1 INTRODUCTION
This section presents a short description of DEMO fusion reactor and the role of breeding blankets. Then, objectives of thesis activity are illustrated.
1.1 Main features of DEMO reactor
DEMO reactor will mark the very first step of fusion power into the energy market by supplying electricity to the grid. DEMO will largely build on the ITER experience. Beyond that:
• DEMO will breed its own fusion fuel tritium;
• DEMO will need materials suitable for handling the flux of neutrons produced in the fusion reactions.
To achieve fusion electricity by 2050, DEMO construction has to start in the early 2030s, immediately after ITER achieves the milestone of a net energy surplus. DEMO engineering design will become a major activity after 2020. The goal of DEMO will be to produce at least 2 GW of fusion power on a continual basis. Tritium self-sufficiency is mandatory for DEMO, which will burn about 0.3 kg of tritium per operational day. Tritium self-sufficiency requires efficient breeding and extraction systems to minimise tritium inventory. The choices of the materials and the coolant of the breeding blanket will have to be made consistently with the choice of the components for the transformation of the high-grade heat into electricity (the so-called Balance of Plant). Details about DEMO developing program are in Ref. [1].
As a typical magnetic confinement fusion reactor, DEMO is a Tokamak machine and is made by three main systems:
1) The vacuum vessel (VV) is a torus-shaped double-walled pressure vessel. It provides the primary vacuum and shields the magnet system from neutrons. It supports the in-vessel components and other systems:
• Breeding blanket (BB) with first wall (FW). These components have to fulfil several functions. In particular they have to assure self-sufficiency of the fusion reactor with regard to tritium (by producing, from lithium, at least the same amount of tritium as that which is consumed in the plasma), to maximize the net efficiency of the power plant (by assuring the highest possible temperature of the coolant), and to act as a radiation barrier (such that the components behind the breeding blanket receive the lowest amount of radiation possible) [5].
• Divertor. All charged particles present in the plasma are subject to a confinement effect caused by the magnetic field. Consequently, if no action is taken, Helium ion concentration within the plasma increases with the combustion rate. This effect must be limited because not only is the fuel diluted, but losses through radiation also increase. A specific device referred to as the divertor manages continuous extraction of reaction ash and, thus, maintains its concentration in the plasma at an acceptable level. It involves creating, by means of the poloidal magnet system, a final closed magnetic surface that stretches in two sections to the divertor. Beyond this separator, all magnetic surfaces are open and reach the wall. A particle first confined close to the magnetic axis moves slowly away by diffusion to reach the closed magnetic surface and will then encounter the divertor wall. Here, it is
neutralized and, thus, may be extracted by pumps that continuously empty this zone [6].
• Auxiliary heating systems. These systems face the plasma and radiate electromagnetic waves or neutral particles as deuterons into the plasma transferring energy to certain particles but also provide additional functions to support e.g. plasma break-down, heating to H-mode, MHD-control, current drive and others.
• Diagnostics. There is variety of different sensors installed. These measure plasma and magnetic field parameters mostly in the framework of the plasma control system.
• Fuelling system. It is located outside tokamak, forms pellets of frozen D, T, or D-T, accelerates these pellets, and guides them through a pipe up to the level of the FW. 2) The magnet system is an assembly of planar superconducting (SC) coils, which
provide the magnetic field required to break-down and confine the plasma, to drive its current and to define its poloidal structure. It is actively cooled by liquid helium at ~4K.
3) The Cryostat is a large, single-walled, passively cooled vacuum vessel at room temperature. It provides the vacuum required to operate the magnet system in cryogenic condition and supports the two backbone structures of the tokamak: the vacuum vessel and the TF coil system.
Tokamak of DEMO reactor with WCLL BB PHTS system connected is reported in 3D view in Fig. 1.1. Plasma parameters corresponding to DEMO 2015 concept are collected in Tab. 1.1.
Parameter Unit Value
Major radius, R0 m 9.072
Minor radius, a m 2.297
Aspect ratio m 3.101
FW/Blanket life years 19.06
Divertor life years 4.3
Plasma current MA 19.6
Toroidal Field, B0 at R0 T 5.667
Plasma volume m3 2502
Plasma surface area m2 1428
Auxiliary heating power MW 50 Average neutron wall load MWm-2 1.05 Nuclear heating in blanket MW 1822 Nuclear heating in shield MW 4.014
Fusion power MW 2037
Thermal power MW 2436
Net electric power MW 500
Net cycle efficiency -- 33.25% Tab. 1.1 – Main plasma parameters for DEMO 2015 concept (Ref.[4]).
1.2 Breeding blankets: generalities
The blanket absorbs the energy of the fusion neutrons and heats up a cooling fluid to drive the turbine for electricity production. It also ensures the tritium breeding process and shields the components outside the reaction chamber from the fast fusion neutrons. The reactor blanket contains lithium and a neutron-multiplying element (i.e. beryllium or lead). Tritium is extracted from the blanket and reprocessed.
Studies are ongoing on four main types of blankets [2]:
• Helium Cooled Pebble Bed (HCPB): He as coolant at pressure of 8 MPa, temperature from 300°C to 500°C, Li4SiO4 or Li2TiO3 as breeder and Be, both in form of pebble bed, T extraction in blanket with purge gas;
• Helium Cooled Lithium Lead (HCLL): He as coolant at pressure of 8 MPa, temperature from 300°C to 500°C, PbLi eutectic as breeder, T extraction from PbLi outside the reactor;
• Water Cooled Lithium Lead (WCLL): water as coolant at pressure of 15.5 MPa, temperature from 295°C to 328°C (i.e. PWR conditions), PbLi eutectic as breeder, T extraction from PbLi outside the reactor;
• Dual Coolant Lithium Lead (DCLL): He at pressure of 8 MPa and temperature from 300°C to 400°C and PbLi at maximum temperature of 500°C as coolant, PbLi eutectic as breeder, T extraction system outside the reactor.
Fig. 1.2 – BB under investigation for DEMO (Ref.[7]-[10]).
WCLL Breeding Blanket System, which is a candidate option for DEMO, is based on the use of reduced activation ferritic-martensitic steel EUROFER as structural material, liquid Lithium-Lead (PbLi) enriched at 90% in 6Li as breeder, neutron multiplier and tritium carrier, and water at typical PWR conditions (pressure 15.5 MPa, inlet temperature 295°C, and outlet temperature 328°C) as coolant. The WCLL Blanket Systems is constituted by the following subsystems:
• The Blanket segments, that can be further divided in:
- The blanket module, including the integrated FW, the Breeding Zone (BZ) and the internal manifold distribution system;
- The module’s Back Supporting Structure (BSS) providing mechanical attachments and insuring the distribution of all fluids to the modules, including the service pipes within the vessel port structures up to closure plate at the level of the cryostat, and supporting structures up to attachment directly to the VV or to intermediate shielding systems.
• The PbLi loop insuring the circulation of the breeder inside the modules and outside the VV for the tritium extraction purposes. Its auxiliary systems like the PbLi purification/chemistry control system are also included.
HCPB DCLL BB 2017 HCLL WCLL BB 2015
• The Tritium Extraction System (TES), which interfaces with the PbLi loop and with the Fuel Cycle main scopes of TES are:
- To extract tritium from the flowing lithium lead alloy in a dedicated sub-system; - To route it to the Tritium Plant for final processing.
The WCLL BB version 2016 is designed with the single module segment approach with a basic breeding cell element (Fig. 1.3) repeated along the poloidal direction, whose main dimensions can be seen in Fig. 1.4. The reference DEMO CAD model is divided in 18 sectors (20°) in toroidal direction. A blanket sector comprises three segments in the outboard blanket (OB) and two segments in the inboard (IB). The segments are separated by a gap of 20 mm. Thus, there is a total of 54 segments in the outboard blanket and 36 on the inboard blanket along the toroidal direction.
Fig. 1.3 - Section of the breeding cell element on toroidal-radial plane (Ref.[3]).
The FW of the segment (inboard and outboard) is single curved in poloidal direction, and its plasma facing area is covered with a tungsten layer. The FW is cooled by water flowing in square channels in counter-current direction along a radial-toroidal path.
Fig. 1.4 – Section of a poloidal plane of a breeding cell (Ref.[3]).
PbLi and water are collected by manifold located in the back wall of module (Fig. 1.5).
Fig. 1.5 – 3D view of BZ and FW inlet and outlet manifold (Ref.[3]).
In order to guarantees the structural integrity of each segments against the over pressurization, the WCLL BB segments (inboard and outboard) are equipped with internal stiffening plates placed along poloidal-radial (PR) and toroidal-radial (TR) planes (Fig. 1.6).
Fig. 1.6 – Stiffening plates on a WCLL BB module (Ref.[3]).
1.3 Framework of the thesis
European Union is committed in the fusion energy field with ITER and DEMO Projects. The EUROfusion Consortium is in charge of developing the DEMO conceptual design and R&D activities. ENEA belongs to the Consortium and the Experimental Engineering Division (FSN-ING) of ENEA C.R. Brasimone manages relevant and innovative experimental laboratories and facilities in support of nuclear R&D with several facilities related to fusion engineering development, such as LIFUS5 [11][12], LIFUS6[13], IELLLO[15], HeFUS3[14], TRIEX[16], THALLIUM[17], HYDREX[14].
This thesis, supported by Università di Pisa – Dipartimento di Ingegneria Civile e Industriale in collaboration with ENEA FSN-ING-PAN laboratory, is conducted in the framework of the Working Package Breeding Blanket (WPBB) of DEMO project, in thigh cooperation with the WCLL BB design team.
1.4 Objectives and structure of the thesis
The objective of the thesis is focused on the design of the experiments relevant for the validation of SIMMER-III code, and in particular of the lithium-lead water reaction chemical model implemented in the code. The objective is achieved by means of accomplishing the following tasks:
• implementation of the LIFUS5/Mod3 experiment (i.e. supporting the choice of the hydrogen system, dimensioning the heating system, testing the injection caps, and taking care of the construction phases);
• numerical simulations by SIMMER-III code in order to design the experimental campaign and the testing procedures;
• collecting and assembling the technical report of the facility, for attending and supervising the different assembling phases (e.g. workshop activities, etc.).
The overall activity is summarized in five sections. Among these, the first and the last are the introduction and the conclusions, respectively.
The introduction presents the EUROfusion Project, the DEMO Fusion Power Plant and the Breeding Blanket, with main reference to the Water Cooled Lithium Lead Breeding Blanket.
The second section addresses the reference postulated event, which is the “in-box-LOCA”, giving details on the thermal hydraulic and chemical phenomena relevant to safety and the simulation. A brief description of SIMMER code is also reported.
LIFUS5/Mod3 facility, including some design choices is presented in Section 3. This provides a description of the facility and of the different supporting activities performed during the design and construction phases.
Section 4, before the conclusions, presents the set-up SIMMER-III model and the pre-test analyses conducted for the design of the experiments, the layout of the injection line and the operative conditions, after the injection, relevant for the operation of the hydrogen measurement system.
1.5 References
[1] F. Romanelli et al., Fusion Electricity – A roadmap to the realisation of fusion energy, EFDA, November 2012, ISBN 978-3-00-040720-8.
[2] L.V. Boccaccini, EU blanket design and R&D for DEMO, 2nd EU-US DCLL Workshop, UCLA, Nov. 14-15TH, 2014.
[3] A. Del Nevo et al., Integration for WCLL-DDD 2016 for WCLL (update of DDD 2015), EUROfusion final report, 2017.
[4] DEMO1 Reference Design - 2015 April (“EU DEMO 2015”) – PROCESS One Page Output (2LBJRY v1.0)
[5] http://fusionwiki.ciemat.es/wiki/Breeding_blanket
[6] P. Magaud et al, Nuclear Fusion Reactors, Encyclopedia of Energy, Volume 4, Elsevier, 2004.
[7] F. Hernandez et al., A new HCPB breeding blanket for EU DEMO: Evolution, rationale
and preliminary performances, Fusion Engineering and Design, Article in press.
[8] A. Del Nevo et al., WCLL breeding blanket design and integration for DEMO 2015:
status and perspectives, Fusion Engineering and Design, Article in press.
[9] D. Rapisarda, DCLL BB design, DEMO BB Design and Analysis Methods, April 2017. [10] Jean-Charles Jaboulay et al., Nuclear analysis of the HCLL blanket for European
DEMO, Fusion Engineering and Design, Article in press.
[11] A. Ciampichetti et al., Water large leaks into liquid Pb-17Li: first experimental results
on LIFUS 5 facility, Fusion Eng. Des. 69 (2003) 563-567.
[12] A. Ciampichetti, et. al., Final report on TW2-TTBA-005-D1. Water large leaks into
liquid Pb-16Li: tests n. 6-7-8 on LIFUS5, LB-A-R-019, September 2003.
[13] A. Tincani et al., Realizzazione e qualifica dell’impianto sperimentale per prove di
corrosione/erosione in litio (LIFUS6), Report RdS, September 2013.
[14] M. Utili, Compatibility of structural materials under high temperature thermal flux
with Helium Cooling System for Fusion Reactor and Coolant Purification System,
MATISSE Workshop, 2015.
[15] M. Utili et al., European Breeding Blanket Test Facility: An Integrated device to test
European cooled TBMs in view of ITER, SOFT-25, September 2008.
[16] I. Ricapito, Tritium extraction from Pb16Li and He: EU experience and proposal, IEA Workshop on T/Pb16Li, Idaho Falls, June 2007.
[17] M. Utili et al., THALLIUM: An experimental facility for simulation of HCLL In-box
LOCA and validation of RELAP5-3D system code, Fusion Engineering and Design,
