Molten salt breeding blanket: Investigations and proposals of pre-conceptual design options for testing in DEMO

In Europe, the DEMO test facility for Breeding Blankets (BB) will be the next step for fusion energy after ITER. The BB is a key component facing the plasma and ensuring tritium self-sufficiency, shielding against neutrons and heat extraction for electricity production. Within the EU research program, the two concepts candidates as driver BB in DEMO consider high-pressure coolants (water or helium) and Lithium-Lead or Beryllium pebble beds as breeder. In this article, an advanced BB concept is propounded as a candidate to be tested in DEMO taking into account Molten Salt (MS) as breeder and coolant and, consequently, named Breeder and Coolant Molten Salt (BCMS) BB. BCMS BB have several advantages, which lead to their investigations. As the MS is not pressurized during operation, high requirements associated to Nuclear Pressure Equipment (ESPN / PED) could be relaxed. Furthermore, the MS magnetohydrodynamic (MHD) flow is low due to its weak electrical conductivity. Besides, MS BB have identified drawbacks such as the structural material compatibility and the MS chemistry control needed during irradiation. A bibliographic review of MS and compatible materials is conducted. Promising BCMS BB are analysed regarding the Tritium Breeding Ratio (TBR) to test the capability of such MS breeders to reach the targeted TBR. The main nuclear quantities are evaluated with various MS on a dedicated model via the TRIPOLI-4® Monte-Carlo code: TBR, nuclear heating, neutron flux, displacement damage and helium production are reported and discussed. Moreover, the thermal and thermo-hydraulic responses of the MS BB are evaluated regarding the pressure drops (ΔP), the Pumping Power (PP) and the maximum temperatures in the structures. Nonetheless, preliminary design options are investigated. Finally, the major purpose is to evaluate the feasibility of a BCMS BB and identify their advantages and drawbacks

Nuclear Analysis of the HCLL "Advanced-Plus" Breeding Blanket with Single Module Segment Structure

International audienceThis paper presents the nuclear analysis carried out with the TRIPOLI-4$^R$ Monte-Carlo code for the Helium Cooled Lithium Lead (HCLL) Advanced-Plus breeding blanket design using the Single Module Segment (SMS) option in the European DEMO 2017 baseline. Compared to the previous one, this baseline is characterized by a radial outboard breeding blanket thickness reduction of 30 cm. Previous study has quantified its impact on Tritium Breeding Ratio (TBR reduction up to -0.08). This major constraint lead to the need of SMS solution development with HCLL Advanced-Plus design to reduce the amount of steel in the breeding blanket for TBR improvement. HCLL Advanced-Plus design is currently developed with the aim to improve the TBR. The main nuclear quantities neutron wall loading, TBR, nuclear heating, neutron flux, displacement damage and helium production are reported and discussed

DEMO Breeding Blanket Helium Cooled First Wall design investigation to cope high heat loads

In the framework of the European “HORIZON 2020” program, EURO fusion develops a fusion power demonstrator (DEMO). According to a recent study about plasma heat loads, it appeared that the First Wall, which is the first part of the Breeding Blanket in front of the plasma, will face some high heat fluxes on some poloidal locations of the Breeding Blanket.In order to cope such high heat fluxes, this paper presents the investigation on different Helium cooled First Wall design integrated to the Breeding Blanket on the basis of standard square and circular smooth channels and with different options for the Tungsten armor surrounding the channels, taking advantages to the different material properties. The performances of the different concepts have been assessed with thermal and mechanical Finite Element Method numerical simulation based on a slice of the Helium Cooled Lithium Lead concept. Results are compared with the RCC-MRx design rules to prevent failure during normal and accidental condition. The results show that the options with channels surrounded by Tungsten could meet some plasma heat loads requirements from design point of view. However, the concept is still at an early stage of development and open issues are discussed

Nuclear analysis of the HCLL blanket for the European DEMO

This paper presents the nuclear analysis of the European DEMO baseline 2015 with HCLL blanket carried out with the TRIPOLI-4® Monte Carlo code and the JEFF-3.2 nuclear data library. The TRIPOLI-4® model was imported from CAD using the McCad tool. A procedure that generates the detailed 3D model describing all the HCLL blanket internal structures was developed. This procedure allows parametrization of the blanket internal structures such as the number of cooling plates, manifolds, etc. and the thickness of the stiffening grid for instance. Different design variants were studied to improve the tritium production. From this previous study a complete nuclear analysis was carried out on a promising design which is a compromise between tritium production and mechanical robustness. All criteria (TBR, nuclear heating in coils and displacement damage in vacuum vessel) are met using this new reference design

Comparison over the nuclear analysis of the HCLL blanket for the European DEMO

This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4® Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4® representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4® Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO

Investigation on the “advanced-plus” Helium Cooled Lithium Lead Breeding Blanket design concept for TBR enhancement regarding thermal and mechanical behavior

In the framework of the European “HORIZON 2020” research program, the EUROfusion Consortium develops a design of a fusion demonstrator (DEMO). The Breeding Blanket (BB) directly surrounding the plasma is a major component ensuring tritium self-sufficiency, shielding against neutrons from D-T plasmas and heat extraction for electricity conversion. CEA-Saclay, with the support of Wigner-RCP and IPP-CR is responsible of developing the Helium Cooled Lithium Lead (HCLL) BB concept. In order to enhance Tritium Breeding Ratio (TBR), an alternative HCLL BB design, called “Advanced-Plus” concept, has been investigated, with the aim to reduce the amount of steel, while ensuring good thermal and mechanical behavior. The thermal behaviors of the horizontal Stiffening Plates (hSPs) with various channels designs in terms of dimensions and layout have been parsed on simplified FE models for verifying and optimizing the HCLL Breeding Zone (BZ) before analyzing a HCLL BB ¼ model. Moreover, sensitive analyses have been carried out for testing various hSPs pitches, alternative Helium hydraulic schemes inside the hSPs as well as the detachable FW option in applying a zero Heat Flux (HF) on the FW. Furthermore, mechanical calculations have also been performed to assess the behavior of the module in faulted condition. The results obtained with the Cast3M-qualified-FEM code are presented and discussed

Nuclear analysis of the HCLL blanket for the European DEMO

International audienceThis paper presents the nuclear analysis of the European DEMO baseline 2015 with HCLL blanket carried out with the TRIPOLI-4® Monte Carlo code and the JEFF-3.2 nuclear data library. The TRIPOLI-4® model was imported from CAD using the McCad tool. A procedure that generates the detailed 3D model describing all the HCLL blanket internal structures was developed. This procedure allows parametrization of the blanket internal structures such as the number of cooling plates, manifolds, etc. and the thickness of the stiffening grid for instance. Different design variants were studied to improve the tritium production. From this previous study a complete nuclear analysis was carried out on a promising design which is a compromise between tritium production and mechanical robustness. All criteria (TBR, nuclear heating in coils and displacement damage in vacuum vessel) are met using this new reference design

Rigorous-two-Steps scheme of TRIPOLI-4® Monte Carlo code validation for shutdown dose rate calculation

International audienceAfter fission or fusion reactor shutdown the activated structure emits decay photons. For maintenance operations the radiation dose map must be established in the reactor building. Several calculation schemes have been developed to calculate the shutdown dose rate. These schemes are widely developed in fusion application and more precisely for the ITER tokamak. This paper presents the rigorous-two-steps scheme implemented at CEA. It is based on the TRIPOLI-4® Monte Carlo code and the inventory code MENDEL. The ITER shutdown dose rate benchmark has been carried out, results are in a good agreement with the other participant