bed is fluidized only when the shear stress in particle unit is larger than shear strength of debris particles. By integrating the debris relocation model into the COMMEN code, the transition process of the bed in depressurization experiments was simulated and compared against the experimental results. Good agreement shows that the debris relocation model presented in this paper can reasonably simulate the relocation behavior.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"193 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":"122650964","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":"Noise Reduction Treatment and Analysis of Accumulated Radon Concentration in Uranium-Like Rock Based on Wavelet Theory","authors":"F. Jiang, Wenchao Yang, Ming Li, Xiangyang Li, Hong Changshou","doi":"10.1115/ICONE26-81068","DOIUrl":"https://doi.org/10.1115/ICONE26-81068","url":null,"abstract":"Based on the principle of similarity, uranium ore samples were prepared from raw materials such as uranium tailings, quartz sand and refined iron powder, and then the samples were treated with different packages for measuring the sample accumulated radon concentration. In the actual measurement process, due to the characteristics of radionuclide decay, instrument reasons and human factors, the data will be a certain deviation. Therefore, the method of wavelet analysis is used to denoise the accumulated radon concentration and obtain radon exhalation rate. The results of the study show: the correlation coefficient of cumulative radon concentration fitted by wavelet denoising is improved greatly, and all of them are above 0.99, the recalculated radon exhalation rate of the single side of the sample is decreased by 0.06Bq · m−2 · s−1, and double-edged is decreased by 0.02Bq · m−2 · s−1. The experiment proved that wavelet theory can be used to correct calculated value of radon exhalation rate of uranium-like rock. At the same time, it provides a new method for further study of uranium mine radiation protection parameters.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"34 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":"126772863","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":"Numerical Research on Shock Resistance Safety Analysis of Ship Power Plant Valve","authors":"Jun Wu, Fan Bai, Yong Liu, X. Lao, C. Dai","doi":"10.1115/ICONE26-82137","DOIUrl":"https://doi.org/10.1115/ICONE26-82137","url":null,"abstract":"In order to increase the operation safety of ship nuclear power platform. A typical butterfly valve with nuclear safety class 2 of ship power plant is studied in this paper. Numerical analysis on shock resistance safety of butterfly valve is conducted by three-dimensional finite element method. Firstly, the load type under working condition and the stress evaluation criteria of valve safety analysis are determined based on national standards and ASME standards. Secondly, the finite element model of butterfly valve is constructed and the equivalent stress distribution of butterfly valve is obtained under static loads and constraint conditions. Moreover, the butterfly valve natural frequencies and vibration shapes are obtained by modal analysis under inspection working condition. Furthermore, the butterfly valve stress distributions on three coordinate directions are analyzed by shock response spectrum analysis according to the modal analysis results. Lastly, the total stress distribution of butterfly valve is obtained by superposition procedure. And the butterfly valve structure shock resistance safety is checked based on the stress evaluation criteria.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"13 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":"127851252","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":"Segmented γ Scanning Device and its Experimental Research","authors":"S. Hou, Chen Chen, Quan-hu Zhang, X. Su, Wenming Zuo","doi":"10.1115/ICONE26-82327","DOIUrl":"https://doi.org/10.1115/ICONE26-82327","url":null,"abstract":"Segmented Gamma Scanner (SGS) is a commonly used nondestructive testing (Non-Destructive Assay NDA) method. SGS uses radial rotation, axial segmentation, segmented scanning of the non-uniform sample of the uniform treatment, making it possible to accurately measure the radioactivity on each segment and is currently widely used in the field of nuclear material management. This paper introduces a self-designed SGS measuring device and uses this device to perform a large number of scanning experiments on different measurement objects in the laboratory. It also studied the detection performance, stability and SGS method of different objects the accuracy of the measurement results.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"21 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":"127598973","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":"Development of Neutronics and Thermal-Hydraulics Coupled Code for Accelerator Driven Subcritical Systems","authors":"Run Luo, Pengfei Wang, Xinyu Wei, S. Revankar, F. Zhao","doi":"10.1115/ICONE26-81276","DOIUrl":"https://doi.org/10.1115/ICONE26-81276","url":null,"abstract":"A new neutronics and thermal-hydraulics coupled code named ARTAP is developed to analyze the steady-state and transient characteristics of accelerator driven subcritical system (ADS) in this paper. Monte Carlo simulations for spallation neutron source and deterministic calculations for the subcritical core are performed in the steady-state analysis module. The ADS core is divided into a number of nodes both along the reactor axis and the fuel pellet radius for the calculation of spatial temperature distributions. The power iteration method is adopted to solve the coupled neutronics and thermal-hydraulics problems. The transient analysis module consists of space-time neutron kinetics model and thermal-hydraulics dynamic model, which is calculated by using numerical differentiation formulas (NDFs) method. The new code is verified by comparing its predictions for both the steady-state and transient cases of the OECD/NEA ADS benchmark. Results of numerical simulations indicate that ARTAP is reliable and efficient to be applied for the ADS analysis.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"95 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":"127406022","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}