Power Plant Physics And Technology Research Articles

Strategical Approach for the Neutronics in the European Fusion Programme

One of the main aims of the European fusion programme is the design of a DEMOnstration fusion power plant (DEMO). The related work is conducted by the EUROfusion consortium and includes a strong supporting research and development programme. Support is also provided to the design of the high intense neutron source IFMIF-DONES (International Fusion Material Irradiation Facility-DEMO Oriented NEutron Source) to be built for the qualification of materials considered for DEMO. Neutronics plays a fundamental role for the design, operation, and safety of these facilities including the evaluation and verification of their nuclear performance. The lessons learned during the ITER design and construction phase point to the need to strengthen nuclear design integration already in the early DEMO design phase and establish an improved nuclear safety culture. This requires a coordinated approach for the neutronics that relies upon the availability of suitable computational procedures, tools, and data, qualified and validated for specific design- and safety-related applications. Accordingly, the approach builds on the development of advanced computation tools and the provision of high-quality nuclear data supported by integral experiments for their validation. Furthermore, configuration and requirement management principles ensure the alignment with the global nuclear design integration. This translates into appropriately chosen design margins and acceptance criteria, along with the specification of the nuclear analyses to be conducted in the various design phases. This paper presents the outlined approach as implemented in the EUROfusion Power Plant Physics and Technology (PPPT) programme and provides a strategical outlook of planned activities. This includes development works on advanced simulation tools with their application in various nuclear design- and safety-related analyses. The efforts to improve the nuclear database, in particular, with regard to radiation damage and activation cross-section data relevant to DEMO and DONES, are highlighted. Furthermore, the methodological approach applied to PPPT nuclear analyses including design, shielding, activation, and radiation dose calculations is discussed on the basis of specific examples.

Pre-conceptual design of EU-DEMO divertor primary heat transfer systems

In the frame of the activities promoted and encouraged by the EUROfusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently addressed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, the Divertor Primary Heat Transfer Systems (PHTSs) are intended to feed coolant to the two main components of the Divertor assembly, namely the Plasma Facing Components (PFCs) and the Cassette Body (CB). Since the DEMO Divertor must withstand high heat flux loads together with a considerable neutron deposited power, very tight tolerances may be allowed to the coolant inlet conditions. Therefore, the design of reliable PHTSs is of the utmost importance towards the development of an EU-DEMO fusion reactor. Within this framework, a study has been jointly carried out by University of Palermo, Ansaldo Nucleare and CREATE to design the Ex-Vessel PHTSs of both the PFCs and the CB for a DEMO reactor equipped with a Helium Cooled Pebble Bed Breeding Blanket. The paper describes criteria and rationale followed with the aim to achieve simple PHTS designs based on the adoption of easy-to-manufacture main components avoiding too much extrapolation from the state-of-art technology. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, pressurizers, piping and pumps are presented and critically discussed, with particular attention to those integration, safety and feasibility constraints that may deeply affect the design of such components. Finally, the evaluation of PHTS key parameters as total pumping power and coolant inventory is reported.

Comparative activation analyses for the HCPB breeding blanket in DEMO

The Helium Cooled Pebble Bed (HCPB) breeding blanket is one out of two concepts considered as driver blanket for the European DEMOnstration fusion power plant. The latest design, developed within the Breeding Blanket project of EUROfusion’s Power Plant Physics and Technology (PPPT) programme, assumes a fission-like breeder pin arrangement with breeder ceramics and Be12Ti as neutron multiplier.For the assessment of safety and waste management issues related to the use of the HCPB blanket in DEMO, a thorough activation analysis is required. Such an analysis should provide the total activity inventory and decay heat as well as the breakdown into components and materials as resulting from the irradiation over the assumed blanket lifetime in DEMO. An accurate calculation of the activity inventories requires the application of a suitable computational approach coupling the neutron transport simulation and the activation calculation with fully heterogeneous 3D geometry model of the DEMO.Precise materials specifications, including impurities and tramp materials, are essential prerequisites for adequate activation calculations. For the materials used in the HCPB blanket, the material compositions are provided with the material specifications set up for PPPT neutronics and activation calculations. These include the Eurofer steel, the breeder ceramics, tungsten and the beryllium in the Be12Ti used as neutron multiplier.Two difference tritium breeding material compositions of activation analyses were performed using each Li4SiO4+Li2TiO3 the latest material and previous material definition Li4SiO4. And beryllium titanate, the activation calculations were performed using the impurities specifications as provided by different potential beryllium suppliers. The most essential here is a uranium content in the ore used for the Be12Ti metallurgical engineering. This can permit a justification of the potential environmental risks associated with the application of the various materials options. The effect of the different Lithium compositions on the activity, both on the short- and long-term time scale, is evaluated in this work in the context of the radioactive waste management issue for DEMO.

Towards large area CVD diamond disks for Brewster-angle windows

Abstract In the frame of the EUROfusion Work Package Heating and Current Drive (WP HCD) of the Power Plant Physics and Technology (PPPT) program, CVD diamond disk Brewster-angle windows for gyrotron operation at multi-megawatt RF power levels and long pulses are under development. These windows allow for frequency step-tuneable operation. The Brewster-angle of 67.2° for diamond leads to an elliptical connection of the disk to the copper waveguides (WGs), requiring an advanced joining process. For proper transmission of the RF power, the disk consists of low loss CVD diamond of optical grade. The current target for the WG aperture of DEMO is 63.5 mm. It allows for an RF power transmission of 2 MW, but it requires a disk diameter of 180 mm for the 67.2° angle. In addition, a thickness of approximately 2 mm is needed to achieve the proper mechanical stability. State of the art microwave plasma reactors are not capable of growing disks of such size. The maximum available diameter of a polycrystalline CVD diamond disk suited to microwave applications is currently 140 mm. Thus, the industrial partner Diamond Materials GmbH (Freiburg, Germany) is doing extensive diamond growth experiments. A first of its kind, 180 mm thermal grade, crack-free, diamond disk was produced in the microwave plasma reactor with an average unpolished thickness of about 2 mm. First loss tangent measurements have been also performed. This paper describes the steps and the first results of this non-straightforward path, a challenging new field for diamond manufacturers and a major breakthrough for future frequency step-tuneable operation.

Shut-Down Dose Rate Distribution Calculation in the Access Cell With Activated HFTM in the IFMIF-DONES Facility

International Fusion Materials Irradiation Facility–DEMOnstration Power Station (DEMO)-Oriented Neutron Source (IFMIF-DONES) is an accelerator-based d-Li neutron source that is aimed at the qualification of materials at the irradiation conditions of a DEMO fusion power reactor. It is being developed in the frame of EUROfusion’s Power Plant Physics and Technology (PPPT) work program. The intense neutron radiation produced in the liquid lithium target results in a strong activation of the test cell (TC) and its high flux test module (HFTM). This article presents the analyses performed within the Early Neutron Source (ENS) project of EUROfusion/PPPT directed to provide up-to-date estimates of the activity inventories and the decay heat production in the DONES TC facility. To this end, a series of coupled Monte Carlo N-Particle code (MCNP)+ McDelicious transport and FISPACT inventory calculations were performed using the 2017 DONES TC model and nuclear cross section data from the Joint Evaluated Fission and Fusion (JEFF)–3.1.2 data library. The article presents the shut-down dose rate maps generated by HFTM that was irradiated at the TC and later relocated to the access cell. Performed calculations used to addresses the issue of the radiation dose loads that are to be expected due to the activated components/systems including the HFTM. In addition to that, activation calculations of HFTM materials are also provided.

Required, achievable and target TBR for the European DEMO

Abstract This work presents an up-to-date assessment of the TBR (Tritium Breeding Ratio) requirements for the European DEMO which is under development in the EUROfusion PPPT (Power Plant Physics and Technology) programme. Previous requirements were updated in accordance with the recent advances in the fuel cycle concept, the progress achieved with the DEMO design, and the additional requirement to generate the start-up tritium inventory for a follow-up fusion power reactor. The assessment results in a requirement of TBR ≥ 1.05 which is needed to provide the amount of tritium to the fuel cycle required to attain self-sufficiency. The TBR design target, which includes margins for calculation uncertainties and incomplete models, was set to TBR ≥ 1.15. The attainability of this design target is to be demonstrated with a neutronics calculation using an appropriate simulation model, without considering auxiliary systems, limiters, extra ports, etc. With the latest blanket design versions, this requirement can be fulfilled by the driver blanket candidates considered for DEMO, the HCPB (Helium Cooled Pebble Bed) and the WCLL (Water Cooled Lithium Lead) breeding blankets.

Nuclear data activities of the EUROfusion consortium

The activities of the EUROfusion consortiums on the development of high quality nuclear data for fusion applications are presented. The activities, implemented in the Power Plant Physics and Technology (PPPT) programme of EUROfusion, include nuclear data evaluations for neutron and deuteron induced reactions and the production of related data libraries which satisfy the needs for nuclear analyses of the DEMO fusion power plant and the IFMIF-DONES neutron source. The activities are closely linked to the JEFF initiative of the NEA Data Bank. The evaluation work is complemented by extensive benchmark, sensitivity and uncertainty analyses to check the performance of the evaluated cross-section data and libraries against integral experiments.

Progress in the design development of EU DEMO helium-cooled pebble bed primary heat transfer system

In the frame of the activities promoted and encouraged by the EURO-fusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently posed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, a very important role is played by the Breeding Blanket (BB) Primary Heat Transfer System (PHTS) as it is responsible to extract more than 80% of the fusion plasma power.In this framework, University of Palermo, Ansaldo Nucleare and CREATE have focused their work to improve thermal-hydraulic, safety and integration features of the Ex-Vessel PHTS for the Helium-Cooled Pebble Bed (HCPB) BB concept of DEMO.Starting from the outcomes obtained in 2016 that have allowed to highlight some criticalities which needed design changes, the paper describes progress and developments of the 2017 HCPB PHTS as well as the goals which have been achieved. The results of the research activity carried out show, in fact, i) an increase of the thermal-hydraulic performances, since a reduction in both total coolant inventory and total pressure drop has been respectively reached, ii) a potential improvement of the overall safety characteristics of the system. Nevertheless a critical assessment of these key parameters reveals that some issues are still open in terms of design integration and feasibility of the whole DEMO BoP for the helium-cooled blanket option, indicating that additional efforts are required to make this technology more attractive.

Neutronics studies for the novel design of lower port in DEMO

The conceptual design activity of the Demonstration Fusion Power Reactor (DEMO) is in progress in the Power Plant Physics and Technology (PPPT) programme, within the EUROfusion Consortium. In this work neutronics studies, fundamental for the nuclear design of DEMO, are presented for a novel design of the lower port (LP). Two possible configurations of the LP have been investigated: the vacuum pumping port, with the pumping unit located inside the port, and an empty port designed for remote handling. For both configurations 3-dimensional Monte Carlo calculations have been performed with MCNP5 to assess the neutron flux inside and around the port and the nuclear heating in sensitive components, such as the toroidal field coil conductor, the vacuum pumps, the shielding elements and the port closure plate. Different shielding configurations have been considered, by adding shielding blocks at the lower port entrance. Single and double wall port walls and closure plates with different thickness have been studied to reduce nuclear loads and neutron flux. Nuclear quantities under analysis were found to be within the limit for all the components, with the exception of the nuclear heating on the toroidal field coils. The absence of any shield of the divertor cassette pumping duct, as in the DEMO 2017 baseline configuration, is responsible for a huge amount of radiation streaming inside the pumping duct that can cause the excess of heating on the coil conductor. The use of a liner, or of an equivalent shielding component, is proposed, but further improvements are needed to keep the nuclear loads on the coil conductor within the limit.

Equatorial electron cyclotron port plug neutronic analyses for the EU DEMO

Within the Power Plant Physics and Technology (PPPT) programme in the EUROfusion Consortium design activities are currently in progress for the development of a DEMOnstration Fusion Power Plant (DEMO). The design of the machine and the integration of in-vessel components require neutronics analyses to verify the tritium self-sufficiency, the shielding requirements, and the structural integrity of its components. In particular, the effect of penetrations in the blanket and in equatorial port plug introduced by the electron cyclotron (EC) heating system, namely due to openings for the antenna waveguides (WGs), were analyzed. In this study three-dimensional MCNP calculations were conducted for the pre-conceptual designs of the EC port plugs and shielding optimization were performed in order to ensure that the DEMO design limits are not exceeded.Two configurations of the EC heating system were performed using a DEMO Water Cooled Lithium Lead (WCLL) with integrated EC configurations model of a half of the sector (i.e. 10° model) and relevant ones repeated with a full sector model (i.e. 20° model) to test the reliability of results. Additionally, the effect of the radiation on the WGs closest to the plasma was analyzed as well as the impact on Tritium Breeding Ratio (TBR).

Thermomechanical analysis supporting the preliminary engineering design of DONES target assembly

The design of the Demo Oriented NEutron Source (DONES) is the main objective of the Work Package Early Neutron Source (WPENS) of EUROfusion Power Plant Physics and Technology (PPPT) programme. DONES is an IFMIF-based neutron source with the goal of testing and qualifying candidate materials to be used in DEMO and future fusion power plants.In the framework of these activities, a research campaign has been carried out at the University of Palermo, in close cooperation with ENEA labs, in order to assess the thermomechanical performances of the DONES target system, endowed with an integrated Target Assembly (TA) when it undergoes the thermomechanical loads typical of its nominal operating scenario. The effects on TA behaviour of two possible different configurations for the accelerator beam foot-print have been investigated, considering an entire (20 × 5 cm2) or a halved (10 × 5 cm2) beam foot-print area with respect to that of IFMIF.The study has been carried out following a theoretical-numerical approach, based on the Finite Element Method (FEM), and adopting the qualified ABAQUS v. 6.14 commercial FEM code.The obtained thermomechanical results have been assessed in order to check whether the BP deformation field is such to prevent the contact of the BP itself with the High Flux Test Module, as well as to verify their compliance with the design criteria foreseen for the structural material. To this purpose, a stress linearization procedure has been performed along the most critical paths located within the BP, in order to check the fulfilment of the rules prescribed by the ITER structural design code SDC-IC. The obtained results are herewith presented and critically discussed.

The role of nuclear data for fusion nuclear technology

The role of nuclear data for neutronics analyses of Fusion Technology (FT) facilities is presented in this paper. Nuclear data evaluations of high quality are required for transport simulations, uncertainty assessments, activation and radiation damage calculations affecting the design and performance of the FT facilities, as well as safety, licensing, waste management and decommissioning issues. A corresponding programme on the nuclear data development (NDD) and qualification is conducted within the Power Plant Physics and Technology (PPPT) programme of EUROfusion supporting the development of the DEMO fusion power plant and the IFMIF-DONES neutron source. The programme builds on the achievements of pre-ceding NDD activities within the European fusion programme as addressed in the paper. Further needs for design, shielding, activation and radiation dose calculations are discussed, and recommendations are given to further improve and qualify the nuclear data base for PPPT nuclear analyses. This includes dedicated experimental activities tailored to the needs of DEMO and IFMIF-DONES.

Methodological approach for DEMO neutronics in the European PPPT programme: Tools, data and analyses

The methodological approach employed for the neutronics in the PPPT (Power Plant Physics and Technology) programme of EUROfusion is presented. It encompasses development works on advanced computational tools and activities related to the nuclear design and performance evaluation of the DEMO power plant including safety, maintenance, and waste management issues. Development work is conducted on Monte Carlo codes, on the CAD geometry conversion for Monte Carlo simulations, and on coupled radiation transport and activation computation systems. The role of nuclear data for reliable DEMO neutronics design analyses and uncertainty assessments is also addressed. Specific examples of nuclear analyses are presented including breeder blanket and shielding analyses for the different DEMO blanket concepts as well as related activation, decay heat and shut-down dose rate analyses.

Sensitivity and uncertainty analysis for the tritium breeding ratio of a DEMO fusion reactor with a helium cooled pebble bed blanket

An uncertainty analysis was performed for the tritium breeding ratio (TBR) of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB), currently under development in the European Power Plant Physics and Technology (PPPT) programme for a fusion power demonstration reactor (DEMO). A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design calculations, was employed. The nuclear cross-section data were taken from the JEFF-3.2 data library. For the uncertainty analysis, the isotopes H-1, Li-6, Li-7, Be-9, O-16, Si-28, Si-29, Si-30, Cr-52, Fe-54, Fe-56, Ni-58, W-182, W-183, W-184 and W-186 were considered. The covariance data were taken from JEFF-3.2 where available. Otherwise a combination of FENDL-2.1 for Li-7, EFF-3 for Be-9 and JENDL-3.2 for O-16 were compared with data from TENDL-2014. Another comparison was performed with covariance data from JEFF-3.3T1. The analyses show an overall uncertainty of ± 3.2% for the TBR when using JEFF-3.2 covariance data with the mentioned additions. When using TENDL-2014 covariance data as replacement, the uncertainty increases to ± 8.6%. For JEFF-3.3T1 the uncertainty result is ± 5.6%. The uncertainty is dominated by O-16, Li-6 and Li-7 cross-sections.

Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR)≥1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

Development of Sandwich Flow Channel Inserts for an EU DEMO Dual Coolant Blanket Concept

The dual-coolant lead-lithium (DCLL) blanket concept, which is considered as a candidate for fusion power plants and possibly for a demonstration reactor (DEMO), is being investigated within the framework of the European Power Plant Physics and Technology (PPPT) study. One of major issues of the DCLL concept development is the design of the flow channel inserts (FCIs), which are essential for the reduction of magneto-hydrodynamic (MHD) pressure losses. Due to the tight schedule for the short-term PPPT DEMO, a low-temperature DCLL concept with a liquid metal outlet temperature below 500 °C has been proposed. This allows the use of a simpler type of FCI (taking into account the LM corrosion issues), e.g. Eurofer-Alumina-Eurofer sandwich FCI, instead of the SiCf/SiC version for high temperature case, the production thereof is challenging. This paper discusses the technological study on manufacturing of some FCI design variants and post-examination of the samples.

Neutronics requirements for a DEMO fusion power plant

This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts.

Neutronic analyses and tools development efforts in the European DEMO programme

The European Fusion Development Agreement (EFDA) recently launched a programme on Power Plant Physics and Technology (PPPT) with the aim to develop a conceptual design of a fusion demonstration reactor (DEMO) addressing key technology and physics issues. A dedicated part of the PPPT programme is devoted to the neutronics which, among others, has to define and verify requirements and boundary conditions for the DEMO systems. The quality of the provided data depends on the capabilities and the reliability of the computational tools. Accordingly, the PPPT activities in the area of neutronics include both DEMO nuclear analyses and development efforts on neutronic tools including their verification and validation. This paper reports on first neutronics studies performed for DEMO, and on the evaluation and further development of neutronic tools.

Pre-conceptual studies and R&D for DEMO superconducting magnets

The DEMO plant will demonstrate by mid century the feasibility of electric power generation by nuclear fusion. Since 2011, conceptual design studies are coordinated by the EFDA Power Plant Physics and Technology (PPPT) Division, with the aim of identifying requirements, propose design approaches and start R&D for the magnet system of DEMO. The input and generic boundary conditions are given by the system codes: the major radius of the tokamak is about 9m. The proposed operating current at 13.6T peak field is 82kA, placing the DEMO TF conductor at substantially higher performance compared to ITER TF (68kA/11.5T). The innovative winding layout is a graded, layer wound with Nb3Sn/NbTi hybridization, aiming at minimizing the size and the cost of the superconductor. Two options are considered for the Nb3Sn conductor: one a “wind&react” cable-in-conduit (CICC) with reduced void fraction and rectangular shape. The other conductor is a “react&wind” flat cable with copper segregation and thick steel conduit assembled by longitudinal weld. The conductor designs were first drafted in 2012 and updated in 2013 based on a first round of assessments, which includes electromagnetic, thermal-hydraulic and mechanical analysis. The manufacture of full size prototype conductors is planned in 2014.The technical requirement of the DEMO superconducting magnets is highlighted in comparison to ITER and other fusion devices. The large size of the DEMO tokamak is the main challenge for the demonstration of the feasibility of power generation by fusion. Together with the technical issues, the cost of the superconducting magnets will be eventually the crucial aspect to promote the establishment of nuclear fusion as a primary energy source in the coming centuries.