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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Numerical investigation of two-dimensional buoyancy-driven eddies in liquid metal magnetohydrodynamic flows in breeding blankets *

Abstract Lead–lithium flows are key features in the design of tokamak breeding blanket concepts such as the dual-coolant lead–lithium (DCLL). Since they flow under magnetic fields, they are affected by magnetohydrodynamic (MHD) effects. The neutron flux originating in the tokamak plasma heats the breeding blanket channels in a non-uniform manner, inducing buoyancy forces in the liquid metal. Buoyancy may become a source of quasi-two-dimensional (Q2D) turbulence, and the appearance of eddies may affect the transport of heat and tritium across the blanket. Blankets characterized by high-speed liquid metal flows (such as DCLL) will need ceramic insulating walls to reduce the MHD-related pressure drop in the channels. In our simulations, we have used the Q2D model proposed by Sommeria and Moreau (SM82) which is especially suitable for modeling electrically insulating channel flows. On top of that, we have modelled buoyancy forces in the momentum equation using the Oberbeck–Boussinesq approximation. In this work, we include a validation of the implemented Q2D model in buoyancy-driven cases and identify a thermohydraulic configuration that promotes the generation of eddies and the accumulation of tritium. We also show the results provided by our post-processing tool based on the bi-dimensional fast Fourier transform for eddy detection and characterization. We complete our investigation by performing an initial assessment of how eddies can accumulate tritium in the breeding blanket. We conclude by discussing the relevance of preventing accumulation of tritium within the blanket and suggest a possible solution.

Suarez, Daniel (ORCID:0000000251740100)↗
DOE OSTI2016

Detection of Breeding Blankets Using Antineutrinos

The Plutonium Management and Disposition Agreement between the United States and Russia makes arrangements for the disposal of 34 metric tons of excess weapon-grade plutonium. Under this agreement Russia plans to dispose of its excess stocks by processing the plutonium into fuel for fast breeder reactors. To meet the disposition requirements this fuel would be burned while the fast reactors are run as burners, i.e., without a natural uranium blanket that can be used to breed plutonium surrounding the core. Here, we discuss the potential application of antineutrino monitoring to the verification of the presence of a breeding blanket. We found that a 36 kg antineutrino detector, exploiting coherent elastic neutrino-nucleus scattering and made of silicon could determine the presence of a breeding blanket at a liquid sodium cooled fast reactor at the 95 percent confidence level within 90 days.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2015

Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

Future fusion power plants based on the deuterium-tritium (DT) fuel cycle will be required to breed the T fuel via neutron reactions with lithium, which will be incorporated in a breeding blanket that surrounds the fusion source. Recent work by LLNL proposed the used of liquid Li as the breeder in an inertial fusion energy (IFE) power plant. Subsequently, an LDRD was initiated to develop alternatives ternary alloy liquid metal breeders that have reduced chemical reactivity with water and air compared to pure Li. Part of the work plan was to experimentally investigate the phase diagrams of ternary alloys. Of particular interest was measurement of the melt temperature, which must be low enough to be compatible with the temperature limits of the steel used in the construction of the chamber and heat transfer system.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Tritium Breeding Blanket for a Commercial Fusion Power Plant - A System Engineering Assessment

The goal of developing a new source of electric power based on fusion has been pursued for decades. If successful, future fusion power plants will help meet growing world-wide demand for electric power. A key feature and selling point for fusion is that its fuel supply is widely distributed globally and virtually inexhaustible. Current world-wide research on fusion energy is focused on the deuterium-tritium (DT for short) fusion reaction since it will be the easiest to achieve in terms of the conditions (e.g., temperature, density and confinement time of the DT fuel) required to produce net energy. Over the past decades countless studies have examined various concepts for TBBs for both magnetic fusion energy (MFE) and inertial fusion energy (IFE). At this time, the key organizations involved are government sponsored research organizations world-wide. The near-term focus of the MFE community is on the development of TBB mock-ups to be tested on the ITER tokamak currently under construction in Caderache France. TBB concepts for IFE tend to be different from MFE primarily due to significantly different operating conditions and constraints. This report focuses on longer-term commercial power plants where the key stakeholders include: electric utilities, plant owner and operator, manufacturer, regulators, utility customers, and in-plant subsystems including the heat transfer and conversion systems, fuel processing system, plant safety systems, and the monitoring control systems.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Initial Neutronics Investigation of a Chlorine Salt-Based Breeder Blanket

Tritium breeding blankets within D-T-fueled fusion reactors contain lithium compounds and typically require neutron multiplier materials to achieve a tritium breeding ratio (TBR) consistent with self-sustaining operation. Liquid breeder blankets have some advantages over solid blankets, and previous blanket studies have investigated liquid metal as well as liquid salt–based blankets. Liquid salts have reduced magnetohydrodynamic effects as compared to liquid metals, but typically have a lower TBR. Recently, advanced fission reactor concepts have considered chloride-based salts in their design, and there is a significant amount of research work occurring to study these salts. Chloride salts have previously been considered for fusion reactors, but studies have typically found lower breeding ratios than for fluoride salts, such as 2(LiF)-BeF 2 (flibe) so they have not been further developed. In this work, we use a one-dimensional cylindrical radiation transport model of a conceptual tokamak fusion reactor to investigate the neutronics feasibility of using a chloride salt–based blanket that uses chlorine enriched in 37 Cl, which has both a low capture cross section and a substantial (n,2n) cross section. It is found that chloride salts (LiCl mixed with BeCl 2 and/or PbCl 2 ) can potentially achieve a ~3% to 5% higher TBR than fluoride molten salts, notably flibe, in the absence of a solid multiplier. Including a solid multiplier, however, does narrow this advantage, with TBRs estimated within ~1% of flibe with a 2-cm Be multiplier. Chloride salts can also reach lower melting points than flibe, potentially improving the scope for the use of reduced activation ferritic-martensitic steel as a structural material. There is substantial uncertainty in the calculations driven by limited thermochemical data for the Cl salts, plus cross-section uncertainties. The production of 36 Cl through 35 Cl(n,g) and 37 Cl(n,2n) has the potential to challenge the waste disposal rating of the blanket. Calculations indicate that, while this is not an immediate showstopper, this case depends upon the exact waste disposal rating criteria used for 36 Cl. Further work could reduce these uncertainties with improved thermochemical data, higher-fidelity modeling for downselected salts, and more refined waste disposal calculations and regulatory guidance. Lastly, it must be recognized that, as for all molten salts, corrosion and chemistry can present appreciable technical challenges that require further assessment in developing a practical blanket concept, and also that the enrichment of chlorine presents an additional technical and supply chain challenge.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Methodology for accident analyses of fusion breeder blankets

'Fusion for Energy' (F4E) develops designs and implements the European Test Blanket Systems (TBS) in ITER - Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB). Safety demonstration is an essential element for the integration of TBS in ITER and accident analyses are one of its critical segments. A systematic approach to the accident analyses had been acquired under the F4E contract on TBS safety analyses. F4E technical requirements and AMEC and INL efforts resulted in the development of a comprehensive methodology for fusion breeding blanket accident analyses. It addresses the specificity of the breeding blankets design, materials and phenomena and at the same time is consistent with the one already applied to ITER accident analyses. Methodology consists of several phases. At first the reference scenarios are selected on the base of FMEA studies. In the second place elaboration of the accident analyses specifications we use phenomena identification and ranking tables to identify the requirements to be met by the code(s) and TBS models. Thus the limitations of the codes are identified and possible solutions to be built into the models are proposed. These include among others the loose coupling of different codes or code versions in order to simulate multi-fluid flows and phenomena. The code selection and issue of the accident analyses specifications conclude this second step. Furthermore the breeding blanket and ancillary systems models are built on. In this work challenges met and solutions used in the development of both MELCOR and RELAP5 codes models of HCLL and HCPB TBSs will be shared. To continue the developed models are qualified by comparison with finite elements analyses, by code to code comparison and sensitivity studies. Finally, the qualified models are used for the execution of the accident analyses of specific scenario. When possible the methodology phases will be illustrated in the paper by limited number of tables and figures. Description of each phase and its results in detail as well the methodology applications to EU HCLL and HCPB TBSs will be published in separate papers. The developed methodology is applicable to accident analyses of other TBSs to be tested in ITER and as well to DEMO breeding blankets.

Fusion Safety↗
DOE OSTI2016

Experimental Activities with the IELLLO Lithium-Lead Facility

Fusion power offers the prospect of an almost inexhaustible source of energy for future generations. Although this may be true, it also presents so far serious scientific and engineering challenges. Breeding Blanket is one of key components of future fusion reactors, since it directly involves tritium breeding and energy extraction, both of them critical to development of fusion power. To deal with the development of Breeding Blankets, ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) designed and built a lead lithium loop (IELLLO: Integrated European Lead Lithium Loop): the experiment aims to support R and D activities for the conceptual and preliminary design of the European test blanket systems. IELLLO was completed, installed and operated at ENEA Brasimone Research Centre. The purpose of this work is to describe the experimental activities that have been carried out concerning the qualification of the facility, and relative instrumentations, for lead lithium applications. A series of thermo-hydraulic investigations on IELLLO facility has been elaborated: lead lithium pressure transducers, heat exchanger, permanent magnet pumping system, mass flow meters and air cooler have been tested. The absolute pressure meter in contact with Pb-16Li was qualified in pressure range up to 10 bar and temperature up to 400 deg. C. The installation of new kinds of Pb-16Li differential pressure transducers will allow to check the pump characteristic curve and accurately calculate valves pressure drops. These factors could give a better information for the thermo-hydraulic qualification of the facility. Vortex Mass Flow Meter has shown good results in terms of accuracy. The performance has been validated by thermal balance with a dedicated heating system. Nevertheless, the installations of different types of mass flow meters will enable to test IELLLO at ITER relevant conditions, and at lower mass flow rates than those tested up to now. The new pumping system with a PM Pump, was qualified for the first time in Pb-16Li, and the procedure for the operation and maintenance of the pump was developed. The pump is able to supply a mass flow rate in the range from 1.2 to 2.4 kg/s with a temperature of Pb-16Li up to 550 deg. C in the test section. Significant results have been obtained also for the heat exchanger, which has shown high efficiencies, even more than expected. Likewise, the Air Cooler has been verified to be able to cool Pb-16Li in a relatively short time, even at the worst temperature and mass flow conditions. (authors)

42 ENGINEERING↗
DOE OSTI2014

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Neutron Measurement Instrumentation Development at KIT for the European ITER TBM

Fusion power reactors will rely on the internal production of the fuel tritium from lithium in the tritium breeding blanket. Test Blanket Modules (TBM) will be installed in ITER with the aim to investigate the nuclear performance of different breeding blanket designs. Currently there is no fully qualified nuclear instrumentation available for the measurement of neutron fluxes and tritium production rates which would be able to withstand the harsh environment conditions in the TBM such as high temperature (>400 deg. C) and, depending on the operation scenario, intense radiation levels. As partner of the European Consortium on Nuclear Data and Measurement Techniques in the framework of several F4E specific grants and contracts, KIT and ENEA have jointly studied the possibility to develop and test detectors suitable to operate in ITER-TBMs. Here we present an overview of ongoing work on three types of neutron flux monitors under development for the TBMs with focus on the KIT activities. A neutron activation system (NAS) with pneumatic sample transport could provide absolute neutron flux measurements in selected positions. A test system for investigating activation materials with short half-lives was constructed at the DT neutron generator laboratory of Technical University of Dresden to investigate the neutronics aspects. Several irradiations have been performed with focus on the simultaneous measurement of the extracted activated probes. An engineering assessment of a TBM NAS in the conceptual design phase has been done which considered issues of design requirements and integration. Last but not least, a mechanical test bench is under construction at KIT which will address issues of driving the activation probes, solutions for loading the system etc. experimentally. Self-powered neutron detectors (SPND) are widely applied in fission reactor monitoring, and the commercially available SPNDs are sensitive to thermal neutrons. We are investigating novel materials for SPND which would be sensitive also to the fast neutron flux expected in the TBMs. To this end simulations were done with the European Activation System EASY and neutron flux spectra which were calculated with MCNP for the HCPB TBM. Preliminary tests with commercial SPND in a fast reactor were performed. As a result of these activities, several materials have been found which may be suitable for the measurement of fast neutron fluxes in the TBM. Test detectors are under preparation for testing with DT neutron generators. Within the I{sub S}MART project, funded by KIC InnoEnergy, KIT is developing an online detector based on silicon carbide electronics for the TBMs. The operation of such detectors at TBM relevant temperatures is expected to incur lower accumulated radiation damage to them than at room temperature due to annealing effects. Detectors of several designs have been already irradiated with DT neutrons. Irradiation tests at elevated temperatures have been done and further tests are currently underway. This paper summarizes the status of the work for these three neutron flux monitor systems. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Design and Assessments of Liquid Metal Systems for a Fusion Power Reactor

This three-year research program at UCLA builds on the Fusion Engineering System Study (FESS) and has the major goal of identifying and accessing the most critical MHD/Mass & Heat Transfer phenomena associated with LM MHD flows in fusion breeding/cooling applications. The computed results and their analyses provide a strong basis to overcome present design limitations and will eventually lead to feasible, more effective solutions for LM systems of a fusion power reactor, such as breeding blankets and LM plasma facing components (PFCs). The project focuses on three research thrusts: Identification and characterization of the most important generic issues common to LM breeders (hereinafter called “Blanket”: ~47% of the resources); Design, implementation and assessments of two LM breeding-blanket concepts, DCLL and HCLL, for particular fusion devices, such as the US Fusion Nuclear Science Facility (FNSF) or the Fusion Pilot Plant (FPP) (hereinafter called “System Studies”: ~40% of the resources); Design and analysis of LM (Li) divertor concepts for FNSF or FPP, including scoping and detailed 2D and 3D computations (hereinafter called “LM PFC”: ~13% of the resources).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Numerical Simulation of Irradiation Induced Swelling for STAR 4.1 Blanket

Here, this article presents the preliminary results of FEM implementation of irradiation-induced swelling, hardening, and creep effect using ANSYS user programmable features (UPFs) applied to material properties. The first liquid metal breeding blanket model for STAR 4.1 tokamak as a part of the virtual prototyping system is under development at Princeton Plasma Physics Laboratory (PPPL). Intense neutron irradiation produces significant changes in the physical and mechanical properties of Fe-(8%–9%)Cr-based reduced activation ferritic martensitic (RAFM) steels. An ANSYS model of the breeding blanket was built to involve these irradiation-induced material property changes with neutron fluence distribution mapped from the MCNP model. Simulation results show that significant structure deformation forms from nuclear swelling, stress, and plastic strain arising from DPA gradient through wall thickness. The material degradation effect is not negligible. This modeling feature can either assist in analyzing the structural behaviors with the influence of nuclear swelling or provide guidance to design the structure to withstand irradiation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗