Increasing Fuel Utilization of Breed and Burn Reactors PDF Download

Author: Christian Diego Di Sanzo
Publisher:
ISBN:
Category :
Languages : en
Pages : 155

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Breed and Burn reactors (B & B), also referred to Traveling Wave Reactors, are fast spectrum reactors that can be fed indefinitely with depleted uranium only, once criticality is achieved without the need for fuel reprocessing. Radiation damage to the fuel cladding limits the fuel utilization of B & B reactors to ~ 18-20% FIMA (Fissions of Initial Metal Atoms) - the minimum burnup required for sustaining the B & B mode of operation. The fuel discharged from this type of cores contain ~ 10% fissile plutonium. Such a high plutonium content poses environmental and proliferation concerns, but makes it possible to utilize the fuel for further energy production. The objectives of the research reported in this dissertation are to analyze the fuel cycle of B & B reactors and study new strategies to extend the fuel utilization beyond ~ 18-20% FIMA. First, the B & B reactor physics is examined while recycling the fuel every 20% FIMA via a limited separation processing, using either the melt refining or AIROX dry processes. It was found that the maximum attainable burnup varies from 54% to 58% FIMA - depending on the recycling process and on the fraction of neutrons lost via leakage and reactivity control. In Chapter 3 the discharge fuel characteristics of B & B reactors operating at 20% FIMA and 55% FIMA is analyzed and compared. It is found that the 20% FIMA reactor discharges a fuel with about ~ 80% fissile plutonium over total plutonium content. Subsequently a new strategy of minimal reconditioning, called double cladding is proposed to extend the fuel utilization in specifically designed second-tier reactors. It is found that with this strategy it is possible to increase fuel utilization to 30% in a sodium fast reactor and up to 40% when a subcritical B & B core is driven by an accelerator-driven spallation neutron source. Lastly, a fuel cycle using Pressurized Water Reactors (PWR) to reduce the plutonium content of discharged B & B reactors is analyzed. It was found that it is possible to burn the B & B discharged fuel up to an additional 105.6 GWd/MTIHM and 66 GWd/MTIHM, for melt refining and AIROX, respectively.

Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 216

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The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B & B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B & B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B & B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

Author: Florent Heidet
Publisher:
ISBN:
Category :
Languages : en
Pages : 412

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The primary objective of this study was to estimate the maximum fuel utilization that is achievable using fast reactors that are designed to operate with fuel reconditioning when the fuel reaches its radiation damage constraint. The primary functions of the fuel reconditioning are to relieve the pressure of the gaseous fission products, replace the clad and reduce radiation induced defects in the fuel. That is, the fuel recycling processes considered cannot be used for the separation of actinides from most of the fission products and to extract plutonium or any other actinide from the fuel. Hence, these recycling processes are highly proliferation resistant and, hopefully, less expensive than processes traditionally considered for used fuel recycling. With the fuel reconditioning recycling, the maximum discharge burnup is dictated by the reactivity of the fuel and is not limited by the material damage constraints. A couple of fuel management strategies were examined: the conventional multi-batch fuel management and a breed-and-burn mode of operation. The two recycling processes examined are an AIROX-like process and the melt-refining process. It is found that using fuel reconditioning it is possible to increase the fuel utilization by up to one order of magnitude with the conventional mode of operation and by two orders of magnitude using the breed and burn (B & B) mode of operation, relative to that achievable in once-trough LWRs. The conventional mode of operation, in which the fast reactor is constantly fed with enriched fuel, enables to achieve a discharge burnup ranging from 52.4% FIMA for a medium size 1200 MWth fast reactor to 65.3% FIMA for a large 3000 MWth fast reactor. With the innovative breed and burn mode of operation only depleted uranium is required for the fuel feed with the exception of the fissile fuel required for establishing the initial criticality. The achievable burnup is up to 57% FIMA in a 3000 MWth B & B reactor and up to 44% FIMA in a 1200 MWth B & B reactor. In order to sustain the breed and burn mode of operation it is necessary to accumulate in the depleted uranium feed an average burnup of at least 20% FIMA, when using metallic uranium fuel alloyed with 10 weight % zirconium in a tight-lattice core. By discharging the fuel at this minimum required burnup and loading it, after reconditioning, into a new reactor along with fresh depleted uranium, it is possible to spawn additional B & B reactors without need for any additional fissile fuel. With this spawning mode of operation the achievable B & B reactor capacity growth rate is 3.85% per year, without need for uranium enrichment capability or actinides separation capability. The energy value of the depleted uranium currently accumulated in the USA, when used in the proposed breed-and-burn fast reactors, is equivalent to at least seven centuries of the total 2009 USA supply of electricity, all sources included. Relative to LWR operating with the once-through fuel cycle, the fuel discharged from the B & B fast reactors at ~57% FIMA features, per unit of electricity generated: (a) ~40% the amount of TRU and Pu; (b) ~10% the inventory of 237Np and its precursors; (c) ~12% of the decay heat from TRU; (d) ~28% of the radiotoxicity; (e) ~7% the neutron emission rate; the latter three are measured one year following discharge. The fraction of the fissile isotopes in the discharged plutonium is comparable but the decay heat and neutron emission rate per unit mass of discharged plutonium are nearly half as large. The proposed modes of operation are expected to improve the economics and the proliferation resistance and, hence, may justify sooner deployment of fast reactors. The deployment of the suggested fast reactor system will constitute a significant step forward towards sustainable nuclear energy. However, technologies for fuel reconditioning need be developed and their economic viability need be established.

Author: Thomas James Dolan
Publisher: Elsevier
ISBN: 0323993567
Category : Technology & Engineering
Languages : en
Pages : 1068

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Molten Salt Reactors and Thorium Energy, Second Edition is a fully updated comprehensive reference on the latest advances in MSR research and technology. Building on the successful first edition, Tom Dolan and the team of experts have fully updated the content to reflect the impressive advances from the last 5 years, ensuring this book continues to be the go-to reference on the topic. This new edition covers progress made in MSR design, details innovative experiments, and includes molten salt data, corrosion studies and deployment plans. The successful case studies section of the first edition have been removed, expanded, and fully updated, and are now published in a companion title called Global Case Studies on Molten Salt Reactors. Readers will gain a deep understanding of the advantages and challenges of MSR development and thorium fuel use, as well as step-by-step guidance on the latest in MSR reactor design. Each chapter provides a clear introduction, covers technical issues and includes examples and conclusions, while promoting the sustainability benefits throughout. A fully updated comprehensive handbook on Molten Salt Reactors and Thorium Energy, written by a team of global experts Covers MSR applications, technical issues, reactor types and reactor designs Includes 3 brand new chapters which reflect the latest advances in research and technology since the first edition published Presents case studies on molten salt reactors which aid in the transition to net zero by providing abundant clean, safe energy to complement wind and solar powe

Author: Etienne Parent
Publisher:
ISBN:
Category :
Languages : en
Pages : 236

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A comparative analysis of nuclear fuel cycles was carried out. Fuel cycles reviewed include: once-through fuel cycles in LWRs, PHWRs, HTGRs, and fast gas cooled breed and burn reactors; single-pass recycle schemes: plutonium recycle in LWRs and direct-use of spent PWR fuel in CANDU reactors (DUPIC); multi-pass recycle schemes: transmutation of transuranics in LWRs, fast reactors, double strata systems, and molten salt reactors. Mass flow calculations for the fuel cycles at equilibrium were carried out based on data available in the open literature, and results were used to compare the performance of the fuel cycles with respect to uranium utilization, waste management, proliferation resistance, and economics. Potential for mid-century deployment was assessed based on these results. Once-through fuel cycles based on solid fuel thermal reactors are found to be the best candidates for mid-century deployment because the substantial increase in electricity costs entailed by reprocessing schemes is unlikely to be justified by the afforded reductions in long-term proliferation and waste management risks. Furthermore, once-through cycles present lower proliferation and waste management risks in the short-term and their inefficient use of uranium is not likely to become an important issue before the middle of the century even under a high growth scenario.

Author: Robert C. Petroski
Publisher:
ISBN:
Category :
Languages : en
Pages : 351

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A new theoretical framework is introduced, the "neutron excess" concept, which is useful for analyzing breed-and-burn (B&B) reactors and their fuel cycles. Based on this concept, a set of methods has been developed which allows a broad comparison of B&B reactors using different fuels, structural materials, and coolants. This new approach allows important reactor and fuelcycle parameters to be approximated quickly, without the need for a full core design, including minimum burnup/irradiation damage and reactor fleet doubling time. Two general configurations of B&B reactors are considered: a "minimum-burnup" version in which fuel elements can be shuffled in three dimensions, and a "linear-assembly" version composed of conventional linear assemblies that are shuffled radially. Based on studies of different core compositions, the best options for minimizing fuel burnup and material DPA are metal fuel (with a strong dependence on alloy content), the type of steel that allows the lowest structure volume fraction, and helium coolant. If sufficient fuel performance margin exists, sodium coolant can be substituted in place of helium to achieve higher power densities at a modest burnup and DPA penalty. For a minimum-burnup B&B reactor, reasonably achievable minimum DPA values are on the order of 250-350 DPA in steel, while axial peaking in a linear-assembly B&B reactor raises minimum DPA to over 450 DPA. By recycling used B&B fuel in a limited-separations (without full actinide separations) fuel cycle, there is potential for sodium-cooled B&B reactors to achieve fleet doubling times of less than one decade, although this result is highly sensitive to the reactor core composition employed as well as thermal hydraulic performance.

Author: International Atomic Energy Agency
Publisher:
ISBN:
Category : Technology & Engineering
Languages : en
Pages : 72

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Author:
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ISBN:
Category :
Languages : en
Pages :

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Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium (LEU) fuels. Examples of systems in this class include the small modular reactors being considered internationally; e.g. 4S [Tsuboi 2009], Hyperion Power Module [Deal 2010], ARC-100 [Wade 2010], and SSTAR [Smith 2008]. (2) Systems for Resource Utilization - In recent years, interest has developed in the use of advanced nuclear designs for the effective utilization of fuel resources. Systems under this class have generally utilized the breed and burn concept in which fissile material is bred and used in situ in the reactor core. Due to the favorable breeding that is possible with fast neutrons, these systems have tended to be fast spectrum systems. In the once-through concepts (as opposed to the traditional multirecycle approach typically considered for fast reactors), an ignition (or starter) zone contains driver fuel which is fissile material. This zone is designed to last a long time period to allow the breeding of sufficient fissile material in the adjoining blanket zone. The blanket zone is initially made of fertile depleted uranium fuel. This zone could also be made of fertile thorium fuel or recovered uranium from fuel reprocessing or natural uranium. However, given the bulk of depleted uranium and the potentially large inventory of recovered uranium, it is unlikely that the use of thorium is required in the near term in the U.S. Following the breeding of plutonium or fissile U-233 in the blanket, this zone or assembly then carries a larger fraction of the power generation in the reactor. These systems tend to also have a long cycle length (or core life) and they could be with or without fuel shuffling. When fuel is shuffled, the incoming fuel is generally depleted uranium (or thorium) fuel. In any case, fuel is burned once and then discharged. Examples of systems in this class include the CANDLE concept [Sekimoto 2001], the traveling wave reactor (TWR) concept of TerraPower [Ellis 2010], the ultra-long life fast reactor (ULFR) by ANL [Kim 2010], and the BNL fast mixed spectrum reactor (FMSR) concept [Fisher 1979]. (3) Thermal systems for resource extension - These systems were primarily considered during the INFCE/NASAP evaluations [NASAP 1979] and include various LWR designs for increasing resource utilization (both uranium and thorium). This class would include the Radkowsky seed-blanket concept. Also included in this class are the thermal reactor systems being considered for deployment as small modular reactors, such as IRIS [Carelli 2004], mPower [mPower], and NuScale [NuScale] that are all water cooled reactors. The purpose of this work is to provide relevant systems and fuel cycle information for some of these once-through fuel cycle systems. In this report, the intent is on providing information on most of the systems from open sources and from scoping studies recently done within the program. As there is insufficient fuel cycle information on the first class of systems, they are not discussed in this report.

Author: Kevan D. Weaver
Publisher:
ISBN:
Category :
Languages : en
Pages :

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The Idaho National Laboratory (INL) contribution to the Nuclear Energy Research Initiative (NERI) project number 2002-005 was divided into reactor physics, and thermal-hydraulics and plant design. The research targeted credible physics and thermal-hydraulics models for a gas-cooled fast reactor, analyzing various fuel and in-core fuel cycle options to achieve a true breed and burn core, and performing a design basis Loss of Coolant Accident (LOCA) analysis on that design. For the physics analysis, a 1/8 core model was created using different enrichments and simulated equilibrium fuel loadings. The model was used to locate the hot spot of the reactor, and the peak to average energy deposition at that location. The model was also used to create contour plots of the flux and energy deposition over the volume of the reactor. The eigenvalue over time was evaluated using three different fuel configurations with the same core geometry. The breeding capabilities of this configuration were excellent for a 7% U-235 model and good in both a plutonium model and a 14% U-235 model. Changing the fuel composition from the Pu fuel which provided about 78% U-238 for breeding to the 14% U-235 fuel with about 86% U-238 slowed the rate of decrease in the eigenvalue a noticeable amount. Switching to the 7% U-235 fuel with about 93% U-238 showed an increase in the eigenvalue over time. For the thermal-hydraulic analysis, the reactor design used was the one forwarded by the MIT team. This reactor design uses helium coolant, a Brayton cycle, and has a thermal power of 600 MW. The core design parameters were supplied by MIT; however, the other key reactor components that were necessary for a plausible simulation of a LOCA were not defined. The thermal-hydraulic and plant design research concentrated on determining reasonable values for those undefined components. The LOCA simulation was intended to provide insights on the influence of the Reactor Cavity Cooling System (RCCS), the containment building, and a Decay Heat Removal System (DHRS) on the natural circulation heat transfer of the core's decay heat. A baseline case for natural circulation had to be established in order to truly understand the impact of the added safety systems. This baseline case did not include a DHRS, although the current MIT design does have a DHRS that features the highly efficient Printed Circuit Heat Exchangers (PCHEs). The initial LOCA analysis revealed that the RCCS was insufficient to maintain the reactor core below the fuel matrix decomposition temperature. A guard containment was added to the model in order to maintain a prescribed backpressure during the LOCA to enhance the natural circulation. The backpressure approach did provide satisfactory natural convection during the LOCA. The necessary backpressure was 1.8 MPa, which was not especially different from the values reported by other gas fast reactor researchers. However, as the model evolved to be more physically representative of a nuclear reactor, i.e., it included radial peaking factors, inlet plenum orificing, and the degradation of SiC thermal properties as a result of irradiation, the LOCA-induced fuel temperatures were not consistently below the decomposition limit.

Author: Willem Frederik Geert van Rooijen
Publisher: IOS Press
ISBN: 9781586036966
Category : Technology & Engineering
Languages : en
Pages : 160

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The Generation IV Forum is an international nuclear energy research initiative aimed at developing the fourth generation of nuclear reactors, envisaged to enter service halfway the 21st century. One of the Generation IV reactor systems is the Gas Cooled Fast Reactor (GCFR), the subject of study in this thesis. The Generation IV reactor concepts should improve all aspects of nuclear power generation. Within Generation IV, the GCFR concept specifically targets sustainability of nuclear power generation. The Gas Cooled Fast Reactor core power density is high in comparison to other gas cooled reactor concepts. Like all nuclear reactors, the GCFR produces decay heat after shut down, which has to be transported out of the reactor under all circumstances. The layout of the primary system therefore focuses on using natural convection Decay Heat Removal (DHR) where possible, with a large coolant fraction in the core to reduce friction losses.