Multiobjective optimization for nuclear fleet evolution scenarios using COSI

The consequences of various fleet evolution options on material inventories and flux in fuel cycle and waste can be analysed by means of transition scenario studies. The COSI code is currently simulating chronologically scenarios whose parameters are fully defined by the user and is coupled with the CESAR depletion code. As the interactions among reactors and fuel cycle facilities can be complex, and the ways in which they may be configured are many, the development of optimization methodology could improve scenario studies. The optimization problem definition needs to list: (i) criteria (e.g. saving natural resources and minimizing waste production); (ii) variables (scenario parameters) related to reprocessing, reactor operation, installed power distribution, etc.; (iii) constraints making scenarios industrially feasible. The large number of scenario calculations needed to solve an optimization problem can be time-consuming and hardly achievable; therefore, it requires the shortening of the COSI computation time. Given that CESAR depletion calculations represent about 95% of this computation time, CESAR surrogate models have been developed and coupled with COSI. Different regression models are compared to estimate CESAR outputs: first- and second-order polynomial regressions, Gaussian process and artificial neural network. This paper is about a first optimization study of a transition scenario from the current French nuclear fleet to a Sodium Fast Reactors fleet as defined in the frame of the 2006 French Act for waste management. The present article deals with obtaining the optimal scenarios and validating the methodology implemented, i.e. the coupling between the simulation software COSI, depletion surrogate models and a genetic algorithm optimization method

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle

Prospects in China for nuclear development up to 2050

International audienceAn ambitious plan for nuclear development exists for long in China. The study of scenarios with prospect of 150 GWe to 400 GWe in 2050 is carried out using the COSI6 simulation software, and aims at analyzing the evolution of nuclear energy currently planned in China. Results rely on natural uranium supplies, fuel fabrication, spent fuel reprocessing, quantities of proliferating materials and the opportunity of a rapid deployment of fast reactors (FBR). It seems impossible for China to start two fast reactors before 2020 without any external source of plutonium. Anyway, FBR may represent at the most around 30% of the total nuclear capacity in the country by 2050. Indeed the deployment of FBR only can start from 2035 to 2040. Finally, the pace of FBR development should be controlled carefully by the proportion of FBR and PWR with respect to the reprocessing capacity. Natural uranium savings appear rather low by 2050, because the transition toward a fast reactor fleet independent from uranium ore lasts decades. However, this independence may be reached by the end of the century, before uranium resources are dwindling, if the first corners to close the fuel cycle are turned in the next decade

FRENCH SCENARIOS TOWARD FAST PLUTONIUM MULTI-RECYCLING IN PWR

International audienceIn France, the COSI6 software can simulate prospective scenarios of nuclear energy evolution. Nuclear scenarios focused these last years on the development of SFR technology. However, SFR are more expensive to build than thermal reactors. In case SFR would not become economically competitive in the next decades, MOX spent fuels would pileup in the backend of the fuel cycle, unless alternative solutions of plutonium management in PWR were found. In this study, advanced EPR (European Pressurized water Reactor) fuel designs are applied to enable plutonium multi-recycling and stabilization of all spent fuel: CORAIL refers to fuel assemblies containing LEU and MOX rods, and MIX (also called MOXEUS) to assemblies where fuel rods are composed of plutonium mixed with enriched uranium. Scenarios results reveal that introducing MIX and CORAIL in EPR by the middle of the century can lead to a fast stabilization of spent fuel and plutonium inventories. With respect to open cycle, more minor actinides (MA) accumulate (about +70%), but the production of transuranic elements (Pu + MA) remains almost 3 times less. Furthermore, all high-level wastes are now packaged for long-term storage. Besides, spent fuels still contain significant quantities of fissile uranium. In MIX scenarios however, this uranium may be enriched and easily recycled into dedicated EPR for efficient natural uranium savings. In this case, the resource balance is significantly better than in open cycle (-30%). Multi-recycling in PWR appears therefore to be a viable temporary solution, allowing for spent fuels and wastes management until we expect the running out of natural uranium

A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics

Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the ‘C-lite’, is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results

Verification of dose rate calculations for PWR spent fuel assemblies

International audienceUnder the framework of the Nuclear Energy Agency's Expert Group on Advanced Fuel Cycle Scenarios (NEA/AFCS), a benchmark on dose rate calculations for Pressurized Water Reactor (PWR) Spent Fuel Assembly is currently underway. This multinational effort was first proposed by CEA (French Commissariat a l'Energie Atomique et aux Energies Alternatives) and DOE (U.S. Department of Energy), after having conducted their own bilateral comparative study on dose rate calculations for typical UOX and MOX spent fuel assemblies [1]. The goals of this benchmark are to expand on that work by including more international participants, to verify the dose rate results and potentially include validation efforts depending on the availability of appropriate experimental data

A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics

International audienceRadiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the ‘C-lite’, is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results