Helium Cooled Pebble Bed Test Blanket Research Articles

Preliminary Experimental Quantification of Helium Leakages from Flanged Connections at HCPB TBS Operative Conditions

The HCPB TBS (Helium-Cooled Pebble Bed Test Blanket System) is one of the two European TBSs that will be installed and tested in the ITER reactor. The use of flanged connections in the Helium Coolant System and the Tritium Extraction System of the HCPB TBS would make the remote maintenance operations easier and faster. Therefore, investigating the helium leakage from flanges becomes a fundamental step toward the control of the tritium activity in the Port Cell, as the helium flow will contain a variable but not negligible amount of tritium. The first set of experiments on helium leakages from flanged connections is described in this paper. The experiments were performed in a HeFUS3 facility, an eight-shaped helium loop designed to work at HCPB-TBS-relevant conditions. The facility can provide a helium mass flow rate in the range of 0.27–1.4 kg/s and can reach a pressure as high as 80 bar and a temperature up to 530 °C. Two types of gaskets were tested in this campaign: a spiral-wound gasket and an oval ring joint. The gasket/flange assemblies are described in detail in this paper, together with the test section that hosts them and the performed commissioning tests. The tests were carried out at 500 °C and 80 bar. In these conditions, the leak rate from the flange with the oval ring joint resulted in being, on average, 1.42·10−6 mbar∙L/s, while the leak rate from the flange with the spiral-wound gasket resulted in being, on average, 3.73·10−3 mbar∙L/s.

Advances in Additive Manufacturing of fusion materials

Recent analysis result in significant impact on the design of the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB). Especially the main demanding sub-component of the BB in terms of fabrication, the First Wall (FW) is highly affected. As a matter of fact, the relevance of the developments made for the industrial fabrication of the FW used e.g. in the EU HCPB Test Blanket Module (TBM) for ITER is limited: On one hand in terms of the overall dimensions, but also in terms of geometry and cooling channel configuration. Thus, a fabrication strategy for the FW was proposed as an option in 2018 based on Additive Manufacturing (AM) offering a solution to cover the new aspects. The dedicated fabrication routine is based on Cold Spray (CS) in alternation with machining. In terms of manufacturing of non-plasma facing complex shaped thin- or double wall BB structures (e.g. Fuel Pins proposed for the HCPB BB) the use of AM also provides benefits compared to conventional technologies. Therefore, the AM process of Selective Laser Melting (SLM) is also investigated in terms of applicability for BBs. This paper gives an overview about the advances in different AM options applied for nuclear fusion structural low activation steels. Possibilities for spin-offs to other technological fields are discussed and conclusions are drawn reflecting licensing aspects and technological limits.

Preliminary RAMI assessment for HCPB and HCLL ITER test blanket module ancillary systems

A RAMI (Reliability, Availability, Maintainability and Inspectability) assessment performed on the ITER Test Blanket Module (TBM) ancillary systems is presented. The assessment is aimed at evaluating design that may jeopardize the achievement of the overall 75% availability requirements for the TBM plant breakdown structure. The ancillary systems of the European Test Blanket Modules to be operated in ITER that have been analyzed in this assessment are the helium cooling systems (HCSs) of both the Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB), the lithium-lead loop of the HCLL TBM and the Tritium Extraction System (TES) of the HCPB TBM. RAMI performance was assessed by means of reliability block diagrams (RBDs) over a foreseen mission of 20 years, with operating cycles reflecting ITER schedule. In particular, a set of events leading to unavailability of the systems was initially defined by means of a failure mode and effect analysis. Then RBDs were implemented according to reliability-wise integration of the components included in such systems. Different RBDs were defined depending on the component being judged to impact normal operation or start-up operation mode. Systems analysis was performed exploiting a modular approach in order to elicit the relative contribution to overall availability so to support possible design improvement by highlighting critical sub-systems. In particular the unavailability of cooling, tritium extraction and trapping functions were assessed separately for HCLL and HCPB concepts. Finally the integrated impact of all ancillary systems on overall tokamak operational availability was estimated and the assessed systems, considering the current preliminary level of design development, appear able to meet the expected performance.

Conceptual design of the enhanced coolant purification systems for the European HCLL and HCPB test blanket modules

The Coolant Purification Systems (CPSs) is one of the most relevant ancillary systems of European Helium Cooled Lead Lithium (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Modules (TBMs) which are currently in the preliminary design phase in view of their installation and operation in ITER.The CPS implements mainly two functions: the extraction and concentration of the tritium permeated from the TBM modules into the primary cooling circuit and the chemistry control of helium primary coolant.During the HCLL and HCPB-TBSs (Test Blanket Systems) Conceptual Design Review (CDR) in 2015 it was recognized the need of reducing the tritium permeation into the Port Cell #16 of ITER. To achieve this and, then, to lower the tritium partial pressure in the Helium Cooling Systems in normal operation, the helium flow-rate treated by each CPS has been increased of almost one order of magnitude.In 2017, to satisfy the CDR outcomes and the new design requirements requested by Fusion for Energy (F4E, the European Domestic Agency for ITER), ENEA performed a preliminary design of the “enhanced” CPSs.This paper presents the current design of the “enhanced” CPSs, focusing on design requirements, assumptions, selection of technologies and preliminary components sizing.

A new HCPB breeding blanket for the EU DEMO: Evolution, rationale and preliminary performances

The Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) is one of the 4 BB concepts being investigated in the EU for their possible implementation in DEMO. During 2014 the former “beer-box” BB concept based on the ITER’s HCPB Test Blanket Module suffered several design changes so as to meet the different counteracting nuclear, thermohydraulic and thermomechanical requirements. These studies evidenced that the concept is too rigid to meet the tight TBR requirements imposed for the EU DEMO (i.e. TBR≥1.10). Additionally, the complex manifold system with unbalanced helium mass flow in each of the 2 parallel cooling loops made the concept thermohydraulically complex. However, parametric studies during 2015 revealed that the HCPB concept have potential for a better nuclear performance, as well as margin for a significant simplification of the cooling internals by redefining the cooling plates and the architecture of the blanket, building a symmetric flow scheme.This paper describes the new HCPB concept based on an integrated FW with the breeding zone thermohydraulics and helium manifold systems. The former complex manifold backplates have been compacted and integrated in the cooling plates, releasing ≈300mm of radial space that can be used now to increase breeder zone, the neutron shielding, to reinforce the Back Supporting Structure (BSS) or basically to reduce the reactor size. Detailed neutronic analyses have yielded a TBR of ∼1.20 for the baseline design. Initial analyses show a correct thermohydraulic behavior. Preliminary thermomechanical analyses also indicate that the design can potentially withstand an in-box LOCA at 9 MPa at a level C according to the RCC-MRx code. Future consolidation activities are described, which shall lead to a concept meeting the BB requirements.

Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design.For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of high dose irradiation on it.

Two Different Medium-Scale First Wall Fabrication Experiments at KIT for ITER Helium-Cooled Pebble Bed Test Blanket Module

Subcomponent manufacturing and assembly concepts for the fabrication of the helium-cooled pebble bed test blanket module (TBM) for ITER have been developed over more than one decade at KIT, in particular the first wall (FW), which is a key element for the TBM fabrication. The design of this subcomponent foresees the manufacturing of a large U-bended plate of EUROFER with built-in channels for helium cooling. Manufacturing technologies developed at KIT are based on diffusion welding of two half-plates as the most promising option. This paper deals with the manufacturing of two medium-scale TBM FW mock ups according to two different industrial processes: a uni-axial diffusion welding process realized in a mechanic press at high temperature and a hot isostatic pressing process applied to a canned assembly at relatively low pressure.The qualification of the welds produced is described, and the results are compared to previous small- and medium-size scale experiments. The results of the recent FW fabrication mock ups are presented with regard to material data (e.g., ultimate strength, ductile-brittle transition temperature) and TBM-relevant parameters (e.g., deformation of cooling channels). The paper concludes with an overview of the strategy to evolve from 1/8th-scale experiments to TBM-relevant dimensions.

Thermo-mechanical analyses and assessment with respect to the design codes and standards of the HCPB-TBM Breeder Unit

The Helium Cooled Pebble Bed-Test Blanket Module (HCPB-TBM) is one of the two breeding blanket concepts currently under development in Europe. Key component of the HCPB-TBM, the Breeder Unit (BU), has entered the detailed engineering design phase. After establishing a base design, thermal and thermo-mechanical analyses have been performed under typical ITER operational conditions: the results are presented and discussed in this paper.A full scaled finite element model of the base design of the HCPB-TBM BU has been built to run thermal analyses of the beryllium and Li4SO4 pebble beds and thermo-mechanical analyses of the BU structure, both in steady state and in a typical transient regime during a pulse of the ITER D–T phase. The temperatures reached in the Li4SO4 and beryllium pebble beds in the BU base design are 930.8°C and 712.9°C, respectively, which are above the recommended values of 920°C for Li4SO4 and 650°C for the beryllium pebbles. The maximum temperature in the structural steel is 548.4°C, which remains under the design limit defined for the TBM studies (550°C). In order to decrease the temperature in the hot spots identified in the pebble beds, a reduction of the Li4SO4 and beryllium bed volumes has been adopted. As for the structural material, the thermo-mechanical analyses have been assessed with respect the RCC-MR design code (completed for irradiation damages with ITER SDC-IC). The results reveal some problematic points in the base design, concentrated in the coolant inlet and outlet pipes and in the connection region of the BU cooling plates with the BU backplate. Submodeling technique has been used to improve the design in these regions. An increase in the thickness of the coolant inlet and outlet pipes and a redesign of the BU backplate have led to a fulfillment of the codes and standards. The design modifications of pebble bed region and structural material have been implemented in the final design of the BU that is presently used as reference for the design and test of a BU mockup in KIT.

Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

Abstract The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.

The coolant purification system of the European test blanket modules: Preliminary design

The HCPB (Helium Cooled Pebble Bed) and HCLL (Helium Cooled Lithium Lead) Test Blanket Modules (TBMs), developed in EU to be tested in ITER, adopt helium at 80 bar as primary coolant. This paper contains a conceptual design of the TBMs Coolant Purification System (CPS) based on the need to remove permeated tritium and gas impurities. The following steps have been considered: identification of CPS design requirements; review of the purification systems developed for Helium Cooled Fission Reactors and proposed for DEMO Fusion Reactor; selection of the most promising technologies for CPS; indications on instrumentation and procedures for tritium balance. The proposed solution is a three-stage process constituted by an oxidiser to convert Q 2 and CO to Q 2O and CO 2, an adsorption step, performed on molecular sieve at room temperature to remove Q 2O and CO 2, and a final step performed on a heated getter to remove residual impurities.

Neutronic analysis of the HCPB TBM in ITER utilizing an advanced integral approach

Nuclear design analyses were performed for the new vertical design of the HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM) in ITER utilizing an advanced integral approach. The related neutronics calculations were performed with the Monte Carlo code MCNP5 using the latest MCNP geometry model of ITER, named “A-lite”. The TBM model was generated with the geometry conversion tool McCad on the basis of a dedicated CAD model and integrated into one half of the equatorial test blanket port of the A-lite model. The neutronic calculations provided assessments of the total tritium and heat production in the TBM and its components as well as evaluations for the detailed spatial distributions of the neutron flux and the power density. Fine mesh tally calculations were utilized to provide high fidelity results for the distributions that enable identification of crucial issues.

On the theoretical–numerical study of the HEXCALIBER mock-up thermo-mechanical behaviour

Within the framework of the R&D activities promoted by European Fusion Development Agreement on the Helium-Cooled Pebble Bed Test Blanket Module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo (DIN) performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, the DIN developed a thermo-mechanical constitutive model for these pebble beds to be validated against the HEXCALIBER mock-up test campaign, carried out at the ENEA HE-FUS3 facility. The paper presents the main results of the mock-up experimental tests and of their numerical simulations performed adopting the finite element method, which allowed the DIN constitutive model to be assessed and validated.

Experimental tests and thermo-mechanical analyses on the HEXCALIBER mock-up

Within the framework of the R&D activities promoted by European Fusion Development Agreement on the helium-cooled pebble bed test blanket module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, which are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, at the DIN a thermo mechanical constitutive model was developed for both lithiated ceramics and beryllium pebble beds and it was successfully implemented on a commercial finite element code to analyze the experimental results of the ENEA test campaigns on TAZZA, HELICHETTA and HELICA mock-ups. The paper presents the preliminary theoretical–numerical study performed at the DIN to support the experimental investigations on the HEXCALIBER mock-up, a more complex test section set-up by ENEA to investigate the thermo-mechanical behaviour of mutually interacting pairs of beryllium and lithium orthosilicate pebbles beds. In particular, after the descriptions of the HEXCALIBER mock-up and its finite element model, the numerical predictions of temperatures and/or contact pressures in the pebble beds, in the structural materials and in the helium coolant are presented and critically discussed.

Monte Carlo based sensitivity and uncertainty analysis of the HCPB Test Blanket Module in ITER

This paper presents the first real test application of the Monte Carlo based sensitivity and uncertainty computational approach to the Test Blanket Module (TBM) in ITER. To this end the standard three-dimensional MCNP model of ITER (20° torus sector) with integrated TBM of the Helium-cooled Pebble Bed (HCPB) type was employed. Sensitivity profiles and integrated sensitivities were calculated for the total tritium production in the HCPB TBM by using the track length estimator approach of MCSEN, a local extension to the Monte Carlo code MCNP. The associated uncertainties of the tritium production due to nuclear data uncertainties were assessed by making use of available co-variance data from different sources. The major contribution to the sensitivities was shown to come from the materials contained in the TBM, in particular the beryllium neutron multiplier and the breeder ceramics constituents 6Li and 16O. With the calculated sensitivity profiles and the available co-variance data, the cross-section induced uncertainties of the tritium production have been assessed at a level of 2.8%.

Analysis of three LOCAs for the European HCPB test blanket system

Three LOCA sequences have been analysed in the frame of the safety assessment of the European helium cooled pebble bed (HCPB) test blanket module (TBM) system. These three sequences have been selected as representative of accidents judged to cover all scenarios envisaged in the design basis involving the TBMs. The three groups of accidents investigated with their parametric variations have as initiating event a loss of coolant accident at different location: in-vessel, in-box, and ex-vessel. In general the reduced inventory of activation products and Tritium associated with the TBM System makes the impact of this test system almost negligible to the overall safety risk of ITER. Nevertheless, the possibility to jeopardise the ITER safety concept has been analysed especially in connection to particular sequences that imply the pressurisation of the vacuum vessel due to the He coolant blow-down, the hydrogen and/or heat production from Be-steam or Be-air reaction and the necessity to assure the heat removal in the short and long period.

A constitutive model for the thermo-mechanical behaviour of fusion-relevant pebble beds and its application to the simulation of HELICA mock-up experimental results

Within the framework of the R&D activities promoted by EFDA on the Helium-Cooled Pebble Bed Test Blanket Module to be irradiated in ITER, attention has been focused on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramic pebble beds that are envisaged to be used respectively as neutron multiplier and tritium breeder. This behaviour depends, mainly, on the reactor-relevant conditions, the pebble sizes and the breeder cell geometries and a general constitutive model has not yet been validated, especially for fusion-relevant applications. ENEA-Brasimone and the Department of Nuclear Engineering (DIN) of the University of Palermo have performed intense research activities on this topic by means of an integrated experimental and theoretical approach. In particular, HELICA mock-up has been set-up and tested at the HE-FUS3 facility of ENEA-Brasimone to investigate the behaviour of lithiated ceramic pebble beds in reactor-relevant geometries, providing useful data sets on which DIN has validated its constitutive model. The paper presents the main features of the DIN constitutive model and results of two test campaigns on HELICA-mock-up. The comparison between experimental results and numerical predictions is reported too, showing a good agreement that encourages the adoption of the DIN constitutive model.

Nuclear performance analyses for HCPB test blanket modules in ITER

Neutronic, shielding and activation analyses have been performed for recent design variants of the Helium Cooled Pebble Bed (HCPB) test blanket module (TBM) in ITER on the basis of 3D Monte Carlo calculations. The main objective has been to assess and optimise the nuclear performance of the HCPB test blanket modules in terms of the tritium generation, the nuclear heating and the radiation shielding and provide, among others, the data required for the engineering design of the test modules. The shielding efficiency of the TBM system was shown to be sufficient to allow access of work personnel to the port extension after a waiting time of 10 days after shut down as required by ITER. The activation analyses provided the afterheat and activation data for quality assured safety analyses assuming a representative irradiation scenario.

Nuclear analyses for two “look-alike” helium-cooled pebble bed test blanket sub-modules proposed by the US for testing in ITER

The US is proposing two “look-alike” sub-modules, based on helium-cooled pebble bed (HCPB) ceramic breeder, to be tested in the same test blanket module (TBM) that will occupy a quarter of a port in ITER and placed next to the Japanese TBM. The TBM has a toroidal width of 73 cm, a radial depth of 60 cm and a poloidal height of 91 cm. The ceramic breeder is made of Li 4SiO 4 with 75% Li-6 enrichment (60% packing factor) and beryllium is used as the multiplier. The two sub-modules are arranged in two configurations, namely a layered configuration and an edge-on configuration. In the present work, we analyze these two sub-modules using two-dimensional discrete ordinates transport codes in R-θ model that accounts for the presence of the ITER shielding blanket and the surrounding frame of the port. The objectives are: (1) to examine the profiles of heating and tritium production rates in the two sub-modules, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings, (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanics analysis, and (3) to compare the tritium production capabilities of two variants for the HCPB blanket concept, mainly the parallel and the edge-on configurations. We present the main findings from this study in this paper.

The European breeding blankets development and the test strategy in ITER

In Europe, two breeding blankets are currently developed for DEMO reactor specifications, corresponding to two different research lines, one using a liquid breeder, the Helium-Cooled Lithium-Lead (HCLL) blanket, and one using lithiated solid breeder, the Helium-Cooled Pebble-Bed (HCPB) blanket. Referring to the so called “fast-track” EU strategy, those concepts are intended to be tested in ITER, getting the maximum of information required for launching the DEMO blanket design and construction after the first 10 years of ITER operation. For that, the EU has adopted a blanket testing strategy based on the development of Test Blanket Modules (TBM) that are expected to use DEMO relevant technologies and are designed for each ITER plasma phase to optimize the added value of results and to avoid any impact on ITER availability. Prior to their insertion in ITER, EU is running a significant R&D program for the development and qualification of HCLL and HCPB Test Blanket Modules, including at a later stage functional tests of integrated TBM mock-ups.

Design description and performance analyses of the European HCPB test blanket system in ITER feat

The helium cooled pebble bed (HCPB) blanket is one of the two European DEMO blanket concepts proposed for testing in international thermonuclear experimental reactor (ITER). The purpose of the tests is to validate the design principles and the operational feasibility for the demonstration blanket system. This includes the basic support functions like tritium extraction, helium cooling and heat transport, and helium purification. In addition, the basic properties and operating characteristics of the system's materials will be validated. Safety, reliability, maintenance and dismantling will be equally addressed. To assess these qualities, the ITER horizontal ports will be used to provide a relevant fusion plasma and the appropriate nuclear environment. At conclusion of the ITER feat EDA phase (July 2001), a revised design of the HCPB test blanket system have been completed to adapt the previous design (for ITER FDR) to the new operational conditions of ITER and to a new strategy for the blanket testing in this reactor. Design description, performance and safety analyses are presented in this paper.