Breeder reactor

Summary

A breeder reactor is a nuclear reactor that generates more fissile material than it consumes. These reactors can be fuelled with more commonly available isotopes of uranium and thorium, such as uranium-238 or thorium-232, as opposed to the rare uranium-235 which is used in conventional reactors. These materials are called fertile materials since they can be bred into fuel by these breeder reactors. Breeder reactors achieve this because their neutron economy is high enough to create more fissile fuel than they use. These extra neutrons are absorbed by the fertile material that is loaded into the reactor along with fissile fuel. This irradiated fertile material in turn transmutes into fissile material which can undergo fission reactions. Breeders were at first found attractive because they made more complete use of uranium fuel than light-water reactors, but interest declined after the 1960s, as more uranium reserves were found, and new methods of uranium enrichment reduced fuel co

Official source

https://en.wikipedia.org/wiki/Breeder_reactor

About this result

This page is automatically generated and may contain information that is not correct, complete, up-to-date, or relevant to your search query. The same applies to every other page on this website. Please make sure to verify the information with EPFL's official sources.

Related publications (17)

Related publications

Related people (1)

Related people

Related units

Related concepts (79)

Related concepts

Related courses (11)

Related courses

Related lectures (41)

Related lectures

The general license application (RBG) for the Swiss deep geological repository will be submitted in 2024. Furthermore, the decommissioning of the nuclear power plants (NPPs) is coming soon, with the Mühleberg NPP (KKM) scheduled to shut down in 2019. In support of the associated analyses and planning, availability of accurate descriptions of the nuclear waste streams is of paramount importance. The work described in this thesis represents a significant improvement of Nagra's most important radioactive waste characterization methodologies: for spent fuel, NPP decommissioning waste, and reactor waste. The resulting high-fidelity nuclide inventories will decisively support the future waste disposal research and development, as well as serve the NPP decommissioning planning in general. The spent fuel characterization demonstrated is based on Polaris. This approach is compared against the code sequence used in the industry (Studsvik CMS), in order to confirm that, when the necessary fuel data is available, either one of these codes can be used to provide detailed nuclide vector for the spent fuel. Afterwards, possible ways to produce the desired nuclide vectors in the future are identified. For decommissioning waste, every step of the old characterization methodology has been further developed and improved upon. This includes a new, detailed, fully three-dimensional approach to NPP modeling and the development of a new activation sequence (based on the ORIGEN depletion solver), which outputs a highly-resolved component-wise activation distribution and visualization. Additionally, the methodology is extended to apply the method's results for algorithm-optimized packaging concepts and waste volume minimization through decay storage and free release analyses. At the end, recommendations for the future improvements are presented. The decommissioning waste characterization methodology is then expanded to accommodate reactor waste, and the more complex nature of the non-stationary components of this waste stream (such as the control rods). A systematic general approach, based additionally on the component dose rate measurements, allows for a more efficient and flexible characterization of all reactor waste. These developments serve as an important foundation in the ongoing transition from a conservative approach towards a best-estimate plus uncertainty (BEPU) waste characterization, which is vital for accurate safety analysis studies, as well as waste and cost minimization. The results produced by these methodologies will form the basis for the next version of MIRAM (the Swiss Model Inventory of Radioactive Materials), MIRAM2020, which is to be used for the RBG. They will also be used as part of the next Swiss NPP decommissioning cost study, KS2021. At the same time, the obtained results are already being provided to the Swiss NPPs planning their decommissioning, specifically KKM and Beznau (KKB), allowing detailed component-wise segmentation strategies and packaging concepts decision making, finally leading to notable cost reduction for the utilities.

EPFL2019

, , ,

Aqueous effluent treatment system (2) including a separation reactor (4) having a reactor chamber fluidly connected to an aqueous effluent source (3), connected via a pump (5) to an inlet (34) of the reactor chamber, a fluid extraction system (6) connected to a liquid effluent outlet (38) at a top of the reactor chamber, and a solid residue extraction system (8) connected to a solid residue outlet (36) at a bottom of the reactor chamber. The separation reactor is operable to generate pressures exceeding 22 MPa and temperatures exceeding 300°C in the reactor chamber configured to generate a supercritical zone (40) in an upper portion of the reactor chamber to which the liquid effluent outlet (38) is connected, and a subcritical zone (42) in a lower portion of the chamber within the reactor chamber to which the solid residue outlet (36) is connected. The solid residue extraction system comprises an output circuit (10) comprising a collector (18) coupled to the solid residue outlet via a collector input valve (V1) and to a water output tank (32) via a filter (26) and a collector liquid output valve (V4) operable to be opened to cause a pressure drop at the solid residue outlet to draw solid residue out of the reactor chamber, the solid residue extraction system further comprising a gas feed circuit (14) connected via a gas supply valve (V5) to the collector, the gas supply valve operable to be opened to extract solid residues in the collector to a solids output tank (20) connected to the collector via a collector solids output valve (V6) .

2020

Improved gamma-heating calculational methods for fast reactors and their validation for plutonium-burning configurations

Rakesh Chawla, Gérald Rimpault

A new calculational scheme has been developed for the accurate assessment of gamma heating in fast reactors, its special feature being the determination of the gamma source distribution that is formulated in a near-to-exact manner. The improved methodology, which has been implemented into the ERANOS (European Reactor Analysis Optimized System) code package, is currently validated for Puburning configurations, for which gamma-heating target accuracies are particularly high. This has been accomplished through comparisons with new integral measurements conducted at the MASURCA facility, as well as with reevaluated earlier experiments. In the new measurements, absolute gamma-heating rates were determined in PuO2/UO2-fueled cores surrounded by a steel/sodium reflector, mainly using TLD-700 thermoluminescent dosimeters. Thereby, a considerable effort was undertaken to minimize systematic errors. The calculation/experiment values determined from the analysis of the critical experiments are 0.90 for the PuO2/UO2 core region, 0.84 for the steel/sodium reflector, and 0.89 for an internal steel/sodium diluent zone. The most plausible causes for the observed discrepancies have been identified to be data related, i.e., too low fission gamma energies and too low capture cross sections for the structural elements. The transferability of the current validation findings to a modified Superphenix configuration, in which the radial fertile blanket is replaced by a steel/sodium reflector, and to the 1500 MW(electric) Pu-burning CAPRA 4/94 reference design has been demonstrated.

American Nuclear Society2001

EPFL2019