High-temperature Shielding Research Articles

Electromagnetic interference shielding performance of alternatively-deposited multilayer SiC/PyC porous ceramics

Herein, alternatively-deposited multilayer SiC/pyrolytic carbon (PyC) porous ceramics are fabricated by low-pressure chemical vapor infiltration (LPCVI) method, using nickel (Ni) foam as template. The influence of number of deposition layers (n) on morphology, physical properties, electrical conductivity, and electromagnetic properties of the porous ceramics was systematically investigated. The results revealed that the density of as-prepared SiC/PyC porous ceramics increased from 0.376 g/cm3 to 0.863 g/cm3, whereas the porosity decreased from 86.97% to 70.24% with increasing number of alternately-deposited SiC and PyC layers. Moreover, electrical conductivity and shielding effectiveness gradually increased with increasing number of deposition layers. At n = 6, the electrical conductivity of SiC/PyC porous ceramics increased to 0.289 S/cm, meanwhile room-temperature average total shielding and absorption effectiveness values increased to 26.7 dB and 16 dB, respectively, corresponding to an increase of 44.32% and 58.42% as compared to the sample of n = 0 (AS). In addition, total specific shielding effectiveness of AS and PS2-PS6 samples ranged from 30 to 50 dB cm3/g, and gradually decreased with increasing number of deposition layers. Lastly, high-temperature shielding effectiveness has been assessed in the temperature range of 25 to 600 °C. Excellent stability has been rendered by alternatively-deposited multilayer SiC/PyC porous ceramics.

Effect of the pyrolytic carbon (PyC) content on the dielectric and electromagnetic interference shielding properties of layered SiC/PyC porous ceramics

A novel layered SiC/pyrolytic carbon (PyC) porous ceramic was synthesized from a nickel foam substrate via low-pressure chemical vapor infiltration (LPCVI) with SiCl3CH3-NH3-BCl3-H2-Ar. The microstructure and phase composition of the PyC deposited via Ni catalysis were investigated. In addition, the effect of the PyC content on the microstructure, conductivity, and electromagnetic shielding effectiveness of a two-layered SiC/PyC porous ceramic were discussed. Both the electrical conductivity (from 0.090 to 0.319 S/cm) and the total shielding effectiveness (from 19.2 to 29.0 dB) of the two-layer SiC/PyC porous ceramic (pore size: 200–400 µm) increased with the PyC content. The high-temperature shielding effectiveness of the sample showed an outstanding stability with temperature and remained nearly unchanged (only 2 dB variation) over the 25–600 °C temperature range.

Boron carbide composites with highly aligned graphene nanoplatelets: light-weight and efficient electromagnetic interference shielding materials at high temperatures.

B4C-based ceramic composites containing 0–2 vol% highly aligned graphene nanoplatelets (GNPs) are fabricated. The electromagnetic interference (EMI) shielding properties of the obtained composites are investigated at X-band (8.2–12.4 GHz) frequency range from room-temperature up to 800 °C. All composites exhibit outstanding EMI shielding properties with satisfactory frequency- and thermal-stability. The shielding effectiveness (SE) of GNP/B4C composites increases monotonically with increasing GNP loading. Superior room-temperature SE close to 40 dB is achieved with only 2 vol% GNPs and high SE around 35 dB still persists at 800 °C. Considering their relatively low density, GNP/B4C composites possess a high specific shielding effectiveness (SSE) of 16 dB cm3 g−1 which is among the highest values in reported ceramic-based shielding composites. Especially, the GNP/B4C composite with 2 vol% GNPs exhibits the highest SSE/t (SSE divided by thickness) values at temperatures above 200 °C for all reported shielding composites, indicating that GNP/B4C composites belong to the most promising high-temperature shielding composites. The excellent shielding properties of GNP/B4C composites arise mainly from the high electrical conductivity, high dielectric loss and the multiple reflections by the highly aligned and large-sized GNP layers.

Activation and Environmental Aspects of In-Vacuum Vessel Components of CFETR**supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015BG108002, 2014GB122000, 2014GB119000), National Natural Science Foundation of China (No. 11175207)

The water-cooled ceramic breeder (WCCB) blanket is one of the three candidates of China's Fusion Engineering Test Reactor (CFETR). The evaluation of the radioactivity and decay heat produced by neutrons for the in-vacuum vessel components is essential for the assessment of radioactive wastes and the safety of CFETR. The activation calculation of CFETR in-vacuum vessel components was carried out by using the Monte Carlo N-Particle Transport Code MCNP, IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, and the nuclear inventory code FISPACT-2007 and corresponding EAF-2007 libraries. In these analyses, the three-dimensional (3-D) neutronics model was employed and the WCCB blanket, the divertor, and the shield were modeled in detail to provide the detailed spatial distribution of the neutron flux and energy spectra. Then the neutron flux, energy spectra and the materials specification were transferred to FISPACT for the activation calculation with an assumed irradiation scenario of CFETR. This paper presents the main results of the activation analysis to evaluate the radioactivity, the decay heat, the contact dose, and the waste classification of the radioactive materials. At the time of shutdown, the activity of the WCCB blanket is 1.88×1019 Bq and the specific activity, the decay heat and the contact dose rate are 1.7 × 1013 Bq/kg, 3.05 MW, and 2.0 × 103 Sv/h respectively. After cooling for 100 years, 79% (4166.4 tons) radioactive wastes produced from the blanket, divertor, high temperature shield (HTS) and low temperature shield (LTS) need near surface disposal, while 21% (1112.3 tons) need geological disposal. According to results of the contact dose rate, all the components of the blanket, divertor, HTS and LTS could potentially be recycled after shutdown by using advanced remote handling equipment. In addition, the selection of Eurofer97 or RAFM for the divertor is better than that of SS316 because SS316 makes the activity of the divertor-body keep at a relatively high level.

ARIES-ACT1 System Configuration, Assembly, and Maintenance

ARIES-ACT1 engineering design efforts were devoted to developing a credible configuration that allows for rapid removal of full-power core sectors followed by disassembly in hot cells during maintenance. The power core evolved with the main objective of achieving high performance while maintaining attractive design features and credible configuration, maintenance, and fabrication processes. To achieve high availability and maintainability of a fusion power plant, the power core components of a sector, including inboard and outboard first wall (FW)/blankets, upper and lower divertors, and structural ring or high-temperature shield, were integrated into one replacement unit to minimize time-consuming handling inside the plasma chamber. As with the ARIES-AT design, the FW/blanket design was based on Pb-17Li as coolant and breeder, and low-activation SiC/SiC as structural material; however, the Pb-17Li mass flow rate control, flow path, FW and blanket cooling channels, coolant access pipes, and blanket structural configuration have been revised and improved to provide about the same thermal performance (∼58% thermal efficiency) while keeping the magnetohydrodynamic pressure drop and pumping power, material temperature, and stresses at an acceptable level. Helium-cooled W or W-alloy divertor concepts were developed to accommodate a peak surface heat flux up to ∼14 MW/m2. They include a smaller finger-based divertor and a midsized T-tube and larger plate-type divertor concepts, which take advantage of a simple configuration, and the smaller number of plate units and joints in a power plant. The two-zone divertor concept, with the combination of a finger-based divertor and plate-type divertor, was selected and integrated into the ARIES-ACT1 power core. The fingers are used to accommodate the designed peak heat flux of ∼13 MW/m2, while the plate-type divertor is used for the lower heat flux region. The overall power core configuration and system integration, as well as the definitions of major power core components, such as the FW/blankets, divertor, structural ring, and the vacuum vessel, are described here and the main design features are highlighted. Sector maintenance operations have been investigated and motion demonstrations for removing the power core sectors have been performed using state-of-the-art three-dimensional CAD to analyze the clearances and spaces in all directions. The maintenance sequence and procedure for removing the replacement unit from the plasma chamber to the hot cell for exchange and refurbishment are also discussed in this paper.

Overall Power Core Configuration and System Integration for ARIES-ACT1 Fusion Power Plant

ARIES-ACT1 power plant has been designed and configured to allow for rapid removal of full power core sectors followed by disassembly in hot cells during maintenance operation. To achieve high availability and maintainablity of a fusion power plant, power core components of a sector, including inboard and outboard FW/blankets, upper and lower divertor, structural ring or high temperature shield were integrated into one replacement unit to minimize time comsuming handling inside plasma chamber. In this paper, the overall power core configuration and system integration, as well as the definitions of major power core components are described and main design features are highlighted.

Heatline analysis on thermal management with conjugate natural convection in a square cavity

Conjugate natural convection finds various thermal applications in chemical industries, where the heat transfer is controlled by the presence of solid walls such as heat exchanger, nuclear reactors, fin type cooling and solar storage systems. Heat flow distribution within square cavity enclosed by vertical conducting walls of definite thickness (t1 and t2) is analyzed for various fluids (Pr) in this study based on the location of the wall thickness [left wall (case 1)/ right wall (case 2)/ both side walls (case 3)] and conductivity ratios between solid and fluid regions (K). At Ra=105, circular heatlines are observed near core of the cavity at K=0.1 whereas they are distorted and pushed towards the hot wall at K=10 with low Pr (Pr=0.015). On the other hand, heatlines are horizontally stretched at core of the cavity for higher Pr (Pr=0.7 and 1000) at K=10. End to end heatlines are highly compressed near top portion of the cavity at K=10 irrespective of Pr. Closed loop heatlines are absent for case 2 at Ra=103 whereas closed loop heatlines with lesser magnitude than cases 1 and 3 are observed for case 2 at Ra=105 due to less heat transfer from the hot solid wall irrespective of Pr at K=0.1 and 1. The heat transfer rate can be maintained constant at low thermal conductivity ratio (K) even for the high convective regime (Ra=105) irrespective of wall thickness (t1 and t2). Average Nusselt number shows overall larger heat transfer rate for higher K (K=10), which is almost identical with classical natural convection (zero wall thickness) compared to lower K (K=0.1 and 1) irrespective of location of wall thickness (cases 1–3). In order to achieve almost invariant or lower fluid temperature at Ra=105 for t1+t2=0.2 and 0.8, solid wall at hot side (case 2) may be useful. This heating strategy may be viewed for high temperature shielding or minimization of thermal runaway for temperature sensitive applications, such as environmental control system, chemical storage reservoirs, etc., where heat flow is controlled by solid wall resistance.

Neutronic Assessment of Candidate Materials for TF Coils Shielding in a DEMO Fusion Reactor Based on a DCLL Blanket

Under the Spanish Breeding Blanket Technology Program TECNO_FUS, a conceptual design of a dual-coolant lithium-lead (DCLL) blanket for DEMO is being revisited. In this work, different shielding candidate materials are assessed in their ability to satisfy the radiation load requirements that must be fulfilled in the toroidal field (TF) coils: absorbed dose in the insulator (Epoxy), peak fast neutron fluence in the superconductor (Nb3Sn), peak nuclear heating in the winding pack and maximum neutron fluence in the cooper stabilizer. Furthermore, the impact of the material choice on waste management requirements of both shielding and vacuum vessel (VV) materials is evaluated, and the performance of candidate materials is examined in terms of the helium production in the VV SS316LN material and its implications in reweldability. Materials discussed for the High Temperature Shield are Eurofer, graphite, B4C, WC and WB4C, while the metal hydrides ZrH2, Zr(BH4)4, and TiH2 are discussed for the Low Temperature Shield. In the case of DEMO irradiation scenario, all the analyzed material combinations fulfill the design requirements for the waste management of the shield and VV, He production in the VV wall and TF coils radiation loads requirements.

Conceptual study of vertical sector transport maintenance for DEMO fusion reactor

The maintenance scheme is a critical issue for DEMO design, and requires high availability of the reactor. The SlimCS, designed in JAEA, adopts the horizontal sector transport hot cell maintenance scheme. In order to determine the most appropriate DEMO reactor maintenance scheme, it is important to assess the various maintenance schemes. In this paper the vertical sector transport maintenance concept is proposed for the first time. In the sector maintenance scheme, the amount of cutting/re-welding of the piping is minimized. The sector including blanket modules and a high-temperature shield was divided into 10° segments in a toroidal direction. The sectors are designed to be removed and re-inserted through upper alternate vertical maintenance ports. In the case of the vertical sector transport maintenance scheme, inter-coil structures could be adopted for use against turnover force in toroidal field (TF) coils. This shows an advantage in DEMO maintenance compared with horizontal sector transport.

Conceptual design of the blanket mechanical attachment for the helium-cooled lithium–lead reactor

The conceptual design of a new type of fusion reactor based on the helium-cooled lithium–lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 °C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 °C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions.

Transmutation and activation of reduced activation ferritic martensitic steel in molten salt cooled fusion power plants

Neutron activation analysis was conducted for the reduced activation ferritic/martensitic (RAFM) steel used in flibe molten-salt cooled fusion blankets. After 22.4 MW yr/m 2 of neutron exposure, the RAFM steel first wall in a molten salt blanket with 40% lithium-6 enrichment in lithium was found to be within 1 mSv/h in contact dose rate after 100 yr of cooling. The contact dose rate drops to 30 and 20 μSv/h or less, respectively, when the cooling times are 300 and 500 yr after discharge. The RAFM steel discharged from the high-temperature shield component would be allowed for hands-on recycling after 100 yr of cooling, when the contact dose rate is 10 μSv/h or less. The most significant changes found in the RAFM steel first wall due to nuclear transmutation, are 10% decrease in W and 10% increase in Ti. Additionally, there are minor elements produced: Mn – <1.2%, V – <0.26%, Re – <0.2%, Ta – <0.08%, and Os – <0.1%, all in weight percent. The gaseous elements generated are H and He, and the, respectively, accumulated quantities are about 260 and 190 wppm.

Lay-out of the He-cooled solid breeder model B in the European power plant conceptual study

The European helium cooled pebble bed (HCPB) blanket concept is the basis for one of two limited-extrapolation plant models that are being elaborated within the European power plant conceptual study (PPCS). In addition to addressing the case for fusion safety and environmental compatibility, following earlier studies like SEAFP or SEAL, this reactor study puts emphasis on plant availability and economic viability, which are closely related to specific plant models and require a detailed lay-out of the fusion power core and a consideration of the overall plant (balance of plant). Within the development of in-vessel components for the plant model, the major tasks to be carried out were: (i) adaptation of the HCPB concept—featuring separate pebble beds of ceramic breeder and Beryllium neutron multiplier and reduced-activation ferritic-martensitic steel EUROFER as structural material—to the large module segmentation chosen for reasons of plant availability in part II of the PPCS; (ii) proposal of a concept for a Helium cooled divertor compatible with a maximum of 10 MW/m 2 heat flux to satisfy the requirements of reasonably extrapolated plasma physics; (iii) lay-out of the major plant model components and integration into the in-vessel dimensions found from system code calculations for a power plant of 1500 MW electrical output and iterated data on the plant model performance. The paper defines all major in-vessel components of plant model B, as it is called in the PPCS, namely (i) the unit of FW, blanket and high temperature shield that is to be replaced regularly; (ii) the low temperature shield that is laid out as a lifetime component of the reactor; (iii) the divertor; and (iv) the in-vessel manifolding. Results are presented for the thermal-hydraulic performance of the components and for the thermal-mechanical behaviour of the blanket and the divertor target plate. These results suggest, together with results from the wider exploration of the plant model within the PPCS, that the He cooled solid breeder blanket is a credible concept. They stress the positive role that He cooling can play in economically attractive fusion power plants.

Use of concretes for high-temperature shielding of nuclear reactors

The authors have studied the possibility of using chromite and chomotte heat-resistant concretes for the thermal shields of reactors. They observe neutron fluxes of various intensities (up to 1013 neutrons/cm2·sec, with spectrum similar to fission spectrum), absorbed by shields of these materials. They compute the transmission of neutrons and of fluxes of gamma quanta and the heat emission in the shielding. They calculate the temperatures in the shielding for various neutron fluxes, concrete thicknesses and cooling conditions. They perform a statistical calculation of the temperature stresses for shielding constructed of heat-resistant ferroconcrete.

Realistic Evaluation of Shadow Shields for Space Nuclear Systems

The Outside Test Tank is the only nuclear shielding test facility in the United States suitable for investigating the nuiclear characteristics of shadow shield geometries such as will be required to protect components and man near nuclear systems operating in outer space. The OTT is designed so that only the radiation of interest, that is, neutrons and gamma rays penetrating or originating in the shadow shield materials, is measured. No radiation is measured that is scattered by air or ground. The only radiation that reaches the detectors penetrates shield components on the direct line of sight between the test reactor, the shield, and the test radiation detectors. The design and operation of the Outside Test Tank are described. Problems encountered during testing are discussed, as are the instruments and devices used to measure neutron and gamma ray dose rates and fluxes. A number of unusual phenomena that were discovered are examined. Tested configurations included such high temperature shielding materials as W and LiH. In summary, the information presented in this paper proves that realistic evaluation of actual nuclear shadow shield configurations can be made conveniently and relatively inexpensively without having to test in the vacuum of outer space. The data measured in the Outside Test Tank clearly demonstrate that the design of efficient shadow shields for space nuclear systems is not nearly as straightforward as has been suggested.