scholarly journals Study of an ADS Loaded with Thorium and Reprocessed Fuel

2012 ◽  
Vol 2012 ◽  
pp. 1-12 ◽  
Author(s):  
Graiciany de Paula Barros ◽  
Claubia Pereira ◽  
Maria A. F. Veloso ◽  
Antonella L. Costa
Keyword(s):  
Nuclear Fuel ◽  
Monte Carlo Code ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Driven Systems ◽  
The Core ◽  
Combined Use ◽  
Driven System ◽  
Accelerator Driven System ◽  
Accelerator Driven Systems

Accelerator-driven systems (ADSs) are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods wasT232hO2forU233production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use ofT232hO2without the initial requirement ofU233enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations). The multiplication factor (keff) evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use ofT232hO2and reprocessed fuel allowedU233production without the initial requirement ofU233enrichment.

scholarly journals Time-Dependent Neutronic Analysis of a Power-Flattened Gas Cooled Accelerator Driven System Fuelled with Thorium, Uranium, Plutonium, and Curium Dioxides TRISO Particles

2016 ◽  
Vol 2016 ◽  
pp. 1-11 ◽  
Author(s):  
Gizem Bakır ◽  
Gamze Genç ◽  
Hüseyin Yapıcı
Keyword(s):  
Power Density ◽  
High Energy ◽  
Time Dependent ◽  
Monte Carlo Code ◽  
Spent Fuel ◽  
Driven System ◽  
Accelerator Driven System ◽  
Fission Power ◽  
Triso Particles ◽  
Fuel Particles

This study presents the power flattening and time-dependent neutronic analysis of a conceptual helium gas cooled Accelerator Driven System (ADS) loaded with TRISO (tristructural-isotropic) fuel particles. Target material is lead-bismuth eutectic (LBE). ThO2, UO2, PuO2, and CmO2TRISO particles are used as fuel. PuO2and CmO2fuels are extracted from PWR-MOX spent fuel. Subcritical core is radially divided into 10 equidistant subzones in order to flatten the power produced in the core. Tens of thousands of these TRISO fuel particles are embedded in the carbon matrix fuel pebbles as five different cases. The high-energy Monte Carlo code MCNPX 2.7 with the LA150 library is used for the neutronic calculations. Time-dependent burnup calculations are carried out for thermal fission power (Pth) of 1000 MW using the BURN card. The energy gain of the ADS is in the range of 99.98–148.64 at the beginning of a cycle. Furthermore, the peak-to-average fission power density ratio is obtained between 1.021 and 1.029 at the beginning of the cycle. These ratios show a good quasi-uniform power density for each case. Furthermore, up to 155.1 g233U and 103.6 g239Pu per day can be produced. The considered system has a high neutronic capability in terms of energy multiplication, fissile breeding, and spent fuel transmutation with thorium utilization.

scholarly journals Study of a conceptual accelerator driven system loaded with thorium dioxide mixed with transuranic dioxides in TRISO particles

2016 ◽  
Vol 31 (3) ◽  
pp. 197-206 ◽  
Author(s):  
Gizem Bakir ◽  
Gamze Genc ◽  
Huseyin Yapici
Keyword(s):  
Target Material ◽  
High Energy ◽  
Monte Carlo Code ◽  
Spent Fuel ◽  
Thorium Dioxide ◽  
Pure Lead ◽  
Driven System ◽  
Accelerator Driven System ◽  
Triso Particles ◽  
Fuel Particles

Nuclear spent fuel management is one of the top major subjects in the utilization of nuclear energy. Hence, solutions to this problem have been increasingly researched in recent years. The basic aim of this work is to examine the fissile breeding and transuranic fuel transmutation potentials of a gas cooled accelerator-driven system. In line with this purpose, firstly, the conceptually designed system is optimized by using several target materials and fuel mixtures, from the point of neutronic. Secondly, three different material compositions, namely, pure lead bismuth eutectic (LBE), LBE+natural UO2, and LBE+15 % enrichment UO2, are considered as target material. The target zone is separated to two sub-zones but as one within the other. The outer sub-zone is pure LBE target, and the inner sub-zone is either UO2 or pure LBE target. The UO2 target sub-zone is cooled with helium gas. Finally, the thorium dioxide mixed with transuranic dioxides, discharged from PWR-MOX spent fuel, in pebbles composed of graphite and TRISO-coated spherical fuel particles, is used for breeding fissile fuel and transmuting transuranic fuels. Three different thorium-transuranic mixtures, (Th, Pu)O2, (Th, Cm)O2, (Th, Pu, MA)O2, are examined with various mixture fractions. The packing fractions of the fuel pebbles in the transmutation core and the tristructural-isotropic coated fuel particles in a pebble are assumed as 60 % and 29 %, respectively. The transmutation core is also cooled with a high-temperature helium coolant. In order to produce high-flux neutrons that penetrate through the transmutation core, the target is exposed to the continuous beams of 1 GeV protons. The computations have been carried out with the high-energy Monte Carlo code MCNPX using the LA150 library. The numerical outcomes show that the examined accelerator-driven system has rather high neutronic data in terms of the energy production and fissile fuel breeding.

scholarly journals Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder

2020 ◽  
Vol 2020 ◽  
pp. 1-12
Author(s):  
Young-Hwan Kim ◽  
Yung-Zun Cho ◽  
Jin-Mok Hur
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Energy Research ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Compressive Stresses ◽  
Simulated Fuel ◽  
Ceramic Fuel ◽  
Experimental Approaches

We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

scholarly journals The Accelerator Driven Systems, a 21st Century Option for Closing Nuclear Fuel Cycles and Transmuting Minor Actinides

2021 ◽  
Vol 13 (22) ◽  
pp. 12643
Author(s):  
Hamid Aït Abderrahim ◽  
Michel Giot
Keyword(s):  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Building Blocks ◽  
Minor Actinides ◽  
Physical Treatment ◽  
Driven Systems ◽  
Fuel Cycles ◽  
Fuel Fabrication ◽  
Global Vision ◽  
Accelerator Driven Systems

Closing the nuclear fuel cycle and transmuting Minor Actinides (M.As) can be considered as an application of the duty of care principlel principle which says that, “before the final disposal of any waste, any possible chemical and/or physical treatment has to be applied in order to reduce the waste’s toxicity, provided the treatment does not convey unacceptable risks or unacceptable costs”. Forty years of complex research and development has shown that Accelerator Driven Systems could provide a solution to the challenge posed by spent nuclear fuels, by enabling the ability to considerably decrease their radiotoxicity lifetime burden and volume. In particular, a multilateral strategy of treatment of the MAs could be a commendable solution for both the countries phasing out the exploitation of nuclear energy and for those pursuing and developing this exploitation. The pre-industrial assessment of the technical and financial feasibility for industrialization is the next step. This applies to the four R&D and Demonstration building blocks: advanced separation, MAs’ loaded fuel fabrication, dedicated transmuters demonstration (MYRRHA) and provision for MAs’ fuel loaded processing. A global vision of the process leading to a sustainable option is proposed.

scholarly journals Subcritical Multiplication Parameters of the Accelerator-Driven System with 100 MeV Protons at the Kyoto University Critical Assembly

2012 ◽  
Vol 2012 ◽  
pp. 1-9 ◽  
Author(s):  
Jae-Yong Lim ◽  
Cheol Ho Pyeon ◽  
Takahiro Yagi ◽  
Tsuyoshi Misawa
Keyword(s):  
Reaction Rate ◽  
Reaction Rates ◽  
Relative Difference ◽  
Critical Assembly ◽  
The Core ◽  
Driven System ◽  
Accelerator Driven System ◽  
Special Mention ◽  
Basic Experiments ◽  
Mev Protons

Basic experiments on the accelerator-driven system (ADS) at the Kyoto University Critical Assembly are carried out by combining a solid-moderated and -reflected core with the fixed-field alternating gradient accelerator. The reaction rates are measured by the foil activation method to obtain the subcritical multiplication parameters. The numerical calculations are conducted with the use of MCNPX and JENDL/HE-2007 to evaluate the reaction rates of activation foils set in the core region and at the location of the target. Here, a comparison between the measured and calculated eigenvalues reveals a relative difference of around 10% in C/E values. A special mention is made of the fact that the reaction rate analyses in the subcritical systems demonstrate apparently the actual effect of moving the tungsten target into the core on neutron multiplication. A series of further ADS experiments with 100 MeV protons needs to be carried out to evaluate the accuracy of subcritical multiplication parameters.

scholarly journals Transmutations of Long-Lived and Medium-Lived Fission Products Extracted from CANDU and PWR Spent Fuels in an Accelerator-Driven System

2019 ◽  
Vol 2019 ◽  
pp. 1-13
Author(s):  
Alper Buğra Arslan ◽  
İlayda Yilmaz ◽  
Gizem Bakir ◽  
Hüseyin Yapici
Keyword(s):  
Fission Products ◽  
Time Dependent ◽  
Thermal Reactor ◽  
Beam Power ◽  
Spent Fuel ◽  
Driven System ◽  
Helium Gas ◽  
Computer Codes ◽  
Accelerator Driven System ◽  
Additional Fuel

This study presents the time-dependent analyses of transmutations of long-lived fission products (LLFPs) and medium-lived fission products (MLFPs) occurring in thermal reactors in a conceptual helium gas-cooled accelerator-driven system (ADS). In accordance with this purpose, the CANDU-37 and PWR 15 × 15 spent fuels are separately considered. The ADS consists of LBE-spallation neutron target, subcritical fuel zone, and graphite reflector zone. While the considered ADS is fueled with the spent nuclear fuels extracted from each thermal reactor without the use of additional fuel, fission products extracted from same thermal reactor are also placed into transmutation zone in graphite reflector zone. The LLFP transmutation performance of the modified ADS is analyzed by considering three different spent fuels extracted from the thermal reactors. Spent fuels are extracted from CANDU-37 in case A, from PWR-15 × 15 in case B, and from CANDU-37 fueled with mixture of PWR 15 × 15 spent fuel and 46% ThO2 in case C. The LBE target is bombard with protons of 1000 MeV. The proton beam power is assumed as 20 MW, which corresponds to 1.24828·1017 protons per second. MCNPX 2.7 and CINDER 90 computer codes are used for the time-dependent burn calculations. The ADS is operated under subcritical mode until the value of keff increases to 0.984, and the maximum operation times are obtained as 3400, 3270, and 5040 days according to the spent fuel cases of A, B, and C, respectively. The calculations bring out that in the modified ADS, LLFPs and MLFPs, which are extracted from thermal reactors, can be transformed to stable isotopes in significant amounts along with energy production.

scholarly journals The Three-­Dimensional Neutron-­Physical Model of Spent Nuclear Fuel Storage System

2021 ◽  
Vol 20 ◽  
pp. 51-59
Author(s):  
О. R. Trofymenko ◽  
◽  
І. M. Romanenko ◽  
М. І. Holiuk ◽  
C. V. Hrytsiuk ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Storage Facility ◽  
Spent Fuel ◽  
Wide Range ◽  
Fuel Assemblies ◽  
Fuel Storage

The management of spent nuclear fuel is one of the most pressing problems of Ukraine’s nuclear energy. To solve this problem, as well as to increase Ukraine’s energy independence, the construction of a centralized spent nuclear fuel storage facility is being completed in the Chornobyl exclusion zone, where the spent fuel of Khmelnytsky, Rivne and South Ukrainian nuclear power plants will be stored for the next 100 years. The technology of centralized storage of spent nuclear fuel is based on the storage of fuel assemblies in ventilated HI-STORM concrete containers manufactured by Holtec International. Long-term operation of a spent nuclear fuel storage facility requires a clear understanding of all processes (thermohydraulic, neutron-physical, aging processes, etc.) occurring in HI-STORM containers. And this cannot be achieved without modeling these processes using modern specialized programs. Modeling of neutron and photon transfer makes it possible to analyze the level of protective properties of the container against radiation, optimize the loading of MPC assemblies of different manufacturers and different levels of combustion and evaluate biological protection against neutron and gamma radiation. In the future it will allow to estimate the change in the isotopic composition of the materials of the container, which will be used for the management of aging processes at the centralized storage of spent nuclear fuel. The article is devoted to the development of the three-dimensional model of the HI-STORM storage system. The model was developed using the modern Monte Carlo code Serpent. The presented model consists of models of 31 spent fuel assemblies 438MT manufactured by TVEL company, model MPC-31 and model HISTORM 190. The model allows to perform a wide range of scientific tasks required in the operation of centralized storage of spent nuclear fuel.

scholarly journals Structural integrity investigations of spent nuclear fuel with finite element modeling

2021 ◽  
Vol 1 ◽  
pp. 13-14
Author(s):  
Efstathios Vlassopoulos ◽  
Susanne Pudollek ◽  
Olympios Alifieris ◽  
Dimitrios Papaioannou ◽  
Ramil Nasyrow ◽  
...  
Keyword(s):  
Finite Element ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Structural Integrity ◽  
Mechanical Response ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Final Disposal ◽  
National Cooperative

Abstract. Radioactive waste in Switzerland will be disposed of in a deep geological repository (DGR). Responsible for the planning and preparation of realization of this task is National Cooperative for the Disposal of Radioactive Waste (Nagra). Spent fuel assemblies (SFA) constitute the main high-level waste (HLW) stream that will be disposed in the DGR. Prior to final disposal they will be transferred or transported to an encapsulation plant, where they will be loaded into final disposal canisters. To ensure that the structural integrity of SFAs is not compromised during handling and transportation, it is desirable to characterize the expected mechanical parameters of SFAs after long-term interim storage. Experimental research activities performed at the JRC Karlsruhe include safety aspects of radioactive waste management, encompassing also spent fuel storage and spent fuel/HLW disposal activities. Nagra and JRC have established a collaboration to jointly study relevant properties and behaviours of spent fuel rods, with the support of the Gösgen nuclear power plant and of Framatome, and in collaboration with other partners in Europe and internationally. As part of this collaboration, 3-point bending and impact tests were performed at the hot-cell facilities of JRC Karlsruhe, to determine the mechanical response of spent fuel rodlets under quasi-static and dynamic loads. The structural integrity of fuel rods was also evaluated under different handling scenarios using finite element (FE) analysis. Starting with the construction of a static 3D FE model of a Pressurized Water Reactor (PWR) nuclear fuel rodlet in ANSYS Mechanical, Nagra has developed a series of FE models over the years. Mechanical properties of the original rodlet model were derived through an extensive validation process, using experimental data from the 3-point bending tests. To evaluate the mechanical response of an SFA in different loading scenarios, this model was expanded using 1D beam modeling approach. The development of the simplified 1D models is shown in this presentation. In particular, the effect of the contact formulation between the spacer grid and the rods is discussed. Finally, preliminary results of the bending response of a 15×15 PWR SFA sub-model are presented.

scholarly journals Introduction

2021 ◽  
pp. 1-12
Author(s):  
Cheol Ho Pyeon
Keyword(s):  
Feasibility Studies ◽  
Reaction Rates ◽  
Reactor Physics ◽  
Kinetics Parameters ◽  
Neutron Sources ◽  
Combined Use ◽  
Driven System ◽  
Accelerator Driven System ◽  
Pulsed Neutron ◽  
Mev Protons

AbstractAt the Kyoto University Critical Assembly (KUCA), the accelerator-driven system (ADS) is composed of a solid-moderated and solid-reflected core (A-core) and a pulsed-neutron generator (14 MeV neutrons) or the fixed-filed alternating gradient (FFAG) accelerator (100 MeV protons). At KUCA, two external neutron sources, including 14 MeV neutrons and 100 MeV protons, are separately injected into the A-core, and employed for carrying out the ADS experiments. With the combined use of the A-core and two external neutron sources, basic and feasibility studies of ADS have been engaged in the examination of neutronics of ADS, through the measurements of statics and kinetics parameters of reactor physics, including subcritical multiplication factor, subcriticality, prompt neutron decay constant, effective delayed neutron fraction, neutron spectrum, and reaction rates.

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