Academic literature on the topic 'Reactor fuel reprocessing – Computer programs'
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Journal articles on the topic "Reactor fuel reprocessing – Computer programs"
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Blokhin, A. I., P. A. Blokhin, and S. T. Kazieva. "TRACT Computer Code Application in the Assessment of SNF and RW Class 1 Radionuclide Compositions and Radiation Characteristics." Radioactive Waste 13, no. 4 (2020): 99–111. http://dx.doi.org/10.25283/2587-9707-2020-4-99-111.
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The article presents the results of computational studies addressing radionuclide compositions and radiation characteristics of spent nuclear fuel from WWER-440 and BN-600 reactor units performed using new computer code TRACT. These studies are performed to acquire basic information enabling further forecasts of RW Class 1 characteristics generated from the reprocessing of these fuel types. The calculations allowed to identify the characteristics of WWER-440 and BN-600 SNF considering various irradiation and storage conditions, including different burnup levels and pre-reprocessing cooling times. Such data will enable further evaluation of possible ranges describing the characteristics of RW Class 1.
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Fan, Yan Ru, Xiang Jiang Wang, and Hai Tao Huang. "Research of Shear Force during Spent Fuel Reprocessing Shearing Process." Applied Mechanics and Materials 543-547 (March 2014): 533–36. http://dx.doi.org/10.4028/www.scientific.net/amm.543-547.533.
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The process of a power reactor spent fuel reprocessing pilot plant had been completed after many years of scientific research experiment and engineering construction, while the commissioning results indicated that some technological problems exist in a few key equipments, especially the shearing machine. As the main design basis it is needed to obtain accurate shear force so as to improve research of shearing machine. In this paper, shearing fracture process and shear force have been simulated in ABAQUS software, also a sample type of shear machine has been realize to do the shearing experiment. Through the study, simulation result is verified reliable comparing with the experimental data and the research method using computer simulation combined with experiment verification is significance to the research of spent fuel reprocessing.
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FAGHIHI, F., and M. R. NEMATOLLAHI. "NUCLEAR FUEL DEPLETION ANALYSIS USING MATLAB SOFTWARE." International Journal of Modern Physics C 17, no. 06 (June 2006): 805–15. http://dx.doi.org/10.1142/s0129183106009229.
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Coupled first order IVPs are frequently used in many parts of engineering and sciences. In this article, we presented a code including three computer programs which are joint with the Matlab software to solve and plot the solutions of the first order coupled stiff or non-stiff IVPs. Some engineering and scientific problems related to IVPs are given and fuel depletion (production of the 239 Pu isotope) in a Pressurized Water Nuclear Reactor (PWR) are computed by the present code.
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Ritter, Guillaume, Romain Eschbach, Richard Girieud, and Maxime Soulard. "CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning." EPJ Nuclear Sciences & Technologies 4 (2018): 10. http://dx.doi.org/10.1051/epjn/2018008.
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CESAR stands in French for “simplified depletion applied to reprocessing”. The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ∼400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture, CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with “industrial nuclear” constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR’s). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA − Cadarache.
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Seydaliev, M., and D. Caswell. "CORBA AND MPI-BASED “BACKBONE” FOR COUPLING ADVANCED SIMULATION TOOLS." AECL Nuclear Review 3, no. 2 (December 1, 2014): 83–90. http://dx.doi.org/10.12943/anr.2014.00036.
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There is a growing international interest in using coupled, multidisciplinary computer simulations for a variety of purposes, including nuclear reactor safety analysis. Reactor behaviour can be modeled using a suite of computer programs simulating phenomena or predicting parameters that can be categorized into disciplines such as Thermalhydraulics, Neutronics, Fuel, Fuel Channels, Fission Product Release and Transport, Containment and Atmospheric Dispersion, and Severe Accident Analysis. Traditionally, simulations used for safety analysis individually addressed only the behaviour within a single discipline, based upon static input data from other simulation programs. The limitation of using a suite of stand-alone simulations is that phenomenological interdependencies or temporal feedback between the parameters calculated within individual simulations cannot be adequately captured. To remove this shortcoming, multiple computer simulations for different disciplines must exchange data during runtime to address these interdependencies. This article describes the concept of a new framework, which we refer to as the “Backbone,” to provide the necessary runtime exchange of data. The Backbone, currently under development at AECL for a preliminary feasibility study, is a hybrid design using features taken from the Common Object Request Broker Architecture (CORBA), a standard defined by the Object Management Group, and the Message Passing Interface (MPI), a standard developed by a group of researchers from academia and industry. Both have well-tested and efficient implementations, including some that are freely available under the GNU public licenses. The CORBA component enables individual programs written in different languages and running on different platforms within a network to exchange data with each other, thus behaving like a single application. MPI provides the process-to-process intercommunication between these programs. This paper outlines the different CORBA and MPI configurations examined to date, as well as the preliminary configuration selected for coupling 2 existing safety analysis programs used for modeling thermal–mechanical fuel behavior and fission product behavior respectively. In addition, preliminary work in hosting both the Backbone and the associated safety analysis programs in a cluster environment are discussed.
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Okunev, V. S. "Fundamentally New Composite Materials of Fast Reactors Made on the Basis of Nanotechnology." Key Engineering Materials 887 (May 2021): 159–64. http://dx.doi.org/10.4028/www.scientific.net/kem.887.159.
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The main goal of the work is to identify the advantages of fast reactors when using nanotechnology in the manufacture of core materials. The research methods are based on the adaptation of known technologies (including powder metallurgy) to the design of fast reactors and on the numerical simulation of physical processes carried out using computer programs for the analysis of emergency conditions of fast reactors (including anticipated transient without scram - ATWS). The results of the research show that the use of structural materials based on steels hardened by nanooxides in combination with fundamentally new types of fuel based on composite materials can significantly improve the safety of nuclear technics. Sintered mixtures of ceramic microgranules (oxide, nitride) and nanoadditives of metallic beryllium or uranium are considered as nuclear fuel. Such composite nuclear fuel improves reactor safety and power. The following types of composite fuel were analyzed: mixed oxide with additives of a beryllium or uranium nanopowder, mixed mononitride with additives of a beryllium or uranium nanopowder. Most preferably, a ceramic-metal pellet fuel based on mononitride microgranules and uranium metal nanopowder. The use of such fuel (with a volume fraction of metallic uranium up to 20%) significantly increases the safety of the reactor, combining the advantages of metal and ceramics and completely neutralizing their disadvantages. The proposed materials are of practical importance in the development of new concepts of nuclear technics, in the transition to large-scale nuclear power and high-power reactors. The use of a new cermet-based composite fuel increases the power of the reactor and significantly increases the safety of the reactor.
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Jernkvist, Lars O. "Modelling of fine fragmentation and fission gas release of UO2 fuel in accident conditions." EPJ Nuclear Sciences & Technologies 5 (2019): 11. http://dx.doi.org/10.1051/epjn/2019030.
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In reactor accidents that involve rapid overheating of oxide fuel, overpressurization of gas-filled bubbles and pores may lead to rupture of these cavities, fine fragmentation of the fuel material, and burst-type release of the cavity gas. Analytical rupture criteria for various types of cavities exist, but application of these criteria requires that microstructural characteristics of the fuel, such as cavity size, shape and number density, are known together with the gas content of the cavities. In this paper, we integrate rupture criteria for two kinds of cavities with models that calculate the aforementioned parameters in UO2 LWR fuel for a given operating history. The models are intended for implementation in engineering type computer programs for thermal-mechanical analyses of LWR fuel rods. Here, they have been implemented in the FRAPCON and FRAPTRAN programs and validated against experiments that simulate LOCA and RIA conditions. The capabilities and shortcomings of the proposed models are discussed in light of selected results from this validation. Calculated results suggest that the extent of fuel fragmentation and transient fission gas release depends strongly on the pre-accident fuel microstructure and fission gas distribution, but also on rapid changes in the external pressure exerted on the fuel pellets during the accident.
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Shahabuddin, Syed, and Sayeda Daud. "http://habibiaislamicus.com/index.php/hirj/article/view/181." Habibia Islamicus 4, no. 2 (December 27, 2020): 43–56. http://dx.doi.org/10.47720/hi.2020.0402e07.
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Peaceful atomic production is generally separated into five fundamental territories: extraction and preparing of atomic crude materials, the creation of advanced uranium, generation of atomic fuel components, development and task of atomic reactors and reprocessing of energizes. Notwithstanding utilizing atomic vitality to produce power from control reactors, it is likewise generally utilized in agribusiness, prescription, industry, science and hydrology. Power is right now produced by the iota in created nations. Different nations are additionally attempting endeavors to create power from modest atomic power plants, which utilize little amounts of radioactive substances to deliver power on a substantial scale. Radiation from atomic reactors is at times used to regard sicknesses, for example, tumor. These beams are likewise used to eliminate germs and destructive creepy crawlies in farmland and nourishment items. In the atomic division, organizations or associations can work in various parts of the Muslim world. Pakistan facing difficulty to keep nuclear plant powered due to shortage of nuclear electricity and other challenges are concerns are regarding its security and protection from natural disasters due to climate change which would triple energy consumption of nuclear plant. Although generating electricity from nuclear energy is cost effective as compared to generating electricity from oil and gas or coal, and nuclear radiation can be used to diagnose, treat critical diseases like cancer. Most importantly it can be used in neuropsychiatry to produce brain imaging using gamma camera and single photon emission computed tomography to diagnose depression, anxiety and stress. Pakistan has potential to facilitate its nation and can avail benefit by seeking nuclear technology for mass development programs. The objective of the study is to discuss the Civil Nuclear institute of Technology and what are their contributions and its scope in Pakistan. Atomic researchers can progress toward becoming individuals from these social orders to cooperate on the tranquil utilization of atomic vitality. The researchers should make the best utilization of atomic vitality for vitality purposes to serve all in Pakistan which is begin highlighted the areas in this research and concluded Pakistan facing difficulty to keep nuclear plant powered due to shortage of nuclear electricity and other challenges are concerns are regarding its security and protection from natural disasters due to climate change which would triple energy consumption of nuclear plant. Although generating electricity from nuclear energy is cost effective as compared to generating electricity from oil and gas or coal, and nuclear radiation can be used to diagnose, treat critical diseases like cancer. Most importantly it can be used in neuropsychiatry to produce brain imaging using gamma camera and single photon emission computed tomography to diagnose depression, anxiety and stress. It is concluded that Pakistan has effectively connected with atomic science and technology to innovate and help the society while it is improving as well for implementation as well.
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Massih, A. R., and Lars O. Jernkvist. "Solid state phase transformation kinetics in Zr-base alloys." Scientific Reports 11, no. 1 (March 29, 2021). http://dx.doi.org/10.1038/s41598-021-86308-w.
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AbstractWe present a kinetic model for solid state phase transformation ($$\alpha \rightleftharpoons \beta$$ α ⇌ β ) of common zirconium alloys used as fuel cladding material in light water reactors. The model computes the relative amounts of $$\beta$$ β or $$\alpha$$ α phase fraction as a function of time or temperature in the alloys. The model accounts for the influence of excess oxygen (due to oxidation) and hydrogen concentration (due to hydrogen pickup) on phase transformation kinetics. Two variants of the model denoted by A and B are presented. Model A is suitable for simulation of laboratory experiments in which the heating/cooling rate is constant and is prescribed. Model B is more generic. We compare the results of our model computations, for both A and B variants, with accessible experimental data reported in the literature covering heating/cooling rates of up to 100 K/s. The results of our comparison are satisfactory, especially for model A. Our model B is intended for implementation in fuel rod behavior computer programs, applicable to a reactor accident situation, in which the Zr-based fuel cladding may go through $$\alpha \rightleftharpoons \beta$$ α ⇌ β phase transformation.
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Hittner, Dominique, Carmen Angulo, Virginie Basini, Edgar Bogusch, Eric Breuil, Derek Buckthorpe, Vincent Chauvet, et al. "HTR-TN Achievements and Prospects for Future Developments." Journal of Engineering for Gas Turbines and Power 133, no. 6 (February 14, 2011). http://dx.doi.org/10.1115/1.4000799.
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It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the “Strategic Energy Technology Plan (SET-Plan)” recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to industrial process heat applications addressing (1) the development of a flexible HTR that can be coupled to many different process heat and cogeneration applications with very versatile requirements, (2) the development of coupling technologies for such coupling, (3) the possible adaptations of process heat applications required for nuclear coupling, and (4) the integration and optimization of the whole coupled system. As a preliminary step for this ambitious program, HTR-TN endeavors to create a strategic partnership between nuclear industry and R&D and process heat user industries.
Dissertations / Theses on the topic "Reactor fuel reprocessing – Computer programs"
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Monger, Fred A. "KSIG - Kansas State University isotope generation microcomputer program." 1985. http://hdl.handle.net/2097/27506.
Full textBooks on the topic "Reactor fuel reprocessing – Computer programs"
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Klein, M. E. Simulation of in-reactor experiments with the ELOCA.Mk5 code. Chalk River, Ont: Fuel Engineering Branch, Chalk River Laboratories, 1994.
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United States. Congress. Senate. Committee on Environment and Public Works. Subcommittee on Clean Air, Climate Change, and Nuclear Safety. Oversight on NRC's regulatory responsibilities and capabilities for long- and short-term spent fuel storage programs: Hearing before the Subcommittee on Clean Air, Climate Change, and Nuclear Safety of the Committee on Environment and Public Works, United States Senate, One Hundred Ninth Congress, second session, September 14, 2006. Washington: U.S. G.P.O., 2009.
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Oversight on NRC's regulatory responsibilities and capabilities for long- and short-term spent fuel storage programs: Hearing before the Subcommittee on Clean Air, Climate Change, and Nuclear Safety of the Committee on Environment and Public Works, United States Senate, One Hundred Ninth Congress, second session, September 14, 2006. Washington: U.S. G.P.O., 2009.
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Bilanovic, Z. Neutron-photon energy deposition in CANDU reactor fuel channels: A comparison of modelling techniques using ANISN and MCNP computer codes. Chalk River, Ont: System Chemistry and Corrosion Branch, Chalk River Laboratories, 1994.
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F, Chen T., Lawrence Livermore National Laboratory, and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research., eds. CASKS (Computer Analysis of Storage Casks): A microcomputer based analysis system for storage cask design review : user's manual to version 1b (including program reference). Washington, D.C: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1995.
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K, Mikitiouk, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research, Institut I︠A︡dernykh Reaktorov (Rossiĭskiĭ nauchnyĭ t︠s︡entr "Kurchatovskiĭ institut"), and Institut problem bezopasnogo ispolʹzovanii︠a︡ i︠a︡dernoĭ ėnergii (Rossiĭskiĭ nauchnyĭ t︠s︡entr "Kurchatovskiĭ institut"), eds. Modification of IPSN's SCANAIR fuel rod transient code for high burnup VVER fuel. Washington, DC: Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1999.
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A, Shestopalov, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research, Institut problem bezopasnogo ispolʹzovanii︠a︡ i︠a︡dernoĭ ėnergii (Rossiĭskiĭ nauchnyĭ t︠s︡entr "Kurchatovskiĭ institut"), and Institut I︠A︡dernykh Reaktorov (Rossiĭskiĭ nauchnyĭ t︠s︡entr "Kurchatovskiĭ institut"), eds. Modification of USNRC's FRAP-T6 fuel rod transient code for high burnup VVER fuel. Washington, DC: Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1999.
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F, Young Michael, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Research., and Sandia National Laboratories, eds. Integrated fuel-coolant interaction (IFCI 6.0) code: User's manual. Washington, DC: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1994.
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F, Young Michael, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Research., and Sandia National Laboratories, eds. Integrated fuel-coolant interaction (IFCI 6.0) code: User's manual. Washington, DC: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1994.
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F, Young Michael, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Research., and Sandia National Laboratories, eds. Integrated fuel-coolant interaction (IFCI 6.0) code: User's manual. Washington, DC: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1994.
Find full textBook chapters on the topic "Reactor fuel reprocessing – Computer programs"
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De Luca, Domenico, Simone Di Pasquale, Marco Cherubini, Alessandro Petruzzi, and Gianni Bruna. "International Benchmark Activity in the Field of Sodium Fast Reactors." In Recent Advances on Numerical Simulations [Working Title]. IntechOpen, 2021. http://dx.doi.org/10.5772/intechopen.97812.
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Global interest in fast reactors has been growing since their inception in 1960 because they can provide efficient, safe, and sustainable energy. Their closed fuel cycle can support long-term nuclear power development as part of the world’s future energy mix and decrease the burden of nuclear waste. In addition to current fast reactors construction projects, several countries are engaged in intense R&D and innovation programs for the development of innovative, or Generation IV, fast reactor concepts. Within this framework, NINE is very actively participating in various Coordinated Research Projects (CRPs) organized by the IAEA, aimed at improving Member States’ fast reactor analytical simulation capabilities and international qualification through code-to-code comparison, as well as experimental validation on mock-up experiment results of codes currently employed in the field of fast reactors. The first CRP was focused on the benchmark analysis of Experimental Breeder Reactor II (EBR-II) Shutdown Heat Removal Test (SHRT-17), protected loss-of-flow transient, which ended in the 2017 with the publication of the IAEA-TECDOC-1819. In the framework of this project, the NINE Validation Process– developed in the framework of NEMM (NINE Evaluation Model Methodology) – has been proposed and adopted by most of the organizations to support the interpretation of the results calculated by the CRP participants and the understanding of the reasons for differences between the participants’ simulation results and the experimental data. A second project regards the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the Fast Flux Test Facility (FFTF), the Loss of Flow Without Scram (LOFWOS) Test #13, started in 2018. A detailed nodalization has been developed by NINE following its nodalization techniques and the NINE validation procedure has been adopted to validate the Simulation Model (SM) against the experimental data of the selected test. The third activity deals with the neutronics benchmark of China Experimental Fast Reactor (CEFR) Start-Up Tests, a CRP proposed by the China Institute of Atomic Energy (CIAE) launched in 2018 the main objective of which is to improve the understanding of the start-up of a SFR and to validate the fast reactor analysis computer codes against CEFR experimental data. A series of start-up tests have been analyzed in this benchmark and NINE also proposed and organized a further work package focused on the sensitivity and uncertainty analysis of the first criticality test. The present chapter intends to summarize the results achieved using the codes currently employed in the field of fast reactor in the framework of international projects and benchmarks in which NINE was involved and emphasize how the application of developed procedures allows to validate the SM results and validate the computer codes against experimental data.
Conference papers on the topic "Reactor fuel reprocessing – Computer programs"
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Quang, R. Do, V. Petitjean, F. Hollebecque, O. Pinet, T. Flament, and A. Prod’homme. "Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA’s Cold Crucible Melter." In ASME 2003 9th International Conference on Radioactive Waste Management and Environmental Remediation. ASMEDC, 2003. http://dx.doi.org/10.1115/icem2003-4594.
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The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA’s R&D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a soldified glass layer that protects the melter’s inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities: the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybednum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification cell. This paper will present the results obtained in the framework of these qualification programs.
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Shropshire, David E. "Lessons Learned From GEN I Carbon Dioxide Cooled Reactors." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49380.
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This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.
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Wren, D. J., N. Popov, and V. G. Snell. "Validation of Physics and Thermalhydraulics Computer Codes for Advanced CANDU Reactor Applications." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49102.
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Atomic Energy of Canada Ltd. (AECL) is developing an Advanced CANDU Reactor (ACR) that is an evolutionary advancement of the currently operating CANDU 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular CANDU concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in CANDU 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the CANDU 6, most of the key phenomena associated with the safety analyses of ACR are common, and the CANDU industry standard toolset of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design features that have impact on the physics and thermalhydraulics codes.
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Doucet, M., L. Durand Terrasson, and J. Mouton. "Safety Criticality Standards Using the French CRISTAL Code Package: Application to the AREVA NP UO2 Fuel Fabrication Plant." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89017.
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Criticality safety evaluations implement requirements to proof of sufficient sub critical margins outside of the reactor environment for example in fuel fabrication plants. Basic criticality data (i.e., criticality standards) are used in the determination of sub critical margins for all processes involving plutonium or enriched uranium. There are several criticality international standards, e.g., ARH-600, which is one the US nuclear industry relies on. The French Nuclear Safety Authority (DGSNR and its advising body IRSN) has requested AREVA NP to review the criticality standards used for the evaluation of its Low Enriched Uranium fuel fabrication plants with CRISTAL V0, the recently updated French criticality evaluation package. Criticality safety is a concern for every phase of the fabrication process including UF6 cylinder storage, UF6-UO2 conversion, powder storage, pelletizing, rod loading, assembly fabrication, and assembly transportation. Until 2003, the accepted criticality standards were based on the French CEA work performed in the late seventies with the APOLLO1 cell/assembly computer code. APOLLO1 is a spectral code, used for evaluating the basic characteristics of fuel assemblies for reactor physics applications, which has been enhanced to perform criticality safety calculations. Throughout the years, CRISTAL, starting with APOLLO1 and MORET 3 (a 3D Monte Carlo code), has been improved to account for the growth of its qualification database and for increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating a modern basic microscopic cross section set based on JEF2.2 and the comprehensive APOLLO2 and MORET 4 codes. APOLLO2 is well suited for criticality standards calculations as it includes a sophisticated self shielding approach, a Pij flux determination, and a 1D transport (Sn) process. CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and the implementation of user-friendly graphical interfaces. Due to its comprehensive physical simulation and thanks to its broad qualification database with more than a thousand benchmark/calculation comparisons, CRISTAL V0 provides outstanding and reliable accuracy for criticality evaluations for configurations covering the entire fuel cycle (i.e. from enrichment, pellet/assembly fabrication, transportation, to fuel reprocessing). After a brief description of the calculation scheme and the physics algorithms used in this code package, results for the various fissile media encountered in a UO2 fuel fabrication plant will be detailed and discussed.
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Iversen, Klaus, Seppo Salmenhaara, Steinar Backe, Anna Cato, Staffan Lindskog, Clas Callander, Henrik Efraimsson, Inga Andersson, and Rolf Sjo¨blom. "Cost Calculations at Early Stages of Nuclear Research Facilities in the Nordic Countries." In The 11th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2007. http://dx.doi.org/10.1115/icem2007-7168.
Full textAbstract:
The Nordic countries Denmark, Norway and Sweden, and to some extent also Finland, had very large nuclear research and development programs for a few decades starting in the nineteen fifties. Today, only some of the facilities are in use. Some have been decommissioned and dismantled while others are at various stages of planning for shutdown. The perspective ranges from imminent to several decades. It eventually became realized that considerable planning for the future decommissioning is warranted and that an integral part of this planning is financial, including how financial funds should be acquired, used and allocated over time. This necessitates that accurate and reliable cost estimates be obtained at all stages. However, this is associated with fundamental difficulties and treacherous complexities, especially for the early ones. Eventually, Denmark and Norway decided not to build any nuclear power plants while Finland and Sweden did. This is reflected in the financing where the latter countries have established systems with special funds in which money is being collected now to cover the future costs for the decommissioning of the research facilities. Nonetheless, the needs for planning for the decommissioning of nuclear research facilities are very similar. However, they differ considerably from those of nuclear power reactors, especially with regard to cost calculations. It has become apparent in the course of work that summation types of cost estimation methodologies give rise to large systematic errors if applied at early stages, in which case comparison based assessments are less biased and may be more reliable. Therefore, in order to achieve the required quality of the cost calculations, it is necessary that data and experience from authentic cases be utilized in models for cost calculations. It also implies that this calculation process should include a well adopted learning process. Thus, a Nordic co-operation has been established for the exchange and evaluation of cost-related information on nuclear research facilities. The aim is to identify good practices, accumulate experience, compile data from actual plants and projects, and to derive methodology for cost calculations, especially for early stages. The work includes the following tasks which constitutes the bulk of the present paper: • identification of good practice with regard to the following: - strategy and planning, - methodology selection, - radiological surveying, - uncertainty analysis; • descriptions of relevant plants, features and projects: - decommissioning of reactor DR 1 in Denmark, - decommissioning of reactor R 1 in Sweden, - decommissioning of the pilot scale uranium fuel reprocessing plant in Norway, - planning for the future decommissioning of the TRIGA reactor in Finland; • techniques for assessments of costs introduction.
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Oskolkov, Borys Ya, Yuri A. Neretin, Valeryi P. Saliy, Valeryi A. Seyda, and Vyascheslav V. Fomin. "Problems of Radioactive Waste Management at Chornobyl Nuclear Power Plant (ChNPP)." In ASME 2003 9th International Conference on Radioactive Waste Management and Environmental Remediation. ASMEDC, 2003. http://dx.doi.org/10.1115/icem2003-4827.
Full textAbstract:
According to the assessments the overall amount of radioactive waste (RAW) to be reprocessed and buried at the Chornobyl NPP site amounts to 1,696,738 m3 (without regard for reactor metal structures, dismantling of building structures and plan constructions, and the Unit Shelter building). The overall activity of radioactive waste are evaluated at 1,640,504.64 ΤBq. The RAW management activities are implemented at the Chornoby1 NPP within the frame of several programs of different hierarchy including the State Comprehensive Program for RAW Management in Ukraine, Integrated Program of RAW Management at the Chornoby1 NPP Shutdown Stage and Transformation of the Unit Shelter into an Ecologically Safe System. At the present time a number of key RAW management facilities are being constructed within the frame of the international aid to Ukraine. They are the Liquid Radioactive Waste Plant and Solid Radioactive Waste Reprocessing Complex. As of now, the issues concerning RAW utilization at the Unit Shelter are resolved at a conceptual level. There has not practical decision in relation to a geologic facility. The complexity and scale of ChNPP RAW management problems will require significant efforts of both Ukraine and the whole world community to solve these issues. The task related to removal and final burial of accumulated and generating radioactive waste is one of the main aspects of decommissioning activities at any nuclear power plant. RAW management work is the most important and complicated work performed at the Chornoby1 NPP. The specific features of ChNPP RAW management are as follows: • Variety of RAW generation sources, their types, physical and chemical properties. • Large amount of radioactive wastes which already exist and those generated in the decommissioning process. • Presence of disorganized RAW characterized by wide spatial distribution within the Unit Shelter and at the plant site. • Need to apply a very wide spectrum of various RAW management techniques depending on their location and type. • Need in developing unique techniques to manage special types of RAW located at the site (fuel containing masses of the Unit Shelter). • Large amount and variety of facilities required for RAW final storage. • Absence of reliable and serviceable instrumental procedures and necessary equipment to define RAW properties for RAW separation and classification. • Ecological peculiarities of RAW management within the Chornoby1 zone. • Multiphase decontamination and restoration processes resulting in RAW formation. • Need in integrating RAW management problems at the ChNPP and within the Chornoby1 Exclusion Zone taken as whole. • Long time period required for implementing the whole program of RAW management at the ChNPP. • Large quantity of people involved in RAW management process (local and foreign participants, different organization operated by various departments).
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Mesina, George L. "Visualization of Heat Transfer and Core Damage With RGUI 1.5." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22698.
Full textAbstract:
Graphical User Interfaces (GUI) have become an integral and essential part of computer software. In the ever-changing world of computing, they provide the user with a valuable means to learn, understand, and use the application software while also helping applications adapt to and span different computing paradigms, such as different operating systems. For these reasons, GUI development for nuclear plant analysis programs has been ongoing for a decade and a half and much progress has been made. With the development of codes such as RELAP5-3D [1] and SCDAP/RELAP5 that have multi-dimensional modeling capability, it has become necessary to represent three-dimensional, calculated data. The RELAP5-3D Graphical User Interface (RGUI) [4] was designed specifically for this purpose. It reduces the difficulty of analyzing complex three-dimensional models and enhances the analysts’ ability to recognize plant behavior visually. Previous versions of RGUI [5] focused on visualizing reactor coolant behavior during a simulated transient or accident. Recent work has extended RGUI to display two other phenomena, heat transfer and core damage. Heat transfer is depicted through the visualization of RELAP5-3D heat structures. Core damage is visualized by displaying fuel rods and other core structures in a reactor vessel screen. Conditions within the core are displayed via numerical results and color maps. These new features of RGUI 1.5 are described and illustrated.
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Kawada, Ken-ichi, Ikken Sato, Yoshiharu Tobita, Werner Pfrang, Laurence Buffe, and Emmanuelle Dufour. "Development of PIRT (Phenomena Identification and Ranking Table) for SAS-SFR (SAS4A) Validation." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30679.
Full textAbstract:
SAS-SFR (derived from SAS4A) is presently the most advanced computer code for simulation of the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors (SFR). In the past two decades, intensive model improvement works have been conducted for SAS-SFR utilizing the experimental data from the CABRI programs. The main target of the present work is to confirm validity of these improved models through a systematic and comprehensive set of test analyses to demonstrate that the improved models has a sufficient quality assurance level for applications to reactor conditions. In order to reach these objectives, an approach of PIRT (Phenomena Identification and Ranking Table) on a set of accident scenarios has been applied. Based on the fact that there have been a significant amount of validation studies for decades, development of the code validation matrix concentrated on key issues. Different accident scenarios have been chosen for the PIRT considering typical SFR accident transients that address a large range of phenomena. As the most important and typical Core Disruptive Accident scenarios leading to generalized core melting and to be addressed with SAS-SFR in the present study, ULOF (Unprotected Loss Of Flow), UTOP (Unprotected Transient OverPower) and ULOHS (Unprotected Loss Of Heat Sink) are selected. The PIRT process applied to a given accident scenario consists in an identification of the phenomena involved during the accident, the evaluation of the importance of the phenomena regarding to the evolution and consequences, and the evaluation of the status of knowledge based on the review of available experimental results. The identified phenomena involved in ULOF are explained as follows for the primary phase. Starting from initiating events, a loss of grid power leading to flow coast down without scram is assumed. The scenario up to coolant boiling is the main point within the first part of the ULOF phenomenological chart. Those elements related to reactivity feedback, such as heat up of coolant, fuel and various structures and their deformation due to the thermal transient are picked up. Depending on the time scale before boiling starts, primary, secondary and tertiary loop heat transfer including the DHR (Decay Heat Removal) system response is concerned since it defines the core inlet coolant temperature. Core inlet coolant temperature gives direct impact on the thermal condition of the core. It also affects reactivity through thermal expansion of the grid plate. In the second part of the ULOF phenomenological chart, elements such as coolant boiling, mechanical response of the fuel pin leading to cladding failure, FCI (Fuel-Coolant Interaction) and post-failure material relocation are picked up. This part of the chart is basically common to the ULOHS. Respective identified phenomena are to be simulated in the SAS-SFR code. To validate the function of the models in the code, ten high priority CABRI experiments are selected. Validation studies on these tests are underway. With the present study, important phenomena involved in ULOF, UTOP and ULOHS were identified and an evaluation matrix for the selected CABRI experiments was developed.