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

{"title":"Review and Security Assessment of Red Oil Explosions in Evaporator","authors":"Lian Yiren, Sun Hongchao, Chen Lei, Meng Dongyuan, L. Guoqiang, Zhuang Dajie, Sun Shutang, Zhang Jiangang","doi":"10.1115/ICONE26-82221","DOIUrl":"https://doi.org/10.1115/ICONE26-82221","url":null,"abstract":"Purex progress is widely applied in the nuclear fuel reprocessing plants all over the world. However, various security problems occur in reprocessing facility involving the intense attention of red oil explosion. The exothermic reactions among TBP, nitric salts and nitric acid were responsible for the red oil explosion. In this paper, explosion events at nuclear fuel reprocessing plants initiated by red oil phenomena were reviewed. The formation and reaction mechanisms of red oil causing explosions were analyzed. Moreover, the evaluation and analysis model was built up to evaluate the security assessment of red oil explosion from the associated calculations of data in the typical red oil explosive accidents and the results of subsequent laboratory studies on the chemical reaction.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"340 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":"122785732","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":"Calculation of Core Damage Frequency Caused by Main Control Board Fire in the Main Control Room for Small Modular Reactors","authors":"Wanhong Wang, C. Peng, Yun Guo","doi":"10.1115/ICONE26-81467","DOIUrl":"https://doi.org/10.1115/ICONE26-81467","url":null,"abstract":"Fire accidents have had an adverse impact on nuclear safety in nuclear power plants. Fire probabilistic safety assessment, as one of the most important methods for safety analysis, has played an essential role in fire risk in nuclear power plants. Main control rooms normally contain many cabinets equipped with all kinds of cables. Once cables there get burned, if not tackled immediately in an appropriate way, they would possibly cause severe consequences. In order to assess the fire risk in the referenced main control room (MCR) for a small modular reactor, this study used the software, CFAST, to simulate fire scenarios caused by the main control board in the MCR, and use the software, Risk Spectrum, to calculate the Core Damage Frequency (CDF). The calculation results show that CDF caused by the main control board (MCB) is 6.112E−09/year.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"138 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":"122911061","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 Investigation of Corium Coolability in Core Catcher: Sensitivity to Modeling Parameters","authors":"Liancheng Guo, A. Rineiski","doi":"10.1115/icone26-81841","DOIUrl":"https://doi.org/10.1115/icone26-81841","url":null,"abstract":"To avoid settling of molten materials directly on the vessel wall in severe accident sequences, the implementation of a ‘core catcher’ device in the lower plenum of sodium fast reactor designs is considered. The device is to collect, retain and cool the debris, created when the corium falls down and accumulates in the core catcher, while interacting with surrounding coolant. This Fuel-Coolant Interaction (FCI) leads to a potentially energetic heat and mass transfer process which may threaten the vessel integrity. For simulations of severe accidents, including FCI, the SIMMER code family is employed at KIT. SIMMER-III and SIMMER-IV are advanced tools for the core disruptive accidents (CDA) analysis of liquid-metal fast reactors (LMFRs) and other GEN-IV systems. They are 2D/3D multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics codes coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. However, the experience of SIMMER application to simulation of corium relocation and related FCI is limited. It should be mentioned that the SIMMER code was not firstly developed for the FCI simulation. However, the related models show its basic capability in such complicate multiphase phenomena. The objective of the study was to preliminarily apply this code in a large-scale simulation. An in-vessel model based on European Sodium Fast Reactor (ESFR) was established and calculated by the SIMMER code. In addition, a sensitivity analysis on some modeling parameters is also conducted to examine their impacts. The characteristics of the debris in the core catcher region, such as debris mass and composition are compared. Besides that, the pressure history in this region, the mass of generated sodium vapor and average temperature of liquid sodium, which can be considered as FCI quantitative parameters, are also discussed. It is expected that the present study can provide some numerical experience of the SIMMER code in plant-scale corium relocation and related FCI simulation.","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":"121052771","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":"Blind Simulations of NACIE-UP Experimental Tests by STH Codes","authors":"N. Forgione, M. Angelucci, G. Barone, M. Polidori, Alessandra Cervone, I. Di Piazza, F. Giannetti, P. Lorusso, T. Hollands, A. Papukchiev","doi":"10.1115/ICONE26-81434","DOIUrl":"https://doi.org/10.1115/ICONE26-81434","url":null,"abstract":"In the frame of the SESAME project, a benchmarking activity was proposed to validate the existing system thermal-hydraulics codes for Heavy Liquid Metal reactors. More specifically, blind simulations on three well-defined experiments were carried out on the NACIE-UP facility, using CATHARE by ENEA, ATHLET by GRS, RELAP5-3D by University of Roma and RELAP5/Mod3.3 by University of Pisa. The numerical models were calibrated in terms of system thermal losses and gas enhanced circulation by means of the outcomes from specific experimental preliminary tests.\u0000 The present discussion expose, compare and analyze the numerical results of some representative parameters (primary lead-bismuth eutectic (LBE) mass flow rate, temperatures and pressure) charaterizing the system behaviour in transiet scenarios in a “pre-test” blind numerical assessment.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"27 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":"117014718","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 Continuous Integration Platform for the Deterministic Safety Analyses Code System AC2","authors":"J. Herb","doi":"10.1115/ICONE26-81123","DOIUrl":"https://doi.org/10.1115/ICONE26-81123","url":null,"abstract":"At GRS, a continuous integration platform based on the software Jenkins is used to verify and validate the code system AC2 consisting of the codes ATHLET, ATHLET-CD and COCOSYS. It automates different steps of the verification and validation process. If these steps were done manually, they would take up significant resources. Each time new or changed code is sent to the source code management system, the software is built for all supported operating systems and configurations and verifications as well as validation tests are executed. If test builds do not work or the results of verification or validation show deviations from the specified test targets, immediate actions can be taken to resolve any problems. During the last years the continuous integration platform has helped to improve both the quality of the software and productivity. It helps to maintain the robust quality management regime of the development process of AC2.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"38 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":"128832305","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 Hydrogen Treatment System in Severe Accident: Part 4 — Study of Fission Products and Steam Effect on Hydrogen Treatment Characteristics","authors":"A. Yamada","doi":"10.1115/ICONE26-81759","DOIUrl":"https://doi.org/10.1115/ICONE26-81759","url":null,"abstract":"A large amount of hydrogen is generated by the metal water-reaction in the Primary containment vessel (PCV) of light water reactors in the severe accident (SA). In the present accident management for boiling water reactor (BWR), vent of mixing gas with filtered vent is regarded as the most likely method that prevents the PCV overpressure. However, it is difficult to vent in early stage of SA because of high radioactive dose. Then we have been developing the hydrogen treatment system to prevent excessive pressure without PCV vent. In focusing on the oxidation-reduction reaction of metal oxides (MOs) with high reaction rate, we have been studying hydrogen treatment system using MOs as effective device under oxygen deficit conditions like PCV of BWR. In the previous studies, we evaluated the hydrogen treatment rate using a couple of MOs, and confirmed that CuO, Co3O4, and MnO2 were effective for the hydrogen oxidation under the oxygen-free condition. We also found that granules of these three MOs could achieve the goal of hydrogen treatment rate with reactor of hydrogen treatment system. We predicted that the performance of MOs decreased with exposure to steam and fission products (FPs) in the PCV during the hydrogen treatment, and investigated their influence. The objective of the present research is to investigate how the steam and FPs, which is supposed to be a reaction-inhibiting-factor, influence hydrogen treatment rate. Then, we conducted hydrogen treatment experiments using a fixed bed reactor with MOs layer. As the results, we confirmed that the hydrogen treatment rate of MnO2 decrease from 70 g/s/m3 to 15 g/s/m3 when partial pressure of vapor went above 0.1 MPa-abs, though, that of CuO didn’t depend on the partial pressure of vapor and sustain the same rate about 40 g/s/m3. We also confirmed that the hydrogen treatment rate was decreased with the consumption of granulated MOs faster than our expectation estimated with unreacted-core model*. We also estimated that CsI selected as typical FPs could not affect the hydrogen treatment rate of CuO. From these results, we have evaluated the reaction rate equation including the steam influence in CuO, which could estimate the hydrogen treatment rate of reactor unit.\u0000 *Gas reacts only on the surface of solid and generates shell of products around reactants core. The core shrinks with reaction.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"419 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2016-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124873320","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 Impact Simulation of Aircraft Into Reinforced Concrete Walls With Different Thickness","authors":"Kazuma Hirosaka, M. Nakane, S. Saigo, Norihide Tohyama","doi":"10.1115/ICONE26-82616","DOIUrl":"https://doi.org/10.1115/ICONE26-82616","url":null,"abstract":"Aircraft impact analysis is needed for a safety assessment of nuclear power plants. One of the contents which should be analysed for aircraft impact is physical damage of a concrete building and this can be estimated by a numerical simulation. In order to conduct aircraft impact analysis, simulation model which validated by some experimental data needs to be established.\u0000 In 1990s, impact test using actual F4 Phantom fighter was conducted at Sandia national laboratory in U.S. and a lot of important experimental data were measured. In this paper, the numerical simulation results for this F4 Phantom impact test are introduced. The relationship between the thickness of the shell of the F4 Phantom simulation model and the deceleration of this model is indicated and the differences of the deceleration between simulation and test results are discussed. In addition, the relationship between fracture strain of the shell of the F4 Phantom simulation model and the destruction mode of this model in simulation is indicated and the differences between the destruction mode of the F4 Phantom between simulation and test results are discussed.\u0000 In order to evaluate the physical damage area after the aircraft impact, it is necessary to estimate the aircraft velocity after it perforates the outer concrete wall and to calculate the decrease of the kinematic energy of the aircraft by this perforation. In this paper, several aircraft impact simulations with different concrete wall thickness are conducted and the reduction in kinematic energies of an aircraft by a perforation is estimated. Using these simulation results, the necessary numbers of concrete walls until the impacting aircraft stops is discussed.","PeriodicalId":394688,"journal":{"name":"Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation","volume":"239 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2015-08-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122630908","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}