ITER Shielding Blanket Research Articles

Comparing the Prediction of “Attila” Code to the Experimental Data of Fusion Integral Experiments and to the Results of MCNP Code

The calculation accuracy of the newly developed 3D discrete ordinates code, Attila, is benchmarked by comparing its prediction to the measured data in two mockups bombarded by 14 MeV neutron source of the FNG facility located at Frascati, Italy. The results are also compared to those based on MCNP Monte Carlo code for the same measured reactions. The experimental mock-ups simulate parts of ITER in-vessel components, namely, the tungsten (W) mockup and the ITER shielding blanket. The first mockup was used to validate W data as a material for plasma facing component. A streaming path was introduced in the second configuration. The objective of this paper is to benchmark Attila code to determine its adequacy for fusion application. Another objective is to compare results based on two distinctive 3D calculation tools using the same nuclear data, FENDL2.1, and the same response function (IRDF-90) for measured data. The results of these comparisons are reported in this paper.

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.

Manufacture of blanket shield modules for ITER

A research and development programme for the ITER blanket shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups (small scale and medium scale) and full-scale prototypes of shield blocks (SB) and first wall (FW) panels. The manufacturing feasibility of FW panels has been demonstrated for two copper alloy candidates. Two designs have been developed for the manufacture of the SB, one for a conventional fabrication route and one for a fabrication route based on the hot isostatic press technology. This paper presents the fabrication routes developed in Europe for the manufacture of the ITER Shield modules.

On the Strategy and Requirements for Neutronics Testing in ITER

Neutronics testing is among the several types of fusion technology testing scheduled to be performed in ITER. The three ports assigned for testing will test several blanket concepts proposed by the various parties with test blanket modules (TBM) that utilize different breeders and coolants. Nevertheless, neutronics issues to be resolved in ITER-TBM are generic in nature and are important to each TBM type. Dedicated neutronics tests specifically address the accuracy involved in predicting key neutronics parameters such as tritium production rate, TPR, volumetric heating rate, induced activation and decay heat, and radiation damage to the reactor components. In this paper, we address some strategies for performing the neutronics tests. Tritium self-sufficiency cannot be confirmed by testing in ITER, however, the testing can provide valuable information regarding the main parameters needed to assess the feasibility of achieving tritium self-sufficiency. The paper also addresses the operational requirement (i.e. flux and fluence) as well as the geometrical requirement of the test module (i.e. minimum size) in order to have meaningful and useful tests. Measured neutronics data require high spatial resolution. This necessitates that the measured quantity be as flat as possible in the innermost locations inside the test module. This requirement has been confirmed in the present work based on results from two-dimensional calculations. The US and Japan solid breeder test blanket modules are placed inside half a port in ITER. The R-θ model used accounts for the presence of the ITER shielding blanket and the surrounding frame of the port.

The neutron irradiation effect on mechanical properties of HIP joint material

Dispersion strengthened copper (DSCu) and stainless steel are the candidate material for the heat sink and the structural material of the ITER shielding blanket and these materials are joined by hot isostatic pressing (HIP). In this study, the neutron irradiation effect on mechanical properties of HIP joint material was examined by tensile and impact tests using specimens with irradiation damage of about 1.5 dpa. The results of tensile tests show that tensile strength of HIP joint material was about the same as that of DSCu base material, and this trend did not change after neutron irradiation. On the other hand, the impact value of HIP joint material was smaller than that of DSCu base material because of the diffusion of main elements at joint boundary. It was shown that embrittlement by the neutron irradiation effect is smaller than that of the effect by HIP joint.

Estimation of Helium Production due to Neutron Streaming and Establishment of Shielding Design Conditions in Fusion Shielding Blanket by 3-D Monte Carlo Calculation

In the shielding design of the ITER shielding blanket, helium productions in the branch pipes and the back plate have been a critical concern from the viewpoint of rewelding of stainless-steel. We have performed three dimensional neutron streaming analyses for the ITER shielding blanket by a Monte Carlo method, and we have estimated helium productions in stainless steel. Helium productions at the branch pipes filled in water drastically increase due to a reaction between thermal neutron and boron, which is a trace element of stainless steel. We have calculated helium productions for 310–450 mm thick blanket modules with 10–60 mm wide gaps between adjacent blanket modules and 0.1–20 wppm boron in stainless steel. From the calculated results, we have deduced the approximate equations for the estimation of the helium productions as a function of the blanket module thickness, the gap width and the boron content, and we have clarified shielding design conditions required to satisfy the design limit.

00/00784 On a simplified system for steam production in nuclear power plants with pressurized water reactors

The recently developed FENDL-1 database, both in multigroup form (FENDL/MG-1.0) and continuous energy form (FENDL/MC-1.0) has been tested through analyzing a fusion integral experiment performed at the FNS facility, Japan, on a large bulk shielding assembly made of multilayers of SS316 and water. The assembly is a replica that simulates ITER shielding blanket and is bombarded by a 14 MeV source placed at 30 cm from the cylindrical assembly and housed inside a SS316 cylindrical can. This activity is undertaken as part of co-operation with JAERI on executing the required neutronics R&D tasks for ITER shield design. The objectives are (a) benchmarking FENDL-1 data and identifying any flaws that may exist in this newly developed database, and (b) examining the range of discrepancy between the calculated nuclear parameters inside the assembly and the measured ones in terms of the ratio of the calculated-to-experimental (C/E) data. Both differential and integral experimental data were analyzed along the central axis of the ∼ 120 cm D × 140 cm L assembly. The analyses with the multigroup data, MG also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo (MC) code was used with the FENDL/MC-1 data. The largest range of discrepancy between calculated and measured responses (reaction rates, neutron spectra, gamma ray heating, etc.) was found to be ∼ 20–30% even though in most cases this discrepancy falls within the experimental errors.

00/00782 Neutronics experiment on a mock-up of the ITER shielding blanket at the Frascati Neutron Generator

00/00782 Neutronics experiment on a mock-up of the ITER shielding blanket at the Frascati Neutron Generator

Neutronics experiment on a mock-up of the ITER shielding blanket at the Frascati Neutron Generator

An integral bulk shield experiment has been performed at the Frascati Neutron Generator (FNG) with the main objective to validate the ITER shielding system. To this end, a suitable mock-up of the ITER inboard shielding system including first wall, shielding blanket, vacuum vessel and toroidal field coil, had been assembled at ENEA Frascati. Neutron and γ-ray spectra as well as various nuclear responses were measured inside this mock-up when irradiated by 14-MeV neutrons. Measured and calculated data were compared to validate transport codes and nuclear data used in the design of the ITER system. In particular, the measured neutron and γ-spectra, reaction rates and the nuclear heating were compared with the same quantities calculated with the MCNP Monte Carlo code using the Fusion Evaluated Nuclear Data Library (FENDL)-1.0 and the European Fusion File (EFF)-3.0 neutron cross section data. The neutron-induced activation measured in stainless steel was compared with calculations performed with the FISPACT code and using the most recent activation cross section data file FENDL/A-2.0. The paper includes a comparison of calculated over measured quantities ( C/ E) and a discussion of their relevance to the ITER nuclear design. It is found that, generally, the measured nuclear quantities are predicted by FENDL-1.0 and EFF-3.0 calculations within an uncertainty margin of about ±15% in the shielding blanket and vacuum vessel, and ±30% up to the region of the toroidal field coil.

Decay Heat Analyses for ITER Blanket Designs

Decay heat is a potentially important energy source in the ITER safety analysis. Decay heat analyses were performed for the ITER shielding and breeding blanket designs using the proposed operational scenarios for the base performance and extended performance phases (BPP and EPP). Decay heat due to operational scenarios deviated from the BPP were also investigated.

Fabrication of an ITER Shielding Blanket Prototype

A full-scale prototype mockup of the ITER shielding blanket composed of the first wall with built-in circular tubes and the shield block has been successfully fabricated. Joints of the first wall/the shield block as well as those of the DSCu/DSCu, DSCu/SS and SS/SS in the first wall, were simultaneously HIP bonded. The fabricated prototype mockup fully satisfied the specified design conditions such as dimensional accuracy, pressure-resistance, etc. After HIPping, HIPped interfaces were metallographically observed with pieces cut from the edge of the mockup. There were no harmful voids at the bonded interfaces, and good bondability of simultaneous HIP method was confirmed. In addition, engineering data on fabrication including bonding deformation were obtained.

Design of the ITER Shielding Blanket

The ITER blanket system removes the surface heat flux from the plasma and from bulk heating by the neutrons, reduces the activity in the vacuum vessel (W) structural material to the level allowable to ensure vessel reweldability for the ITER fluence goal and, in combination with the vacuum vessel, protects the superconducting coils and other ex-vessel components from excessive nuclear heating and radiation damage. The blanket system contributes with its eddy currents to the passive stabilization of the plasma motion. It minimizes the effects of electromagnetic loads on the VV due to plasma disruptions, and provides a well defined load path to the VV for net vertical and horizontal loads arising from vertical displacement events (VDE's). The system is designed to allow the possibility of replacing the shield with a breeding blanket, within the same dimensional, maintenance, and coolant constraints, to provide the tritium to meet the technical objectives of the Enhanced Performance Phase.The basic blanket system concept as well as the arrangement and function of its components is essentially unchanged from that established in 19951. However, as discussed in this paper, the design of each component has progressed significantly as a result of the detail design and technical analysis efforts of the last two years. The main components of the blanket system are:• A back plate: a structure comprising a double wall shell that supports the first wall/shield modules and routes the coolant water to them.• First wall/shield modules: comprising a plasma facing first wall (FW) section, and a shielding (or later breeding) section. Primary wall and baffle modules are distinguishable by the function of their FW.• Limiters: define the plasma boundary during plasma start-up and shutdown and are located in equatorial ports.• Flexible connectors, electrical straps, and branch pipes: the remote handling compatible structural, electrical, and cooling connections between the modules and back plate.• Filler shields: shielding permanently mounted to the back plate in the triangular gaps between FW/shield modules.The system will use austenitic stainless steel 316L(N)-IG (ITER Grade) as the primary structural material cooled by water with inlet conditions of 3.8 MPa and 140°C. The plasma facing surface of the FW will be beryllium except the lower region of the baffles, where tungsten is used. The electrical straps and heat sink layer in the FW will be copper alloy. A titanium alloy is the prime candidate material for the flexible connectors.

Optimization of HIP bonding conditions for ITER shielding blanket/first wall made from austenitic stainless steel and dispersion strengthened copper alloy

Optimum bonding conditions were studied on the Hot Isostatic Pressing (HIP) bonded joints of type 316L austenitic stainless steel and Dispersion Strengthened Copper alloy (DSCu) for application to the ITER shielding blanket / first wall. HIP bonded joints were fabricated at temperatures in a 980–1050°C range, and a series of mechanical tests and metallurgical observations were conducted on the joints. Also, bondability of two grades of DSCu (Glidcop Al-25 ® and Al-15 ®) with SS316L was examined in terms of mechanical properties of the HIP bonded joints. From those studies it was concluded that the HIP temperature of 1050°C was an optimal condition for obtaining higher ductility, impact values and fatigue strength. Also, SS316L/Al-15 joints showed better results in terms of ductility and impact values compared with SS316L/Al-25 joints.

Development of first wall/blanket structure by hot isostatic pressing (HIP) in the JAERI

In the ITER shielding blanket design, first wall coolant tubes of type 316LN stainless steel (SS) are embedded in a heat sink of Al 2O 3 dispersion-strengthened Cu (DSCu), which are joined to a SS shield block. Development of fabrication technology using hot isostatic pressing (HIP) for this structure, including testing, is currently being performed at JAERI. Recent progress in this development is reviewed in this paper. A simultaneous joining of DSCu/DSCu, DSCu/SS and SS/SS, especially at the first wall and the first wall to the shield block, by a solid HIP technique, has been pursued. Following the optimization of HIP conditions, mechanical tests of the HIP joints and fabrication of first wall mock-ups were performed. High heat flux tests of the first wall mock-ups resulted in their sufficient heat removal capability and integrity of the HIP joints. Small- and medium-scale mock-ups of the shielding blanket, integrated with the first wall, were fabricated to examine the applicability of the HIP technique and develop a fabrication route. As a result of these activities, fabrication technologies and the route to be applied for the ITER shielding blanket and sufficient thermo-mechanical performance of the HIP bonded structure were foreseen.

European helium cooled pebble bed blanket: Design of blanket: module to be tested in ITER

The helium-cooled pebble bed (HCPB) blanket for a fusion DEMO reactor is based on the use of separate small lithium orthosilicate and beryllium pebble beds placed between radial toroidal cooling plates. The cooling is provided by helium at 8 MPa. The tritium produced in the pebble beds is purged by the flow of helium at 0.1 MPa. The structural material is martensitic steel. It is foreseen, after extensive R&D, to test in ITER a blanket module based on the HCPB design which is one of the two European proposals for the ITER Test Blanket Programme. To facilitate the handling operations, the blanket test module (BTM) is bolted to a surrounding water cooled frame fixed to the ITER shield blanket back plate. For the design of the test module, 3D Monte Carlo neutronic calculations and thermo-hydraulic and stress analyses for the operation during the basic performance phase (BPP) and during the extended performance phase (EPP) of ITER have been performed. The behaviour of the test modules during LOCA, LOFA and electromagnetic transients has been investigated. Conceptual designs of the required ancillary loops have been performed.

Verification of ITER shielding capability and FENDL data benchmarking through analysis of bulk shielding experiment on large SS316/water assembly bombarded with 14MeV neutrons

Abstract The recently developed FENDL-1 database, both in multigroup form (FENDL/MG-1.0) and continuous energy form (FENDL/MC-1.0) has been tested through analyzing a fusion integral experiment performed at the FNS facility, Japan, on a large bulk shielding assembly made of multilayers of SS316 and water. The assembly is a replica that simulates ITER shielding blanket and is bombarded by a 14 MeV source placed at 30 cm from the cylindrical assembly and housed inside a SS316 cylindrical can. This activity is undertaken as part of co-operation with JAERI on executing the required neutronics R&D tasks for ITER shield design. The objectives are (a) benchmarking FENDL-1 data and identifying any flaws that may exist in this newly developed database, and (b) examining the range of discrepancy between the calculated nuclear parameters inside the assembly and the measured ones in terms of the ratio of the calculated-to-experimental (C/E) data. Both differential and integral experimental data were analyzed along the central axis of the ∼ 120 cm D × 140 cm L assembly. The analyses with the multigroup data, MG also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo (MC) code was used with the FENDL/MC-1 data. The largest range of discrepancy between calculated and measured responses (reaction rates, neutron spectra, gamma ray heating, etc.) was found to be ∼ 20–30% even though in most cases this discrepancy falls within the experimental errors.

Corrosion resistance of austenitic steels and alloys in high temperature water

An ampoule corrosion tests technique is presented for making a comparative corrosion-proof estimation of austenitic steels and alloys which are candidate materials for the ITER first wall and shield blanket. Influence of Cl − on the rate and character of corrosion cracking has been determined. The dependence of time before destruction on Cl − contents was obtained by experiments in twice-distilled water at a Cl − content value of less than 0.05 mg/l. Kinetics of a corrosion crack growth has been determined by way of acoustic emission and measurement of electrical resistance. It has been revealed, that corrosion cracking of 316L and 316Ti steels in water with 100 mg/l of Cl − takes place at temperatures over 50°C. For 316L (Russian analogue), 316Ti, 08Cr18Ni10Ti (type 304), 00Cr16Ni15Mo3Nb (type 316) steels and CrNiMo-1 (42–47%Cr, 1%Mo, Ni-base) alloy corrosion long-term strength and crack-resistance characteristics were obtained. Results of tests show, that corrosion characteristics are near the same for all austenitic steels under investigation with CrNiMo-1 alloy being the exception. It was not destroyed during ampoule tests.

Thermal-mechanical analyses of the ITER shielding blanket designs

Abstract Thermal-mechanical analyses were performed for the ITER shielding blanket design to define the performance of the structures when subjected to thermal loads, the internal pressure and the electromagnetic forces. The shielding blanket design under consideration has austenitic steel structure, low temperature and pressure water coolant, and a copper layer coated with beryllium for the first wall. The first wall is structurally integrated with the shielding blanket by thin lateral walls welded to the back plate. At the side wall of the sector, the internal pressure is sustained by the adjacent canisters. Two-dimensional analyses of the equatorial cross-section were carried out using generalized plane strain boundary conditions to simulate the out of plane structure behavior. The mechanical loads used in the analyses consist of an internal coolant pressure of 2 MPa and electromagnetic loads on the first wall due to plasma disruption events of 1 MPa for outboard sections. Two thermal loads were included: a surface heat flux of 50 W cm −2 on the first wall and nuclear heating from the neutronics analyses.