Criticality Uncertainty Dependence on Nuclear Data Library in Fast Molten Salt Reactors

AbstractTo increase the sustainability of the nuclear fuel cycle, and increase security of nuclear energy, we have been inves- tigating Molten Salt Fast Reactors (MSFR) for transmutation of Minor actinoid (MA) isotopes. In the present work we describe the reactor physics analysis of a Th-TRU MSFR using a LiF-ThF4-TRUF3-fuel salt. We investigated the uncertainty of major reactor physics parameters using 3 sets of evaluated nuclear data: JENDL-4.0, JEFF-3.1.2, and ENDF/B-VII.1. The result of our work is that the spread in the multiplication factor is rather large between the sets of nuclear data, while other parameters are by and large the same. The uncertainties due to cross section covariance are large, with Th-232, U-233, and F-19 giving the most important contributions. The isotopic contributions to the uncertainties are quite different between the sets of nuclear data, giving a suspicion that the covariance data may is very different between the evaluations, and a review of the covariance data may be needed

Design of a Spherical Fuel Element for a Gas Cooled Fast Reactor

A study is done to develop a fuel cycle for a Gas Cooled Fast Reactor (GCFR). The design goals are: highly efficient use of (depleted) uranium, application of Pu recycled from LWR discharge as fissile material, high temperature output, and simplicity of design. The design focuses on spherical TRISO-like fuel elements, a homogeneous core at startup, providing for easy fuel fabrication, and self-breeding capability with a flat keff with burnup. Nitride fuel (15N > 99%) has been selected because of it's favourable thermal conductivity, high Heavy Metal density and compatability with PUREX reprocessing. Two core concepts have been studied: one with coated particles embedded inside fuel pebbles, and one with coated particles cooled directly by helium. The result is that a flat keff can be achieved for a long period of time, using coated particles cooled directly, with a homogeneous core at startup, with a closed fuel cycle and a simple refueling and reprocessing scheme

Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle", where only natural uranium is used as raw material, and only fission products are discharged to a repository. Uranium and heavier isotopes (plutonium, americium, etc) are recycled in the reactor and are eventually fissioned. Since the heavy isotopes determine the long-term radiotoxicity of the nuclear waste, application of a closed fuel cycle maximizes the energy output of the fuel material, and can significantly reduce the lifetime of nuclear waste. It is shown that it is possible to obtain a closed fuel cycle with the GCFR. Coated particle fuel and ceramic plate fuel are discussed as fuel candidates for GCFR. A theoretical framework is derived using a combination of eigenvalue perturbation theory and nuclide perturbation theory to estimate the Breeding Gain of the reactor, and the change of Breeding Gain due to changes of the initial fuel composition. It is shown that the GCFR has some potential as an actinide transmutation reactor. To increase the safety of the GCFR, passive elements have been designed to automatically shut down the reactor in case of incidents. These elements use liquid Li-6, which is introduced into special assemblies if necessary. It has been shown that with these elements the reactor can withstand transients without damage to the fuel. The research has been performed in a European framework.Applied Science

Design and neutronic analysis of the intermediate heat exchanger of a fast-spectrum molten salt reactor

Various research groups and private interprises are pursuing the design of a Molten Salt Reactor (MSR) as one of the Generation-IV concepts. In the current work a fast neutron MSR using chloride fuel is analyzed, specially analyzing the power production and neutron flux level in the Intermediate Heat Exchanger (IHX). The neutronic analysis in this work is based on a chloride-fuel MSR with 600 MW thermal power. The core power density was set to 100 MW m −3 with a core H/D [[EQUATION]] 1.0 amd four Intermediate Heat Exchanger (IHX). This leads to a power of 150 MW per IHX; this power is also comparable to the IHX proposed in the SAMOFAR framework. In this work, a preliminary design of a 150 MW helical-coil IHX for a chloride-fueled MSR is prepared and the fission rate, capture rate, and inelastic scatter rate are evaluated

Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all Heavy Metal isotopes are to be recycled in the reactor. In this paper, an overview is presented of recent results obtained in the study of the closed fuel cycle and the influence of the addition of extra Minor Actinide (MA) isotopes from existing LWR stockpiles. In the presented work, up to 10% of the fuel was homogeneously replaced by an MA-mixture. The results are that addition of MA increases the potential of obtaining a closed fuel cycle. Reactivity coefficients generally decrease with increasing MA content. Addition of MA reduces the reactivity swing and allows very long irradiation intervals up to 10% FIMA with a small reactivity swing. Multirecycling studies show that a 600?MWth GCFR can transmute the MA from several PWRs. By a careful choice of the MA-fraction in the fuel, the reactivity of the fuel can be tuned to obtain a preset multiplication factor at end of cycle. Preliminary decay heat calculations show that the presence of MA in the fuel significantly increases the decay heat for time periods relevant to accidents (104–105 s after shutdown). The paper ends with some recommendations for future research in this promising area of the nuclear fuel cycle.Radiation, Radionuclides & ReactorsApplied Science

Analysis of a VVER-1000 in-core fuel management benchmark with DRAGON and DONJON

International audienceThis paper discusses the calculation of the IAEA benchmark entitled In-core fuel management code package validation for WWERs (IAEA-TECDOC-847) with the Version5 code system (DRAGON and DONJON). Calculations were performed for VVER-1000 cores. Cell calculations are done with DRAGON. Core calculations are done with DONJON, using CLE-2000 procedures to determine the reactor state at each depletion step, taking into account the control rod insertion pattern, reactor power, coolant inlet temperature and coolant mass flow. A multi-variate cross-section database for the core calculations is made with DRAGON. Calculated results are the critical boron concentration at each depletion step, and axial and radial power profiles at a few depletion steps. The present work serves as a feasibility study for the application of DRAGON and DONJON to VVER cores, and results are generally acceptable. Suggestions are made to improve the quality of the simulations