scholarly journals A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

2013 ◽  
Vol 2013 ◽  
pp. 1-10 ◽  
Author(s):  
Hangbok Choi ◽  
Robert W. Schleicher ◽  
Puja Gupta
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Fuel Cycle ◽  
Matrix Material ◽  
Reactor Core ◽  
Operation Time ◽  
Fuel Load ◽  
Fissile Material ◽  
Reactor Physics

In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2), which is a compact gas-cooled fast reactor (GFR). The EM2augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

scholarly journals Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 581-590 ◽  
Author(s):  
Przemysław Stanisz ◽  
Jerzy Cetnar ◽  
Grażyna Domańska
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Nuclear Fuel Cycle ◽  
Minor Actinide ◽  
Closed Nuclear Fuel Cycle ◽  
European Collaboration ◽  
Multiple Recycling

Abstract The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR) was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System) and LEADER (Lead-cooled European Advanced Demonstration Reactor) projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA) are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs), and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB) code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

An Overview of US EPA’s Current Radioactive Waste Management and General Radiation Protection Efforts

2009 ◽  
Author(s):  
R. Thomas Peake ◽  
Daniel Schultheisz ◽  
Loren W. Setlow ◽  
Brian Littleton ◽  
Reid Rosnick ◽  
...  
Keyword(s):  
United States ◽  
Radioactive Waste ◽  
Radiation Protection ◽  
Nuclear Power ◽  
Fuel Cycle ◽  
The United States ◽  
Uranium Mining ◽  
Radioactive Waste Disposal ◽  
Environmental Standards

The United States Environmental Protection Agency’s (EPA) Radiation Protection Division is the portion of EPA (or the Agency) that develops environmental standards for radioactive waste disposal in the United States. One current issue of concern is the disposal of low activity radioactive waste (LAW), including wastes that would be produced by a radiological dispersal device (RDD), for which current disposal options may be either inconsistent with the hazard presented by the material or logistically problematic. Another major issue is related to the resurgence in uranium mining. Over the past several years, demand for uranium for nuclear power plant fuel has increased as has the price. The increase in price has made uranium mining potentially profitable in the US. EPA is reviewing its relevant regulations, developed primarily in the 1980s, for potential revisions. For example, in-situ leaching (also known as in-situ recovery) is now the technology of choice where applicable, yet our current environmental standards are focused on conventional uranium milling. EPA has two actions in process, one related to the Clean Air Act, the other related to revising the environmental standards that implement the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA). Separately, but related, EPA has developed over the last several years uranium mining documents that address technologically enhanced natural occurring radioactive materials (TENORM) from abandoned uranium mines, and wastes generated by active uranium extraction facilities. Lastly, in 1977 EPA developed environmental standards that address nuclear energy, fuel fabrication, reprocessing, and other aspects of the uranium fuel cycle. In light of the increased interest in nuclear power and the potential implementation of advanced fuel cycle technologies, the Agency is now reviewing the standards to determine their continued applicability for the twenty-first century.

Possibility of Different Fuel Cycles Usage in GT-MHR

2002 ◽  
Author(s):  
N. Kodochigov ◽  
Yu. Sukharev ◽  
E. Marova ◽  
N. Ponomarev-Stepnoy ◽  
E. Glushkov ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Block Design ◽  
Average Power ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Fuel Load ◽  
Fuel Cycles ◽  
Partial Cycle ◽  
Control Rods ◽  
Neutronic Characteristics

The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis reindicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another.

scholarly journals A Simplified Supercritical Fast Reactor with Thorium Fuel

2014 ◽  
Vol 2014 ◽  
pp. 1-9
Author(s):  
Peng Zhang ◽  
Kan Wang ◽  
Ganglin Yu
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Mechanical Design ◽  
Inherent Safety ◽  
Fuel Conversion ◽  
Reactor Physics ◽  
Mox Fuel ◽  
Feasible Option ◽  
Gen Iv ◽  
Simulation Scheme

Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

scholarly journals On Capabilities and Limitations of Current Fast Neutron Flux Monitoring Instrumentation for the Demo LFR ALFRED

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Luigi Lepore ◽  
Romolo Remetti ◽  
Mauro Cappelli
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Signal To Noise Ratio ◽  
Reactor Power ◽  
Fast Reactors ◽  
Flux Monitoring

Among GEN IV projects for future nuclear power plants, lead-cooled fast reactors (LFRs) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant safety, and waste management. The novelty of this matter causes some open issues about coolant chemical aspects, structural aspects, monitoring instrumentation, etc. Particularly, hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e., source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal hydraulics and chemical conditions for the LFR core environment will be assumed, as the neutron flux will be studied extensively by the Monte Carlo transport code MCNPX (Monte Carlo N-Particles X-version). The core coolant’s high temperature drastically reduces the candidate instrumentation because only some kinds of fission chambers and self-powered neutron detectors can be operated in such an environment. This work aims at evaluating the capabilities of the available instrumentation (usually designed and tailored for sodium-cooled fast reactors) when exposed to the neutron spectrum derived from the Advanced Lead Fast Reactor European Demonstrator, a pool-type LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such a class of instrumentation does follow the power evolution, but is not completely suitable to detect the whole range of reactor power, due to excessive burnup, damages, or gamma interferences. Some improvements are possible to increase the signal-to-noise ratio by optimizing each instrument in the range of reactor power, so to get the best solution. The design of some new detectors is proposed here together with a possible approach for prototyping and testing them by a fast reactor.

A Procedure for the Pre-Conceptual Design of Fast Reactors and Application to a Gas-Cooled Sub-Critical Transmuter

2014 ◽  
Author(s):  
Carlo Fiorina ◽  
Konstantin Mikityuk ◽  
Jiři Křepel
Keyword(s):  
Conceptual Design ◽  
Fast Reactor ◽  
Power Distribution ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Maximum Pressure ◽  
Test Case ◽  
Pressure Losses ◽  
Design And Optimization ◽  
Inert Matrix

A C++ procedure has been developed for the design and optimization of Fast Reactor (FR) cores. It couples the ERANOS based EQL3D procedure developed at PSI for FR equilibrium fuel cycle analysis with a dedicated MATLAB script that evaluates the thermal-hydraulic characteristics of the reactor core. It is conceived to investigate reactors with both standard pins and annular pins. The procedure accepts as input the physical properties of the system, as well as a set of target core parameters presently consisting of core power, maximum fuel burnup, multiplication factor, inner pin diameter (for annular pins) or maximum pressure losses (for standard pins), and core height. It gives as a result a core design fulfilling these design objectives and meeting the constraints on maximum fuel and clad temperatures. In case of annular pins, it also equalizes the temperature rise inside and outside of the core average pin. The procedure considers the possibility of two-zone cores and adjusts the fuel composition in the two zones to achieve an optimal radial power distribution. Finally, it can evaluate safety parameters and fuel cycle characteristics both at beginning-of-life and at equilibrium. As a test case, the procedure has been used for the pre-conceptual design of a sub-critical Gas Fast Reactor core employing inert-matrix sphere-pac fuel and annular pins with SiC cladding.

scholarly journals A study into the dependence of the cladding-fuel pellet gap conductance on burn-up and the effects on the reactor core neutronic performance

2019 ◽  
Vol 5 (2) ◽  
pp. 97-102
Author(s):  
Sergey B. Vygovsky ◽  
Fedor V. Gruzdov ◽  
Rashdan T. Al Malkawi
Keyword(s):  
Fuel Cycle ◽  
Accumulation Rate ◽  
Reactor Core ◽  
Neutron Field ◽  
Fuel Load ◽  
Fuel Pellet ◽  
Load Following ◽  
Fuel Burn ◽  
Numerical Research ◽  
Neutronic Performance

This paper presents the results of the research to study the dependence of the VVER-1000 (1200) cores neutronic characteristics on the cladding – fuel pellet gap conductance coefficient in the process of the fuel burn-up. The purpose of the study was to determine more accurately the dependence of the cladding – fuel pellet gap conductance coefficient on the fuel burn-up as shown in the Final Safety Report for the Bushehr NPP and to determine the extent of the effects this dependence had on the spatial distribution of the neutron field, on the xenon accumulation rate, and on the kinetic and dynamic behavior of the reactor facility. The paper presents the results of calculating the parameters using which the heat engineering safety of the reactor core is monitored in the process of the fuel burn- up (for a generalized fuel load of a VVER-1000) during the transition to an 18-month nuclear fuel cycle. This paper also includes the results of a numerical research to determine the cladding – fuel gap conductance coefficient depending on the fuel burn-up. These results have shown that, in reality, the gap conductance coefficient dependence on the burn-up does not affect greatly the steady-state characteristics. At the same time, it affects to rather a great extent the xenon accumulation rate, specifically in the event of an extended fuel life. In conditions of maneuvering (load following) modes accompanied by the xenon processes in the reactor core. These facts should be into consideration in design of engineering codes, that used to support the operation of the VVER-1000 (1200) and full-scale simulators.

scholarly journals Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

2017 ◽  
Vol 153 ◽  
pp. 07031
Author(s):  
Georgy Tikhomirov ◽  
Mikhail Ternovykh ◽  
Ivan Saldikov ◽  
Peter Fomichenko ◽  
Alexander Gerasimov
Keyword(s):  
Isotopic Composition ◽  
Nuclear Fuel ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Nuclear Fuel Cycle ◽  
Closed Nuclear Fuel Cycle

scholarly journals Study of Helium Cooled Fast Reactor Core Design Fuelled by Thorium Carbide-Uranium Carbide with Modified Candle Axial Direction Scheme

2021 ◽  
Vol 2072 (1) ◽  
pp. 012002
Author(s):  
Z Su’ud ◽  
N R Galih ◽  
M Ariani
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Nuclear Energy ◽  
Reactor Core ◽  
Axial Direction ◽  
Natural Uranium ◽  
Design Calculation ◽  
Uranium Carbide ◽  
Human Need ◽  
Core Height

Abstract Human need of energy will increase time to time. Therefore, a safe, renewable, and efficient source of energy, which is Nuclear Energy, is needed. Nuclear Power Plant (NPP) is the most compatible solution to provide electricity to human race in the future. The problem that came within NPP is the danger of proliferation issues. The method that has been developed to overcome this problem is CANDLE [5] and has been modified by Prof. Zaki Su’ud (Modified CANDLE scheme). This research use Axial Modified CANDLE Scheme to Helium-Cooled Fast Reactor with Natural Uranium Carbide-Thorium Carbide as fuel and applied to various size of core as optimization. Neutronic aspect such as, burn up level, multiplication factor, and conversion ratio are utilized in this paper in order to analyse the behaviour of the reactor. Other than that, percentage of Uranium has been varied to reduce power peaking. The neutronic calculation has been done using SRAC and core design calculation by FI-ITB-CH1. This research concludes that power peaking reduction is able to achieve by combining Uranium Carbide and Thorium Carbide to the fuel. The optimum reactor design reached at 360 cm of core radius and 303 cm of core height.

scholarly journals Evaluation of the neutronic performance of a fast traveling wave reactor in the Th-U fuel cycle

2020 ◽  
Vol 6 (2) ◽  
pp. 77-82
Author(s):  
Alina Ye. Pomysukhina ◽  
Yury P. Sukharev ◽  
German N. Vlasichev
Keyword(s):  
Power Density ◽  
Traveling Wave ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Raw Material ◽  
Initial Time ◽  
Seed Region ◽  
Fissile Material ◽  
Neutron Fission ◽  
Fission Wave

The possibility for all of the uranium or thorium fuel to be used nearly in full is expected in traveling wave reactors. A traveling wave reactor core with a fast neutron spectrum in a thorium-uranium cycle has been numerically simulated. The reactor core is shaped as a rectangular prism with a seed region arranged at one of its ends for the neutron fission wave formation. High-enriched uranium metal is used as the seed region fuel. Calculated power density dependences and concentrations of the nuclides involved with the transformation chain along the core at a number of time points have been obtained. The results were graphically processed for the clear demonstration of the neutron fission wave occurrence and transmission in the reactor. The obtained power density dependence represents a soliton (solitary wave) featuring a distinct time repeatability. Neutron spectra and fission densities are shown at the initial time point, when no wave has yet formed, and at the time of its formation. The wave rate has been calculated based on which the reactor life was estimated. The fuel burn-up has been estimated the ultra-high value of which makes the proposed reactor concept hard to implement. The burn-up of most of both the raw material and the fissile material it produces indicates a high potential efficiency of the developed reactor concept in terms of fuel utilization and nuclear nonproliferation.

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