The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant (original) (raw)
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AIP Conference Proceedings, 2012
A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i th region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.
A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.
Design of small gas cooled fast reactor with two region of natural Uranium fuel fraction
AIP Conference Proceedings, 2012
A design study of small Gas Cooled Fast Reactor with two region fuel has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region fuel i.e. 60% fuel fraction of Natural Uranium as inner core and 65% fuel fraction of Natural Uranium as outer core. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 filled by fresh Natural Uranium. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i th region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium on each region-1. The burn-up calculation is performed using collision probability method PIJ (cell burn-up calculation) in SRAC code which then given eight energy group macroscopic cross section data to be used in two dimensional R-Z geometry multi groups diffusion calculation in CITATION code. This reactor can results power thermal 600 MWth with average power density i.e. 80 watt/cc. After reactor start-up the operation, furthermore reactor only needs Natural Uranium supply for continue operation along 100 years. This calculation result then compared with one region fuel design i.e. 60% and 65% fuel fraction. This core design with two region fuel fraction can be an option for fuel optimization.
HIGH-ENERGY PROCESSES IN CONDENSED MATTER (HEPCM 2020): Proceedings of the XXVII Conference on High-Energy Processes in Condensed Matter, dedicated to the 90th anniversary of the birth of RI Soloukhin
Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium nitride fuel has been done. The purpose of this study was to compare various thermal powers on nuclear power plant (NPP) Gas Cooled Reactor type. Neutronics calculation was design by using SRAC Code version 2006 (Standard Reactor Analysis Code) with the data nuclides from JENDL-4.0 under the Linux Operating System with nuclear data library JENDL4.0. Neutronic calculations were done through some parameter surveys to obtain the optimal results. The first step was calculation of fuel cell (PIJ-method) by using hexagonal cell and then followed by calculation of core reactor (CITATION-method). The power variations carried out are seven variations of power, namely from 100 MWth to 700 MWth. In this power variation all parameters are made the same. The parameters that are made are the same, namely the percentage of plutonium, the volume fraction of the fuel, the reactor geometry type, the diameter and the active core height (terrace). All power variations have an average power density value and maximum power density increases along with the increase in thermal power. When thermal power increases, the k-eff peak value will be increase too. It shows that if the thermal power increases, the burn-up fuel is also increase more than before, so that the fuel is used for the burn-up process which causes k-eff increase. The increasing graph which shows in the figure explains that the reactor breeding occurs, while the declining graph shows the reactor is burning.
Neutronic Design of Uranium-Plutonium Nitride Fuel-Based Gas-Cooled Fast Reactor (GFR)
Jurnal Pendidikan Fisika Indonesia, 2018
This study presents the calculation results of the cell, and core Gas-cooled Fast Reactor (GFR) based fuel Uranium-Plutonium Nitride (U, Pu)N. Parameter survey results of calculations of the fuel cell consisting of a kinf, burnup level, and the conversion ratio and for the calculation of the reactor core produce value keff during a refueling cycle. The calculation was performed by using a set of SRAC program by comparing three types of fuel cell designs. Reactor Design A based on natural uranium could not reach criticality because of keff < 1. Design B used the enrichment of uranium-235 by 9.5% to reach a critical condition at keff > 1. The critical state was also achieved by Design C utilizing natural uranium, and plutonium 5.5% result value keff = 1.015 in the first year of burnup and continues to increase 1.083 in the tenth year without refueling. Moreover, plutonium can replace the uranium enrichment process.Penelitian ini menyajikan hasil perhitungan sel dan teras gas-coo...
Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor
2015
Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.
Startup of “Candle” burnup in fast reactor from enriched uranium core
Energy Conversion and Management, 2006
A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burnup reactivity control mechanisms such as control rods are not required. This burnup strategy can derive many merits. The change of excess reactivity along burnup is theoretically zero, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Therefore the operation of the reactor becomes much easier than the conventional reactors especially for high burnup reactors. The transportation and storage of replacing fuels become easy and safe, since they are free from criticality accidents. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. The average burnup of the spent fuel is about 40 % that is equivalent to 40 % utilization of the natural uranium without the reprocessing and enrichment. The initial core must be prepared with easily available materials. Actinides are simulated by enriched uranium with changing enrichment for each position, and fission products by niobium. The maximum value of enrichment is 13% well below 20%. The obtained effective neutron multiplication factor oscillates with burnup, but the maximum change with time is only 0.0008.
Study on small long-life LBE cooled fast reactor with CANDLE burn-up – Part I: Steady state research
Progress in Nuclear Energy, 2008
Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Leade Bismuth is used as coolant. From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%.
A design study of Pb-Bi-cooled fast reactors with natural uranium as the fuel cycle input
International Journal of Nuclear Energy Science and Technology, 2009
A conceptual design study of Pb-Bi-cooled fast reactors in which the fuel cycle needs only natural uranium input has been performed. In this design, the reactor cores are subdivided into several parts with the same volume. The region with natural uranium is put in the central core and the outer region is arranged with increasing plutonium content. In some cases, the region with natural uranium content can be put in the most outer part of the core. The arrangement of the plutonium content takes account of the criterion that the fuel in a certain part must be fit for fresh fuel in the next higher-enrichment region. Therefore, at the end of a long-life operation, we just need to supply natural uranium fuel to the blanket region and move to the next region. As an example using the SRAC and FI-ITB CH code systems, we have a core with a 15-20 year lifetime per subcycle.
Annals of Nuclear Energy, 2005
In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine-modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: 235 U, which represents the 20% of the fresh uranium, 233 U, which is produced by the transmutation of fertile 232 Th, and 239 Pu, which is produced by the transmutation of fertile 238 U. In order to compensate the depletion of 235 U with the breeding of 233 U and 239 Pu, the quantity of fertile nuclides must be much larger than that one of 235 U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of 235 U. At the same time, the amount of 235 U