scholarly journals Concept of a fast breeder reactor to transmute MAs and LLFPs

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Radioactive Waste ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Fission Products ◽  
Fast Breeder Reactor ◽  
Nuclear Power Reactor ◽  
The Core ◽  
Effective Use

AbstractThe long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MAs: 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFPs: 129I, 99Tc, 79Se, etc.). The purpose of this study was to show a concept that can simultaneously achieve the breeding of fissile materials and the transmutation of MAs and LLFPs in one fast reactor. Transmutation was carried out by loading innovative Duplex-type MA fuel in the core region and LLFP-containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of approximately 1.1 was obtained, and MAs and LLFPs achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237Np, 14.1%/y for 241Am, 9.9%/y for 243Am, 1.6%/y for 129I, 0.75%/y for 99Tc, and 4%/y for 79Se. By simultaneously breeding fissile materials and transmuting MAs and LLFPs in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

scholarly journals Concept of a Fast Breeder Reactor To Transmute MA and LLFP

2021 ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Radioactive Waste ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Fission Products ◽  
Fast Breeder Reactor ◽  
Nuclear Power Reactor ◽  
The Core ◽  
Effective Use

Abstract The long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MA: 237 Np, 241 Am, 243 Am, etc.) and long-lived fission products (LLFP: 129 I, 99 Tc, 79 Se, etc.). The purpose of this study was to show a system that can simultaneously achieve the breeding of fissile materials and the transmutation of MA and LLFP in one fast reactor. Transmutation was carried out by loading innovative Duplex type MA fuel in the core region and LLFP containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of about 1.1 was obtained, and MA and LLFP achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237 Np, 14.1%/y for 241 Am, 9.9%/y for 243 Am, 1.6%/y for 129 I, 0.75%/y for 99 Tc, and 4%/y for 79 Se. By simultaneously breeding fissile materials and transmuting MA and LLFP in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

Void Reactivity Effect Research on Liquid Metal Cooled Fast Reactor

2017 ◽  
Author(s):  
Yang Lyu ◽  
Xiao Liang
Keyword(s):  
Liquid Metal ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Reactor Core ◽  
Nuclear Industry ◽  
Fourth Generation ◽  
Reactivity Effect ◽  
The Core ◽  
Reactivity Coefficient

In the fourth generation of advanced nuclear power systems, the liquid metal cooled fast reactor plays a more and more important role, such as SFR, LFR and ADS system with LBE coolant. Void reactivity effect means bubbles produced in the core area will induce the change of reactivity. And this reactivity will affect the safety of the reactor. Through investigation and comparison of several liquid metal cooled fast reactors in the nuclear industry, this paper studies bubbles in different positions and partial voiding of the active zone inside the core and fuel assemblies with Monte Carlo core physics calculation method and then concludes the main influencing factors of void reactivity coefficient. The results can provide reference for the follow-up research and development of new type liquid metal fast reactor core design.

Physics Characteristics of Nuclear Power Systems with Reduced Long-Term Radioactivity Risk

1995 ◽  
Vol 120 (1) ◽  
pp. 18-39 ◽  
Author(s):  
M. Salvatores ◽  
I. Slessarev ◽  
M. Uematsu
Keyword(s):  
Power Systems ◽  
Nuclear Power

scholarly journals IRRADIATION CHARACTERISTIC OF NATURAL UO2 PIN PHWR TARGET AT PRTF OF RSG – GAS CORE

2017 ◽  
Vol 19 (2) ◽  
pp. 71
Author(s):  
Jati Susilo ◽  
Tagor Malem Sembiring ◽  
Winter Dewayatna
Keyword(s):  
Neutron Flux ◽  
Nuclear Power ◽  
Irradiation Time ◽  
Thermal Power ◽  
Analytical Tool ◽  
Test Facility ◽  
Nuclear Power Reactor ◽  
Ramp Test ◽  
The Core ◽  
Fuel Pin

The RSG-GAS reactor has a facility for irradiation of the fuel pin of nuclear power reactor, namely Power Ramp Test Facility (PRTF). The in-house fabrication PWR fuel pin has prepared for irradiations in the PRTF facility, currently, while the various enrichments of uranium are analyzed using the analytical tool. In the next step, it is planned to perform an irradiation of PHWR fuel pin sample of natural UO2 in the facility. Before irradiation in the core, it should be analyzed by using the analytical tool. The objectives of this paper are to optimize irradiation time based on the burn-up, the generated linear power and the neutron flux level at the target. The 3-dimension calculations have been carried out by using the CITATION code in the SRAC2006 code system. Since the coolant of the reactor is H2O, the effect of moderators in the pressurized tube, H2O and D2O, were analyzed, as well as pellet radius and moderator densities. The calculation results show that the higher linear power as irradiation time longer is occurred preferably in the D2O moderator than in H2O. For the D2O moderator, the higher pressure affects the lower density and longer irradiation time. The maximum irradiation time for natural UO2 fuel pin with the pressurized D2O moderator is about 9.5×104 h, with the linear power of 700 W/cm. During irradiation, neutronic parameters of the core such as excess reactivity and ppf show a very small change, still far below design value.Keywords:  PHWR, Neutron Flux, Thermal Power, PRTF, RSG-GAS KARAKTERISTIK IRADIASI TARGET PIN PHWR UO2 ALAM PADA PRTF TERAS RSG – GAS. Teras RSG-GAS dilengkapi dengan fasilitas untuk uji iradiasi bahan bakar nuklir atau disebut dengan Power Ramp Test Fasility (PRTF). Saat ini sedang dilpersiapkan untuk dilakukan uji sample pin bahan bakar PWR pada fasilitas PRTF. Analisis terhadap uji iradiasi sample pellet UO2 dengan berbagai pengkayaan telah dilakukan menggunakan paket program komputer. Dimasa yang akan datang, uji iradiasi pin bahan bakar PHWR UO2 alam juga sedang dalam perencanaan. Sebelum diiradiasi di dalam teras, maka terlebih dahulu harus dilakukan analisis dengan menggunakan paket program komputer. Tujuan dari penelitian ini adalah optimasi uji iradiasi pin bahan bakar UO2 alam sebagai fungsi waktu iradiasi berdasarkan burn-up, daya linier dan fluks neutron. Perhitungan teras RSG-GAS dilakukan dengan paket program SRAC2006 modul CITATION dalam bentuk geometri 3 dimensi. Analisis dilakukan terhadap pengaruh penggunaan jenis moderator pada tabung tekan iradiasi (H2O dan D2O), perubahan ukuran pelllet UO2 dan perubahan besarnya densitas moderator D2O. Dari analisis hasil perhitungan diketahui bahwa semakin lama waktu iradiasi akan menghasilkan daya termal yang semakin besar jika menggunakan moderator D2O dibandingkan H2O. Semakin tinggi tekanan atau semakin kecil densitas moderator, maka akan menghasilkan daya termal yang semakin besar seiring bertambah lamanya waktu iradiasi. Batas maksimal waktu iradiasi untuk pin bahan bakar UO2 alam dengan moderator D2O bertekanan adalah sekitar 9,5×104 jam, dengan batasan daya linier desain kemampuan peralatan, 700 W/cm. Selama iradiasi, nilai parameter neutronik teras reaktor seperti reaktivitas lebih dan ppf hanya menunjukkan perubahan yang sangat kecil, masih jauh dibawah batas yang ditetapkan dalam desain.Kata kunci: PHWR, Fluks Neutron, Daya Termal, PRTF, RSG-GAS

scholarly journals Assessment of Permian Zubers as the Host Rock for Deep Geological Disposal

Energies ◽  
2020 ◽  
Vol 13 (9) ◽  
pp. 2239
Author(s):  
Leszek Lankof
Keyword(s):  
Radioactive Waste ◽  
Rock Salt ◽  
Host Rock ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Renewable Energy Sources ◽  
Geological Disposal ◽  
Lithostratigraphic Units ◽  
Deep Geological Disposal

Together with renewable energy sources, nuclear power represents an important contribution to a sustainable energy mix in many countries and has an important impact on sustainable development. Nuclear energy production is also a source of high-level radioactive waste (HLW) and spent nuclear fuel (SNF), which require special concern. Disposal in deep geological formations is one of the solutions for the long-term management of HLW and SNF. It requires the development of a concept ensuring long-term safe isolation of waste and its validation applying the safety case methodology, which is a formal compilation of evidence, analyses and arguments that quantify and justify a claim that the repository will be safe. The results of laboratory testing of a potential repository host rock are an important component of the evidence that helps in the safety assessment of the deep geological disposal concept. This paper presents results of research focused on the physical, geomechanical and sorption properties of the Brown and Red Zuber unit rocks from the Kłodawa Salt Mine in Poland, which together with rock salt are an important component of Polish salt domes. The Brown and Red Zubers are typical evaporite lithostratigraphic units for the Polish part of the Zechstein Basin. They consist of halite (15–85%) and water-insoluble minerals, such as anhydrite, clay minerals, carbonates, quartz and feldspar, which occurred in varying proportions in the tested samples. The properties of the zuber rocks have been compared with those of rock salt, which is considered a suitable host rock for deep geological disposal of radioactive waste.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 454-469
Author(s):  
S. H. Abdel-Latif
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Fission Products ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Potential Core ◽  
Safety Requirements ◽  
The Core ◽  
Active Safety Systems ◽  
Reactor Pressure

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

Long-term safety of the extended SFR - Methodology and conclusions from the SR-PSU project

MRS Advances ◽  
2016 ◽  
Vol 1 (61) ◽  
pp. 4075-4080
Author(s):  
Fredrik Vahlund
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Surface System ◽  
Radiological Safety ◽  
The Impact

ABSTRACTSince 1988 the Swedish Nuclear Fuel and Waste Management Co. operates a repository for low- and intermediate-level short-lived radioactive waste, SFR, in Forsmark, Sweden. Due to decommissioning of the nuclear power plants additional storage capacity is needed. In December 2014, an application to extend the repository was therefore submitted. One key component of this application was an assessment of post-closure safety of the extended SFR. For this safety assessment, a methodology based on that developed by SKB for the spent nuclear fuel repository was used and the impact of the degradation of repository components, the evolution of the surface system and changes of future climate on the radiological safety of the repository was assessed over a period of 100,000 years. The central conclusion of the SR-PSU safety assessment is that the extended SFR repository meets requirements on protection of human health and of the environment that have been established by the Swedish radiation safety authority for the final disposal of radioactive waste. Furthermore, the design of the repository was shown suitable for the waste selected and the applied methodology suitable for the safety assessment.

Numerical Simulation of the Behavior of Fission Products in the Primary Circuit of the VVER During the LOCA Severe Accident

2009 ◽  
Author(s):  
S. V. Tsaun ◽  
V. V. Bezlepkin ◽  
A. E. Kiselev ◽  
I. A. Potapov ◽  
V. F. Strizhov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Fission Products ◽  
Severe Accident ◽  
Primary Circuit ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Loss Of Coolant ◽  
Mass Activity

The methods and models for the analysis of the radiological consequences of the design basis and severe accidents in a Nuclear Power Plant (NPP) are presented in this paper when using the system code SOCRAT. The system code SOCRAT/V3 was elaborated for a realistic analysis of radiological consequences of severe accidents in a NPP. The following models of the fission products (FP) behavior are included into the code SOCRAT/V3: (i) the condensation and the evaporation of the FP in the gaseous phase and (ii) the sedimentation, the evaporation, and the coagulation of the aerosol-shape FP. The latter processes are governed by gravity, Brownian and turbulent diffusion, thermophoresis, turbophoresis and so forth. The behavior of the FP during the loss-of-coolant accidents (LOCA) is presented to demonstrate the possibilities of the code SOCRAT/V3. The main stages of the accident (the core dryout, the core reflooding, the core degradation, the hydrogen generation, the FP release, etc.) are described. Corresponding estimations of the mass, activity, and decay heat of the suspended, settled and released into containment the FP (Xe, Te, Cs, CsI, Cs2MoO4, and so forth) are represent as well.

Investigation of the Fission Product Release From Molten Pools Under Oxidizing Conditions With the Code RELOS

2006 ◽  
Author(s):  
Ingo D. Kleinhietpaß ◽  
Hermann Unger ◽  
Hermann-Josef Wagner ◽  
Marco K. Koch
Keyword(s):  
Phase Boundary ◽  
Molten Pool ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Energy Systems ◽  
The Core ◽  
Gas Atmosphere ◽  
Upper Pool

With the purpose of modeling and calculating the core behavior during severe accidents in nuclear power plants system codes are under development worldwide. Modeling of radionuclide release and transport in the case of beyond design basis accidents is an integrated feature of the deterministic safety analysis of nuclear power plants. Following a hypothetical, uncontrolled temperature escalation in the core of light water reactors, significant parts of the core structures may degrade and melt down under formation of molten pools, leading to an accumulation of large amounts of radioactive materials. The possible release of radionuclides from the molten pool provides a potential contribution to the aerosol source term in the late phase of core degradation accidents. The relevance of the amount of transferred oxygen from the gas atmosphere into the molten pool on the specification of a radionuclide and its release depends strongly on the initial oxygen inventory. Particularly for a low oxygen potential in the melt as it is the case for stratification when a metallic phase forms the upper layer and, respectively, when the oxidation has proceeded so far so that zirconium was completely oxidized, a significant influence of atmospheric oxygen on the specification and the release of some radionuclides has to be anticipated. The code RELOS (Release of Low Volatile Fission Products from Molten Surfaces) is under development at the Department of Energy Systems and Energy Economics (formerly Department of Nuclear and New Energy Systems) of the Ruhr-University Bochum. It is based on a mechanistic model to describe the diffusive and convective transport of fission products from the surface of a molten pool into a cooler gas atmosphere. This paper presents the code RELOS, i. e. the features and abilities of the latest code version V2.3 and the new model improvements of V2.4 and the calculated results evaluating the implemented models which deal with the oxygen transfer from the liquid side of the phase boundary to the bulk of the melt by diffusion or by taking into account natural convection. Both models help to estimate the amount of oxygen entering into the liquid upper pool volume and being available for the oxidation reaction. For both models the metallic, the oxidic and a mixture phase can be taken into account when defining the composition of the upper pool volume. The influence of crust formation, i. e. the decrease of the liquid pool surface area is taken care of because it yields the relevant amount of fission products released into the atmosphere. The difference of the partial density between the gas side of the phase boundary and the bulk of the gas phase is the driving force of mass transport.

scholarly journals PRE-TREATMENT OF BITUMINIZED NPP WASTES FOR DISPOSAL IN NEAR-SURFACE REPOSITORY

2019 ◽  
Vol 7 (2A) ◽  
Author(s):  
Vanessa Mota Vieira ◽  
Clédola Cássia Oliveira De Tello
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Transport Model ◽  
Technical Requirement ◽  
Cement Pastes ◽  
Waste Products ◽  
Near Surface ◽  
Evaluation Of Performance ◽  
Pre Treatment

The implementation of the national repository is an important technical requirement, and a legal requirement for the entry into operation of the nuclear power plant Angra 3. The Brazilian repository is being planned to be a near-surface one. In Brazil the low and intermediate level radioactive wastes are immobilized using cement and bitumen for Angra 1 and Angra 2 NPP, respectively. The main problems due to the disposal of bituminized wastes in repositories are swelling of the waste products and their degradation in the long term. To accommodate the swelling of the bituminized wastes, the drums are filled up to 70 - 90% of their volume, which reduces the structural the repository stability and the disposal availability. Countries, which use bitumen in the solidification of NPP´s radioactive waste and have near-surface repositories, need to immobilize this bituminized waste within other drums containing cement pastes or mortars to disposal them. This study aims to find solutions for the storage in surface repository of bituminized radioactive waste products, making them compatible with the acceptance criteria of cemented waste products. It was also performed a modeling with the results obtained in the leaching test using the ALT program and defined the transport model of the cesium leachate element and it was verified that in the early times the leaching was governed by the diffusion model and later by the partition model. The results obtained in this study can be used in the evaluation of performance of repositories.

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