Water Cooled Lithium Lead Blanket Research Articles

PbLi/Water Reaction: Experimental Campaign and Modeling Advancements in WPBB EUROfusion Project

The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is the interaction between PbLi and water caused by a tube rupture in the breeding zone, the so-called in-box LOCA (Loss of Coolant Accident) scenario. This issue has been investigated in the framework of FP8 EUROfusion Project Horizon 2020 and is currently ongoing in FP9 EUROfusion Horizon Europe, defining a strategy for addressing and solving WCLL in-box LOCA. This paper discusses the efforts pursued in recent years to deal with this key safety issue, providing a general view of the approach, a timeline, research and development, and experimental activities. These are conducted to master dominant phenomena and processes relevant to safety aspects during the postulated accident, to enhance the predictive capability and reliability of selected numerical tools, and to validate and qualify models and codes and the procedures for their applications, including coupling and chains of codes.

Analysis of the nuclear loads on Chromium monoblock divertor target for DEMO

The Plasma Facing Components (PFCs) of DEMO divertor play a fundamental role for the heat removal and particle exhaust functions, during fusion plasma. This implies a very harsh operation environment characterized by the combination of complex loading and stress conditions, intense particle bombardment, high heat flux deposition and a significant neutron irradiation. In the framework of the European DEMO divertor project (WPDIV), several target design concepts alternative to the baseline ITER-like technology are under evaluation, such as Chromium block design. It consists of chromium block, tungsten armor tile and copper alloy cooling pipe, and it is very attractive from an activation reduction, waste management and safety point of view. Therefore, the assessment of the most impacting nuclear loads for the PFC design is a pivotal aspect for DEMO divertor R&D program and this study is fully devoted to the comprehension of the spatial distributions of the main nuclear quantities (radiation damage, He-production and nuclear heating density) on the overall divertor targets. Neutronics analyses have been performed with MCNP5 Monte Carlo code using both DEMO MCNP Water Cooled Lithium Lead (WCLL) and Helium Cooled Pebble Bed (HCPB) blanket models, with a fully heterogeneous representation of the divertor cassette and Chromium PFCs geometry. The results of this analysis show that nuclear heating density and radiation damage distributions are higher with WCLL blanket than HCPB, except for He-production which seems not significantly affected by the blanket configuration. For each nuclear quantity, the peak values are achieved in correspondence of the baffle region and especially on the external units of the targets, as expected, being this region the most exposed to irradiation.

Neutronic Activity for Development of the Promising Alternative Water-Cooled DEMO Concepts

An emerging breeding blanket that fulfills performance criteria, meets the safety requirements, and is reliable enough to meet the plant availability is a challenging issue that assumes complex studies involving numerous neutronic analyses based on the Monte Carlo simulations with MCNP code. Two different concepts are now candidates to be implemented as a driver blanket for DEMO fusion reactor: WCLL (Water-Cooled Lithium Lead) and HCPB (Helium-Cooled Pebble Bed). The current R&D work within the EUROfusion DEMO project is concentrated on a search for the new water-cooled blanket layouts: a deep upgrade of the WCLL blanket to ensure a sufficient tritium breeding capability and an elaboration of the hybrid concept coupling technological advantages of water coolant, lead neutron multiplier, and ceramic breeder. To this end, very detailed, fully heterogeneous MCNP geometry models were developed for the newest designs of the WCLL-db (WCLL-double bundle) and WLCB (Water-cooled liquid Lead Ceramic Breeder) DEMO blankets to verify the new engineering solutions. This makes rigorous calculations possible to find an optimal breeder blanket layout. The basic response, tritium breeding ratio (TBR), was assessed for both concepts, and it appeared to be TBR = 1.16 for the WCLL-db and TBR ≤ 1.13 for the WLCB DEMOs, respectively. Several geometry layouts of the WLCB breeder blanket were investigated to reach the TBR sufficient for a sustainable tritium fuel cycle. Two promising novel solutions were suggested to enhance the tritium breeding performance of the WLCB blanket and to achieve TBR ≥ 1.16: heavy water coolant and an advanced breeder ceramic. Various nuclear safety aspects of the technologies utilized in both blanket concepts are addressed in this work to facilitate engineering decisions aimed at the consolidated blanket design for the upcoming DEMO reactor.

Evaluation of the thermal-hydraulic performances of a once-through steam generator in nuclear fusion applications

After decades of operation in nuclear power plants, Once-Through Steam Generators (OTSGs) were recently proposed for nuclear fusion applications. In particular, they are supposed to be installed in the primary cooling systems of the European Union Demonstration fusion power plant (EU-DEMO). One of the key reactor components is the Breeding blanket (BB). Among the BB concepts that are currently under study, Water-Cooled Lithium-Lead (WCLL) option was considered for this work. The WCLL blanket is divided in two main subsystems, the breeder zone (BZ) and the first wall (FW), each one provided with an independent cooling circuit, named Primary Heat Transfer System (PHTS). Thermal power removed from BB by BZ and FW PHTS is driven to the Power Conversion System (PCS) to be converted into electricity. The thermal coupling is ensured by two OTSGs per system. At the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of Sapienza University of Rome, a simulation activity was carried out to understand the component thermal-hydraulic behavior during DEMO normal operations. For calculation purposes, a full model of the steam generator was prepared by using a modified version of RELAP5/MOD3.3 system code. The computational activity performed allows to preliminary characterize the OTSG thermal-hydraulic performances during both pulse and dwell phases.

Activation Analysis and Evaluation of Radionuclide Inventory Decay Heat for EU DEMO Vacuum Vessel Components

Activation inventory data were assessed for different European DEMOnstration Fusion Power Reactor (EU DEMO) vacuum vessel (VV) segments in order to evaluate the activity, dose rates, and decay heat of the structural materials and the impact of heterogeneity as a part of the Safety and Environment project of the EUROfusion/Power Plant Physics Technology program. Activity is a good indicator of the condition of irradiated material while decay heat and dose rate estimates are crucial for power plant operation and public safety. In addition, the VV structures were analyzed in terms of different breeder blanket configurations that were implemented in the model. Helium-cooled pebble bed and water-cooled lithium-lead breeder blanket modules were considered. These modules directly shield the VV from neutron irradiation; therefore, their design has a great influence on subsequent structure activation.

Experiments on the MHD Effect on the Drainage of a LiPb Channel and Supporting Numerical Computations with the Level Set Method

To analyze the impact of the magnetohydrodynamics (MHD) effect on the fast draining of a LiPb channel (lithium-lead eutectic, 15.7 at. % Li) for a liquid metal fusion blanket such as the water-cooled lithium-lead test blanket system of ITER or DEMO, an experimental campaign was carried out with the support of the Integrated European Lead Lithium LOop experimental facility (IELLLO), installed at the ENEA Brasimone research center, Italy. The experiments were carried out by measuring the drainage time of the internal permanent magnet pump channel, normally used to circulate the LiPb in the loop, with and without the magnetic field. Moreover, this paper proposes a new numerical methodology to study the time delay induced by the MHD by using the commercial software COMSOL Multiphysics. In this way, it was possible to evaluate the LiPb fraction present at each time step in the computational domain and to estimate the time necessary for the complete drainage of the channel. The level set method was used to describe the transient behavior of the MHD flow under low-Rm approximation. The developed code was compared with the experimental results and showed good agreement, and it constitutes the first step in model validation as a possible application to ITER and DEMO. The experimental and numerical analyses performed in this work can be used as a benchmark case for MHD code development.

DEMO tritium breeding performances with different in-vessel components configurations

This article presents a global study of the tritium breeding performances in DEMO for the two breeding blanket concepts helium-cooled pebble bed (HCPB) and water-cooled lithium lead (WCLL). In order to consider the impact of cut-outs in the breeding blanket to integrated plasma limiters and auxiliary systems a set of 22.5° sector models was elaborated for diverse geometry configurations. The cut-outs in the breeder blankets were implemented making use of simplified CAD models for limiters (central inboard, upper, equatorial mid plane and equatorial lower limits) as well as for EC and NBI systems. These models were converted into the MCNP geometry representations making use of SuperMC code. The TBR calculations were performed with MCNP code for all geometry models. Interference effects due to non-symmetric distribution of limiters and auxiliary systems around the torus are shown to be negligible making use of a dedicated 360° model. Therefore the total TBR can be calculated considering the separate results regarding the impact of the different In-Vessel Components (IVCs) included in the 22.5° models. The TBR decrease due to the allocation of all currently considered limiters and auxiliary systems is ∼11% and ∼10% for the HCPB and the WCLL blanket respectively. Depending on the TBRtarget value adopted in the DEMO project (assumed range being 1.02–1.05) the required TBR can vary from TBRreq = 1.13÷1.16 accounting for the impact of the integration of limiters and auxiliary systems.

Optimization of the first wall cooling system for the DEMO WCLL blanket

• Thermal-hydraulic analyses of WCLL FW cooling systems in different poloidal position. • FW systems optimization to reduce the number of water channels, not exceeding the temperature limit of 550 °C. • Steady-state analyses to evaluate the cooling capability of the optimized systems. • Transient analysis to demonstrate the fully withstanding of the DEMO thermal loads. • Thermal-hydraulic aspects of FW designs are discussed and CHF evaluated. The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium production, shield the Vacuum Vessel and remove the heat generated by plasma thermal radiation and nuclear reactions. It relies on PbLi eutectic alloy adopted as breeder and neutron multiplier and refrigerate by subcooled pressurized water. The last function is fulfilled by two independent cooling systems: First Wall (FW) that faces the plasma heat flux and the Breeding Zone (BZ) that removes the deposited power of neutron and photon interaction. Several layouts of WCLL BB system have been investigated in the last years to identify a configuration that guarantees Eurofer temperature below the limit (550 °C) and good thermal-hydraulic performances. This research activity focuses on the FW cooling system based on the WCLL BB 2018 design of DEMO 2017 baseline, investigating the cooling performances in order to optimize the FW design, reducing the amount of water that affects the Tritium Breeding Ratio (TBR) and improves the efficiency of the Primary Heat Transfer System (PHTS), verifying the reliability to deliver coolant at adequate design temperature (328 °C) to the PHTS. Different solutions have been considered and analyzed focusing on three specific positions along the poloidal direction: the equatorial cells in the Central Outboard Segment (COB) and Inboard Segment (IB), where there is the maximum deposited power and low Heat Flux (HF), and the apical cell of the IB, impacted by the highest HF but with low deposited power. For each FW design, several thermal-hydraulic steady-state analyses have been performed, as well as sensitivity analyses to evaluate the response of the systems under different cooling configuration, changing the position and the number of channels. Furthermore, a transient analysis reproducing the Dwell-Pulse phase of the DEMO fusion reactor with the optimized configuration of the Outboard FW layout has been performed to study the behavior of the FW cooling system subjected to the pulsed operation. The research activity aims at laying the basis for the finalization of the WCLL BB design, pointing out relevant thermal-hydraulic aspects. The analyses have been carried out using a CFD approach, thus a 3D finite volume model of each configuration has been developed, adopting the commercial ANSYS CFX code. The goal of the study is to compare the different FW cooling system layouts, identifying and discussing advantages and key issues from the thermal-hydraulic point of view. The results show that the coolant system of FW can safely remove the high plasma HF and nuclear heat deposition with a reduction of water channels, delivering coolant to the FW PHTS heat exchanger at the design temperature.

Electromagnetic coupling phenomena in co-axial rectangular channels

In the Water-Cooled Lithium Lead (WCLL) blanket, the eutectic alloy lithium-lead (PbLi) is used as tritium breeder and carrier, neutron multiplier and heat transfer medium. The liquid metal is distributed to and collected from the breeding zone through a compact poloidal manifold composed of two co-axial rectangular channels. The external channel, tasked with distribution, and the internal one, assigned to the collection, are co-flowing and share an electrically conductive wall (cw=0.1). The liquid metal, interacting with the reactor magnetic field, leads to the arising of MagnetoHydroDynamic (MHD) effects that are expected to significantly modify the flow feature and electrically couple the external and internal channels. In this work, the general-purpose CFD code Ansys CFX 18.2 is used to study the coupling phenomena in a wide range of magnetic fields (up to Ha=2000) for a prototypical square co-axial channel. Characteristic flow features and their evolution with increasing magnetic field and varying mass flow rate between the channels are discussed and compared with the uncoupled case, which is in turn composed by a rectangular electro-conductive annulus (external) and a square electro-conductive duct (internal). A correlation is found linking the pressure loss in the studied configuration and an equivalent square channel through a corrective factor ε, which exhibits an asymptotic behavior for Ha > 1000.

MHD pressure drop estimate for the WCLL in-magnet PbLi loop

In the Water-Cooled Lithium Lead (WCLL) blanket, the eutectic alloy lithium-lead (PbLi) is used as tritium breeder and carrier, neutron multiplier, and heat transfer medium. The PbLi hydraulic loop section in the range of the reactor field coils, which includes the blanket and a non-negligible length of the connection pipes, is affected by intense magnetic fields which cause the transition to a MHD regime. Lorentz forces oppose the fluid motion and cause pressure losses several orders of magnitude higher than for the ordinary hydrodynamic regime. An accurate estimate of the MHD pressure drop is mandatory to properly design the PbLi loop and to optimize the flow path in the blanket. In this paper, the so-called “in-magnet” section of the PbLi loop is divided into three main regions (feeding and draining pipe, manifold, and breeding zone) which are further discretized into basic hydraulic elements. Analytical correlations and numerical results available in the literature are then used to calculate the MHD pressure drop terms for each element and then the overall loss. The study highlights that the highest contribution to pressure loss is made by the spinal manifold and connection pipes. Optimization strategies are suggested to minimize losses in these elements.

Long-term corrosion behavior of EUROFER RAFM steel in static liquid Pb-16Li at 550 °C

Liquid Pb-16Li eutectic alloy is the tritium breeding material in several breeding blanket concepts for European DEMO reactor, which are the Helium-Cooled Lithium-Lead (HCLL), the Water-Cooled Lithium-Lead (WCLL) and the Dual-Coolant Lithium-Lead (DCLL). RAFM (Reduced-Activation Ferritic/Martensitic) steels are considered as structural materials for breeding blankets in fusion reactors and EUROFER steel (9Cr-1.1W-0.2 V) is the European current choice for DEMO. Moreover, the WCLL European Test Blanket Module (TBM) to be installed in ITER will be manufactured in EUROFER. Here, the chemical compatibility of EUROFER in Pb-16Li is a major issue due to the dissolution of the material at high temperature in the liquid metal. In this work the corrosion of EUROFER in static Pb-16Li melt was investigated at 550 °C up to 8000 h of exposure to approximate low-flow WCLL blanket conditions. SEM-EDX analysis were performed on corroded samples to assess morphology and compositional changes occurring during the exposure. The corrosion was numerically evaluated by weight loss on samples exposed to various exposure times (2000, 4000 and 8000 h) and the corrosion rate for one year of exposure (8760 h) was extrapolated. The analysis by SEM-EDX showed that EUROFER mostly dissolved uniformly with a depletion of Cr and an enrichment of W at the near interface. The analysis of the weight loss provided an extrapolation of the corrosion rate of 18 μm/yr at 550 °C.

Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept

Abstract In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering selected parameters as a figure of merit to predict possible code outcomes. The uncertainty band will be evaluated through sensitivity analyses programmed, collected and statistically manipulated through RAVEN software tool. MELCOR and RAVEN have been internally coupled through a new Python code interface developed by Sapienza University of Rome, to perform sensitivity and uncertainty quantification analyses during severe accident transient. The Beyond Design Basis Accident (BDBA) analysis of an ex-vessel loss of coolant accident (LOCA) for the water-cooled lithium lead (WCLL) blanket concept has been simulated with the fusion version of MELCOR 1.8.6 code. The postulated initiating event (PIE) is a double-ended break in the first wall (FW) cooling system distributor ring, with simultaneous failure of the plasma shutdown system. An in-vessel breach of the coolant system occurs because of FW failure, with consequent unmitigated plasma shutdown transient. Sensitivity analysis results have shown that the FW temperature at which plasma in-vessel breach occurs is strongly correlated with the mass of hydrogen produced. The same parameter has also an impact on the overall accident scenario, such as the trigger of VVPSS rupture disks and thus source term mobilization.

Preliminary safety analysis of an in-vessel LOCA for the EU-DEMO WCLL blanket concept

Abstract In-vessel Loss Of Coolant Accident (LOCA) is one of the Design Basis Accident to be considered to support the future DEMOnstration power plant safety assessment. The water-cooled lithium-lead (WCLL) Breeding Blanket (BB) concept relies on Lithium-Lead as breeder, neutron multiplier and tritium carrier. The breeding modules are cooled by two independent pressurized water systems: the fist-wall (FW) and the breeding zone (BZ) coolant systems. The postulated initiating event (PIE) considered for this safety analysis is a double ended pipe rupture of the blanket module first wall channels. This event causes the inlet of coolant into the plasma chamber volume triggering an unmitigated plasma disruption and the pressurization of the Vacuum Vessel (VV) volume. The fusion version of MELCOR code (ver. 1.8.6) is used to evaluate accident consequences for two different scenarios, with the presence and absence of the downstream isolation valves, respectively. The chemical reaction between the coolant and the first wall tungsten layer inside the VV has been considered together with the mobilization of the radioactive source term. Pressure and temperature transient behavior in the tokamak volumes demonstrate that safety margins are respected during the accidental sequence.

Nuclear analysis of the Single Module Segment WCLL DEMO

The latest design of the Water Cooled Lithium Lead blanket for the European DEMO reactor is based on a Single Module Segment (SMS) concept for the blanket segments: the removal of the gaps between the modules (present in the former Multi Module structure design), with the consequent reduction of the neutron streaming, improves the tritium production and the shielding capabilities. A nuclear analysis has been carried out with the Monte Carlo N-Particle transport code version 5 (MCNP5) and the Joint Evaluated Fission and Fusion nuclear data libraries version 3.2 (JEFF 3.2). A detailed three-dimensional MCNP DEMO model with the Single Module layout has been generated. Three-dimensional neutron and gamma transport simulations have been carried out in order to assess the tritium production, nuclear power deposition in each subcomponent of the blanket and nuclear heating density, neutron flux, neutron damage (dpa) and Helium production, relevant for the design of the system. The results confirm the fulfilment of tritium self-sufficiency and shielding requirements of DEMO. Furthermore, a parametric analysis has been carried out by varying the thickness of the first wall tungsten layer to evaluate how the blanket performances are influenced by the tungsten layer thickness.

Status of Pb-16Li technologies for European DEMO fusion reactor

Three of the four breeder blanket concepts currently under investigation for the European DEMO Reactor use the eutectic Pb-16Li as breeder material. Those are the Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL) blankets. Moreover, the WCLL is one of the blanket concepts that will be qualified in the ITER reactor, therefore the development and design of lead lithium loops and auxiliary systems is essential. The main functional requirements that Pb-16Li systems have to fulfill are:•to circulate the liquid Pb-16Li through the blanket and ancillaries;•to extract the tritium produced inside the breeder modules from Pb-16Li;•to control Pb-16Li chemistry and to remove accumulated impurities;The present work aims to describe the activities performed in order to achieve the following objectives: i) design and integration of the Pb-16Li loops inside the tokamak building, ii) development and characterization of antipermeation and anticorrosion coatings on structures in contact with Pb-16Li, iii) development and design of an activation products removal system, iv) design of a chemistry control system for Pb-16Li loops, v) performing magnetohydrodynamic analyses taking into account the impact on heat transfer and tritium transport in breeding blankets and performing safety analyses of water/Pb-16Li interaction due to LOCA inside the WCLL blanket.

DEMO WCLL breeding zone cooling system design: Analysis and discussion

The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium self-sufficiency, shielding the Vacuum Vessel and removing the generated power of DEMO reactor. The last function is fulfilled by two independent cooling systems: First Wall (FW) that removes the plasma heat flux and Breeding Zone (BZ) that removes the deposited power of charged particle and neutron. Several layouts of BZ coolant system have been investigated in order to identify a configuration that guarantee Eurofer temperature below the limit (823 K) and good thermal-hydraulic performances (i.e. water outlet temperature 601 K). A research activity is conducted to study and compare four configurations, which rely on different arrangement of the stiffening plates (i.e. toroidal-poloidal and radial-poloidal), orientation of the cooling pipes (i.e. horizontal, vertical) and PbLi flow path. The analysis is performed using a CFD codes, thus a 3D finite volume model of each configuration is developed, adopting the commercial ANSYS CFX code. The objective is to compare the BZ cooling system layouts, identifying and discussing advantages and key issues from the thermal-hydraulic point of view, also considering feedbacks from MHD and neutronic analyses. The research activity aims at laying the groundwork for the finalization of the WCLL blanket design, pointing out relevant thermal-hydraulic aspects.

MHD mixed convection flow in the WCLL: Heat transfer analysis and cooling system optimization

In the Water-Cooled Lithium Lead (WCLL) blanket, a critical problem faced by the design is to ensure that the breeding zone (BZ) is properly cooled to avoid the loss of mechanical properties in the structural materials. CFD simulations are performed using ANSYS CFX to assess the cooling system performances accounting for the magnetic field effect in the sub-channel closest to the first wall (FW). Here, intense buoyancy forces (Gr ≈ 1010) interact with the pressure-driven flow (Re ≈ 103) in a MHD mixed convection regime. A constant magnetic field, parallel to the toroidal direction, is assumed with Ha = 8550. The walls bounding the channel and the water pipes are modeled as perfectly conducting. The magnetic field is found to dampen the velocity fluctuations triggered by the buoyancy forces and the flow is similar to a forced convection regime. The PbLi heat transfer coefficient is reduced to one-third of its ordinary hydrodynamic value and, consequently, hot-spots close to the FW are observed, where TMax ≈ 1000 K. Optimization strategies for the BZ cooling system layout are proposed and implemented in the CFD model, thus fulfilling the design criterion.

Chemistry and corrosion research and development for the water cooling circuits of European DEMO

The European DEMO design will potentially use single-phase water cooling in various components that require protection against corrosion. Coolant conditions will be similar to those of fission plant but with additional considerations arising from materials choices (Eurofer-97, CuCrZr) and 14 MeV neutron irradiation. Presently, many aspects of the water chemistry and corrosion behaviour are not well defined, and several strands of work, reported here, are ongoing to address these challenges under the EUROfusion framework in collaboration with industrial partners to leverage knowledge and expertise from elsewhere in the nuclear industry.Starting with the water-cooled lithium-lead blanket concept and considering the interaction of water with Eurofer-97, the foundation of this work has been the definition of a working water chemistry specification, supported by a review of relevant operating experience from light water fission reactors to understand the potential for technology transfer. Radiolysis modelling has been used to assess options for suppression of oxidising species under high energy neutron irradiation as these can be corrosive to components within the plant. High temperature water corrosion testing facilities have also been employed to expand the corrosion database and supplement existing experimental activities in the EUROfusion programme.In-vessel cooling of the divertor will use CuCrZr tubes under lower-temperature, high flow velocity conditions, which will lead to different considerations compared to the blanket and the potential for flow-accelerated corrosion. Additionally, high, unidirectional, heat fluxes lead to a radial temperature profile and the possibility of sub-nucleate boiling. A separate test setup, currently under construction, to expand this corrosion database is described.

WCLL Blanket Module Structure Variation Influence on Neutron Activation Inventories

One of the conceptual designs of the breeder blanket for a European demonstration fusion power station (DEMO) is the water-cooled lithium lead (WCLL) concept. The design development of the WCLL blanket in recent years has undergone the evolution of the neutronics model employing the MCNP code, with key changes in radial lengths and thickness as well as dimensional differences in inner structures such as breeding zone, shielding zones, and manifold. Furthermore, there has also been a significant increase in projected fusion power. Such changes were made with respect to its nuclear, thermohydraulic, and thermomechanical performances. In this paper, neutronic characteristics of WCLL modules were analyzed in terms of fusion power increase and dimensional changes of the blanket geometry presented in DEMO 2014 and 2015 models. For comparison, outboard and inboard blanket modules of the equatorial region were selected. Investigated neutronic characteristics include activity, decay heat, and contact dose rates. The numerical experiment was carried out with MCNP particle transport code and FISPACT/FISPACT-II activation calculation codes with EAF-2010/TENDL-2015 nuclear data libraries.

Measurements of neutron transport of well defined silicon filtered beam in lead

The correct description of neutron transport in lead is an essential task for correct description of tritium production in the DEMO (DEMOnstration Power Station) breeding blanket because some concepts deal with lead as a major component: namely the WCLL (water cooled lithium lead blanket), HCLL (helium cooled lithium lead blanket), and DCLL (dual cooled lithium lead blanket). Concerning the improvement of the knowledge about the transport of fast neutrons in lead, a set of experiments and calculations was carried out to study this problem with a well-defined neutron beam. The neutron flux behind various lead arrangements positioned along the beam axis was measured using a stilbene scintillation crystal (10 mm × 10 mm) with neutron and gamma pulse shape discrimination. The measurement was performed along the beam axis and in the case of the thick target also above the axis, to estimate the neutron angular scatter in lead. The calculations were realized using MCNP6 with various nuclear data libraries. Discrepancies in the angular distribution description in the energy region of about 1 MeV were discovered by these experiments.