Solid Breeder Research Articles

Surface oxidation effect on deuterium permeation in reduced activation ferritic/martensitic steel F82H for DEMO application

Fuel loss and environmental contamination by tritium permeation through structural materials are critical issues for the establishment of a fusion DEMO reactor. In this study, the effectivity of a chromium oxide layer formed on reduced activation ferritic/martensitic steel F82H as a tritium permeation barrier and its stability under simulated solid/liquid breeder blanket conditions have been investigated. A uniform 100-nm-thick chromium oxide layer was formed by heat treatment at 710 °C for 5 min in 50% argon-50% hydrogen mixed gas with the flow rate of 200 standard cubic centimeter per minute. After exposure to simulated solid breeder blanket conditions, an iron oxide layer and a spinel-type iron-chromium oxide layer formed. In the case of a liquid breeder blanket condition, the chromium oxide layer partly lost at 500 °C for 100 h. The chromium oxide-formed sample decreased deuterium permeation flux by a factor of up to 150. The permeation reduction efficiency deteriorated after exposure to a solid breeder blanket condition due to a change of the chromium oxide layer. However, the chromium oxide formation would play a role to reduce hydrogen isotope permeation even after reduction of the oxide layers.

Reactivity and thermal stability of ternary Be-Zr-V beryllides

A water-cooled solid breeder blanket of a fusion reactor has been adopted in Japan, which focuses on safety concerns; safety is one of the most critical issues. In particular, as a multiplier, Be pebbles are well known to possibly generate hydrogen and react in an exothermic manner due to reaction with H2O during the loss of coolant accident. In contrast to these Be pebbles, beryllium intermetallic compounds (beryllides) demonstrate promise due to their more stable chemical reactivity at high temperatures. Currently, several studies have reported the development of advanced neutron multipliers in Japan, and the EU is part of the demonstration R&D activities at the International Fusion Energy Research Center project, which forms a part of the Broader Approach program. Fabrication methods for beryllide pebbles have been successfully developed by the combination of plasma sintering synthesis and a rotating electrode granulation method.By employing these methods, ternary beryllide pebbles with mixtures of Be13Zr and Be12V were first synthesized with ratios of 1.0:1.0, 1.0:0.8, 1.0:0.6, 1.0:0.4, 1.0:0.2, and 1.0:0.1. The small-sized Be13Zr phase as the precipitate and the Be12V phase were formed without the formation of the Be phase as these compositions did not exhibit the peritectic reaction. In addition, ternary beryllide pebbles exhibited a lower reactivity to water vapor as well as a higher thermal stability than Be pebble. In this study, not only the synthesis but also characterizations at high temperatures were discussed.

Fabrication of Li4SiO4-Li2ZrO3 composite pebbles using extrusion and spherodization technique with improved crush load and moisture stability

The effect of different amount of Li2ZrO3(LZ) (0-15wt%) addition on the properties of composite Li4SiO4 (LS) ceramic pebbles was studied. The Li4SiO4-Li2ZrO3 composite powder was prepared using solid state route at a calcination temperature as low as 900 °C via in-situ method. The composite pebbles were fabricated using a cost-effective and simple technique called extrusion-spherodization. The sintered pebbles were characterized for density, grain size, pore size distribution, crush load and moisture stability. The composite pebbles showed good sintered density (∼87.1%) for LS-5wt% LZ in comparison to LS pebbles (∼80%) when fired at 1000 °C. Moreover, the LS grain size in the composite pebble was significantly reduced by ∼19.7% when compared to LS pebbles. We also found that the average crush load value of the LS-5wt% LZ composite pebbles was improved by nearly 77% (30 N) to that of the pure LS pebbles (17 N). The LS-5wt% LZ pebbles showed excellent stability to moisture in terms of phase and mechanical crush load retention after 30 days of exposure to humidity. The results achieved in this work were found to be fit to the desired properties for use as a ceramic solid breeder.

Density functional theory study of the point defect energetics in γ-LiAlO2, Li2ZrO3 and Li2TiO3 materials

The defect energetics in γ-LiAlO2, Li2TiO3, and Li2ZrO3 materials used in tritium-producing burnable absorber rods and fusion solid breeder applications was investigated using density functional theory calculations. A comprehensive analysis of the charged defects was performed for cation and anion vacancies, interstitials, antisite defects, and 3H interstitial and substitutional defects to understand the defect structures at different charge states and their stability as a function of the electron chemical potential across the electronic band gap of crystals at operating temperature and oxygen partial pressure conditions. Through investigation of the local bonding configurations and projected electronic density of states of the 3H related charged defects, including 3H interstitial (3Hint), 3Hint bound with an oxygen interstitial, and substitutional 3H defects on the Li, O, and M (M = Al, Ti, and Zr) sites (3HLi, 3HO, and 3HM), the 3H states of these defects were distinguished, i.e. triton (3H+), neutral 3H (3H0), and tritide (3H−). The two stable 3H related defects under the Li2O-LiOH rich condition at T = 1000 K, between P(O2) = 10−5 and 10−25 atm were found to be 3H interstitial (3Hint) with charge of +1 and neutral 3HLi (substitutional 3H on the Li site) in γ-LiAlO2, Li2ZrO3, and Li2TiO3 materials. For Li2TiO3, the stability of the 3HO with charge of +1 and 3HM with charge of −3 becomes comparable to that of 3Hint atm at P(O2) = 10−25 atm. Overall, our results indicate the energy cost to form the stable 3H defect species (3Hint with charge of +1 and neutral 3HLi) in γ-LiAlO2 material are higher than those of Li2TiO3 and Li2ZrO3 materials at T = 1000 K, between P(O2) = 10−5 and 10−25 atm. Interactions between the point defects and the 3H species as a function of the electron chemical potential across the band gap of the materials were also examined. All the three materials exhibit strong interactions between the 3Hint and the O vacancy, Li vacancy, and O interstitial defects with the strongest interactions found in the case of γ-LiAlO2 phase, indicating a higher tendency to trap 3H at such point defects in γ-LiAlO2.

Improving the optimization algorithm of NTCOC for application in the HCSB blanket for CFETR Phase II

Blanket is the key component of Chinese Fusion Engineering Test Reactor (CFETR). In our previous work, considering that the radial arrangement of the internal functional zones has the greatest influence on the blanket performances and is also the most important basis of the 3D design, NTCOC, a Neutronics/Thermal-hydraulic Coupling Optimization Code which adopts 1D neutronics and 2D thermal-hydraulic simplified calculation models, has been independently developed for facilitating the radial build design and optimization of the CFETR helium cooled solid breeder (HCSB) blanket. Through setting three leading design requirements as the optimization criteria, this code has been applied on optimizing the radial build arrangement of a conceptual design of HCSB blanket for CFETR Phase I, which preliminarily verifies the reasonability of the code. However, recently, the design phase of CFETR has gradually changed from I to II, the fusion power of which will be quintuple to the previous I. This means that the internal nuclear power deposition in each blanket component under II operation condition will be much larger than the previous I. Therefore, much larger coolant inlet mass flowrate will be needed and more breeder units should to be inserted, both of which will make the total pressure drop of blanket for Phase II much larger than the previous I and even exceed the limit of CFETR HCSB blanket greatly. This can’t be sustained in CFETR, a power plant that should produce electricity. To solve this problem, in this paper, the previous integrated neutronics/thermal-hydraulic optimization algorithm is improved by adding the new pressure drop limit into the optimization criteria. Then the improved code has been applied on optimizing the radial build arrangement of a conceptual design of HCSB blanket for CFETR Phase II through constant and variable step length two methods. The calculation results verifies the feasibility and reasonability of the improved optimization algorithm preliminarily.

A study regarding the optimal radial build of a low-aspect-ratio tokamak fusion system

System parameters and the optimal radial build of a low-aspect-ratio (LAR) tokamak fusion system were found through the coupled analysis of a tokamak system and neutron transport. In a configuration with only an outboard breeding blanket, the minimum major radius to produce a given fusion power was determined by the shielding requirements and the magnetic field at the toroidal field (TF) coil. With a confinement enhancement factor H = 1.4, Q > 10 was possible for fusion power greater than 800 MW with an aspect ratio of A = 1.5; however, Q > 10 was possible for fusion power greater than 1800 MW with an aspect ratio of A = 2.0. The outboard radial build was determined by the tritium breeding and shielding requirements. The tritium breeding capability of blanket concepts proposed for testing in the International Thermonuclear Experimental Reactor (ITER) was evaluated by varying the outboard blanket thickness and the degree of lithium-6 (Li-6) enrichment. Cases with a smaller aspect ratio exhibited better performance since the number of fusion neutrons that contributed to tritium breeding were larger than the case with a larger aspect ratio. Among the blanket concepts, a helium (He)-cooled solid breeder (HCSB) concept showed the best tritium breeding capability and thus allowed for a smaller system size.

High temperature oedometric compression of Li2TiO3 pebble beds for Indian TBM

Lithium Metatitanate (Li2TiO3 or LMT) and Lithium Orthiosilicate (Li4SiO4 or LOS) are the prominent among the suitable candidate materials for breeders. Robust design of the breeder blankets requires a thorough understanding of the thermo-mechanical response of the breeder materials at different loading conditions. In this context, the material characterization plays a vital role. Solid breeder concept of various member nations comprises of pebbles in steel container. Determination of effective properties of the pebble beds helps in understanding and formulation of constitutive relations for such granular systems. In order to develop simulation tools to understand the complex thermo-mechanical response of the breeder units, the properties of the material should be known. Hence, the objective of this work is to characterize crush load of Li2TiO3 pebbles and investigate the thermo-mechanical response of the pebble beds. The study involves determining the crush load of individual pebbles and pebble-bed elastic modulus. Uniaxial oedometric tests are performed to investigate the influence of temperature on the effective elastic modulus of the pebble beds.

Preliminary steady and transient analysis for the CFETR helium cooled solid blanket system with RELAP

Abstract The helium cooled solid breeder (HCSB) blanket, which is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been developed in University of Science and Technology of China (USTC). And the blanket cooling system (HCS) is also designed for HCSB. The steady-state results show that the blanket system has a good relevance to the design parameters. The transient analysis of possible accidents should proceed to evaluate its safety performance. The accident analysis in this paper adopts the ITER-like conditions and the ex-vessel loss of coolant accident (LOCA) case is investigated with RELAP5 code. The results show that the ex-vessel LOCA with detection couldn't cause serious consequence, however the ex-vessel LOCA without detection has potential to melt the blanket that we should take measures to avoid it.

Progress in EU Breeding Blanket design and integration

In Europe (EU), in the frame of the EUROfusion consortium activities, four Breeding Blanket (BB) concepts are being developed with the aim of fulfilling the performances required by a near-term fusion power demonstration plant (DEMO) in terms of tritium self-sufficiency and electricity production. The four blanket options cover a wide range of technological possibilities, as water and helium are considered as possible coolants and solid ceramic breeder in combination with beryllium and PbLi as tritium breeder and neutron multipliers. The strategy for the BB selection and operation has to account for the challenging schedule of the EU DEMO, the ambitious operational requirements of the BBs and the still large development needed to have a BB qualified and licensed for operating in DEMO. In parallel to the continuous design efforts on the four blanket concepts, their integration in-vessel and ex-vessel has started. On the one hand it has become clear that despite the numerous systems to be integrated in-vessel the protection of the blanket first wall has to be addressed with highest priority. On the other hand the ex-vessel interfaces and the requirements imposed by the blanket to the primary heat transfer system and to the PbLi loop components have a considerable impact on the whole DEMO Plant layout.The aim of this paper is: to present the strategy for the DEMO BB down selection and BB operation in DEMO; to describe the status of the design evolution of the four EU BB concepts; to provide an overview of the challenges of the in-vessel and ex-vessel integration of the main systems interfacing the BBs and describe their design status.

Mechanism of vacuum-annealing defects and its effect on release behavior of hydrogen isotopes in Li2TiO3

Li2TiO3 is one of the most promising candidates among solid breeder materials. However, defects introduced into Li2TiO3 will act as the strong trapping sites for tritium. In the present study, mechanism of vacuum-annealing defects and its effect on release behavior of hydrogen isotopes in Li2TiO3 were investigated by means of X-ray diffraction, Raman spectroscopy, electron spin resonance and thermal desorption spectroscopy. The color of samples becomes dark blue and the defects were found to be introduced into Li2TiO3 when annealed in vacuum. This color change suggests the change from Ti4+ to Ti3+ due to decrease in oxygen content. The color recovers to white again after annealing in air. X-ray diffraction and Raman spectroscopy results indicate that there are no modifications on Li2TiO3 crystal phases, but on crystallinity. The main vacuum-annealing defects are E-centers and no other obvious types of defects were observed from electron spin resonance. Based on the experimental results, the production of defects by annealing in vacuum should be satisfied to the following conditions: (1) Li2TiO3 has been exposed in air more than 1 day; (2) Li2TiO3 must be annealed at the temperature higher than 300 °C; (3) Li2TiO3 should be annealed in vacuum lower than 10 Pa. E-centers formed under vacuum-annealing processes have considerable effects on release behavior of hydrogen isotopes investigated by thermal desorption spectroscopy and further should be considered in future fusion reactor. The present work gives some suggestions for future fusion reactors: (1) Li2TiO3 should be preserved in vacuum or kept from water vapor; (2) Li2TiO3 should be annealed at high temperature to remove the adsorbed water before loading into the facility, and must be finished within two days to avoid defects coming from reduction; (3) Li2TiO3 should be improved by adding more oxygen or other elements to refrain from defects introduced by reduction reaction.

First principle study of tritium trapping at oxygen vacancies in Li4SiO4

Tritium trapping at point defects is a major issue concerning tritium extraction efficiency in solid state breeder materials in future fusion reactors. Here, tritium trapping behaviors at oxygen vacancies in Li4SiO4 have been investigated by density functional theory simulations. Formation energies have been calculated to determine the stability of defects in various charge states. Density of states, charge distribution and atomic charge has been calculated to investigate the mechanism of defect formation. The results showed that the neutral and 2 + charged oxygen vacancies are the most stable species and are both diamagnetic. The 1 + charged paramagnetic oxygen vacancy (E′ center) is less stable. In addition, the tritium-trapped O-vacancy is most stable when the defect complex is 1 + charged. The trapped tritium is bonded to Si forming a T-SiO3 unit. Other charge states of the complex may lead to unstable defect structures and high formation energies, forcing tritium to escape the vacancy.

Influence of O-related defects introduced by reduction on release behavior of hydrogen isotopes in Li2TiO3

In fusion reactors, Li2TiO3 is one of the most promising candidates among solid breeder materials. However, Li2TiO3 will be reduced after heating under a reducing atmosphere resulting in a deficiency of oxygen. In the present study, the O-related defects introduced by reduction and their effect on the release behavior of hydrogen isotopes in Li2TiO3 were investigated by means of Raman spectroscopy, electron spin resonance and thermal desorption spectroscopy. O-related defects were confirmed as E-centers by electron spin resonance. The concentration of defects increased as the exposing temperature increased and then decreased when the temperature increased higher above 750 °C. The color of Li2TiO3 samples changed from white to dark blue after heating under deuterium and recovered to white again after annealing in air. This color change suggested a change from Ti4+ to Ti3+ due to a decrease in the oxygen content. Raman spectroscopy results indicated that there are no modifications in Li2TiO3 crystal structure, but on crystallinity. Thermal desorption spectroscopy showed the release behavior of deuterium has been affected considerably by O-related defects.

Plasma Control Requirements for Commercial Fusion Power Plants: A Quantitative Scenario Analysis With a Dynamic Fusion Power Plant Model

The authors constructed a dynamic simulation model of a nuclear fusion power plant on Modelica language to obtain fundamental knowledge on the plasma control requirements for the future commercial fusion power plants. The fusion power plant model was designed with a 1500-MW thermal output tokamak reactor with He-cooled Li2TiO3 solid breeder blanket (coolant outlet conditions: 8 MPa and 515.8 °C). A superheated Rankine cycle was designed to achieve the electrical output of 485.38 MW with the operating pressure of 20.5 MPa. Two plasma output patterns, a step decrease of power and a single pulse decrease of power, were simulated to assess the response of the power plant. A sudden step decrease in fusion neutron led to an immediate decrease in the blanket temperature and the first coolant temperature. In order to avoid the sharp temperature drop, a need for a turbine bypass mechanism or a He coolant boiler bypass mechanism was indicated. On the other hand, because of the delay in the plant responses, the deviation of the electrical output from steady state could be minimized by recovering the plasma output in few tens of seconds. Based on the findings, a new diagram was presented that illustrates an important plasma control requirements for future commercial fusion power plants.

Steady State and Transient Thermal Analysis of the Updated Helium-Cooled Solid Breeder Blanket for CFETR

This paper presents the results of the steady state and transient thermal analysis of the updated helium-cooled solid breeder blanket for the Chinese fusion engineering test reactor (CFETR). The updated design of the helium-cooled solid breeder blanket for the CFETR has been described. The commercial finite element method code ANSYS 15.0 is used for the thermal analysis in this paper. The steady state thermal analysis of the updated blanket has been performed, showing that the temperatures of different materials of the blanket module are below the corresponding temperature limits. The 3-D transient thermal analysis of the updated helium-cooled solid breeder blanket for the CFETR under normal pulsed operation has been conducted. The temperature of different components during flattop burning time calculated by transient thermal analysis matches well with that of steady state thermal analysis. Furthermore, the transient thermal behavior of the blanket under the in-box loss of coolant accidents condition are also described and critically discussed.

Corrosion characteristics of reduced activation ferritic-martensitic steel EUROFER by Li2TiO3 with excess Li

In a solid breeding blanket, ceramic breeder pebbles are in contact with reduced activation ferritic martensitic (RAFM) steel at high temperatures for years and accordingly form a corrosion layer on the blanket structural steel. Present study focuses on corrosion characteristics of EUROFER97 RAFM steel by an advanced breeder material of Li2TiO3 with excess Li (initial ratio of Li/Ti = 2.2) at 623, 823, and 1073 K under sweep gas (He + 0.1% H2) flow. Formation of a thick oxide double layer was found on the surface of the EUROFER plates heated at 823 and 1073 K, while the corrosion layer formed at 673 K was < 1 µm even after 56 days. On the other hand, the changes that appeared on the contacted surface of the breeder pellet were insignificant. The growth of the corrosion layer shows that the corrosion caused by inward migrations mainly of Li and O from the breeder material will have little influence on mechanical property of the blanket structural steel.

The assessment of shutdown dose rate and radioactive waste of HCSB during its replacement in CFETR

Chinese Fusion Engineering Testing Reactor (CFETR) is a new test superconducting tokamak in China aimed to obtain the state-of-the-art technologies and bridge the gaps between ITER and DEMO. The helium cooled solid breeder blanket (HCSB) is one of three main blanket candidates for CFETR. During reactor operations, the highly radioactive nuclides will be produced by the blanket components which are located in the high-energy neutron radiation field. In order to protect the workers from too high radiation and ensure the normal operation of machinery, the Shutdown Dose Rate (SDR) of HCSB during its replacement is evaluated based on the rigorous two-step (R2S) method. Besides, according to the recommended standards provided by Safety and Environmental Assessment of Fusion Power (SEAFP) of the EU fusion program, the initial evaluation of the high level radioactive waste of the blanket is carried out as well. The results show that, based on current design of HCSB, the maximum SDR of the blanket is decreased to 4728 Sv/h after the reactor shutdown one day. And after 50 years of cooling, most of the structural materials can be recycled by simple remote operation.

Numerical study on the purge gas flow and heat transfer characteristics in helium cooled solid breeder blanket of CFETR

The helium cooled solid breeder blanket module was proposed and designed to achieve the goal of self-sustainable tritium breeding in China Fusion Engineering Test Reactor (CFETR), which was regarded as the transition from ITER to DEMO fusion reactors. A series of preliminary analyses, including thermal hydraulics, neutronics, structural mechanics and materials, have been performed for the blanket module design. However, most literatures focus on the thermal hydraulic behaviors of high pressure helium coolant flowing in cooling channels, while the thermal hydraulic characteristics of low pressure purge gas flowing slowly in pebble beds of tritium breeding zones has not been investigated in detail. In this work, the three dimensional model of typical helium cooled solid breeder blanket module was established employing Computational Fluid Dynamics (CFD) method. The corrected Ergun equation was adopted to simulate the coolant flow features in the porous medium. The temperature and flow field distributions in the First Wall (FW), neutron multiplier zones and tritium breeding zones were achieved. The most critical parts in the blanket module during normal operation condition were illustrated. Besides, the influences of purge gas inlet velocity and tritium breeding pebble-bed ball diameter on the whole blanket module thermal hydraulic features were revealed. This work could contribute to the design and optimization of helium cooled solid breeder blanket in CFETR.

Numerical research on the coupling optimization design rule of the CFETR HCSB blanket using the NTCOC code

In this study, a three-step approach of ordinal 3D-1D&2D-3D is independently proposed to derive a coupling optimization design rule of the Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder (HCSB) blanket. Firstly, the three dimensional (3D) neutronics model of CFETR is built with all of the 15 blanket modules adopting the original radial build arrangement parameters. Based on this model, both the neutron wall loading (NWL), which will serve as the normalization factors for the following local neutronics calculations, and the tritium breeding ratio (TBR) distributions on each blanket module are obtained. Secondly, the radial build arrangements of all the 7 inboard and the 8 outboard blanket modules in an 11.25° toroidal CFETR sector are optimized by taking both the tritium breeding and thermal performances into account as the optimization objectives. NTCOC, a Neutronics/Thermal-hydraulic Coupling Optimization Code, which has been independently developed by Xi’an Jiaotong University for accelerating the radial build design and optimization of the fusion reactor HCSB blanket concept with adopting the simplified 1D neutronics and 2D thermal-hydraulic models, is used for these work. Thirdly, based on the final optimized schemes of all the blanket modules, a comprehensive research on the coupling optimization design rule of the CFETR HCSB blanket is performed in-depth and a coupling optimization design integration approach has been finally summarized. Finally, the 3D CFETR neutronics model is updated by replacing all of the original blanket radial build arrangement parameters with the final optimized schemes. The calculation results show that after the radial optimizations, the total TBR of all the 15 blanket modules has increased from the previous 1.092–1.274, which can well meet the self-sufficiency tritium breeding requirement of CFETR.

Development and application of a neutronics/thermal-hydraulics coupling optimization code for the CFETR helium cooled solid breeder blanket with mixed pebble beds

As one of the most important components of fusion reactor, breeding blanket mainly accounts for tritium self-sufficiency, shielding and fusion energy collection. The performance of the blanket is mainly determined by the internal structural arrangement and material composition. However, if we directly adopt a three-dimensional calculation model for the blanket design and optimization, not only will it consume extensive human and computer resources, but also make the computational process extremely complicated and lengthy. Considering that the radial arrangement of the functional zones, which has the greatest influence on the performance, is the basis of the detailed and accurate 3D design of the blanket, NTCOC, a Neutronics/Thermal-hydraulics Coupling Optimization Code, which adopts simplified 1D neutronics and 2D thermal-hydraulic calculation models, is developed for accelerating the radial build design and optimization of the blanket in this work. This code realizes the functions of both the automatic iterative updating the temperature-dependent cross-section libraries and the data transmission between the neutronics and thermal-hydraulic calculation results through the inner and external coupling between the MCNP4C and CFD software, respectively. Therefore, this can make the comprehensive optimization analyses progress much faster and more accurate. In addition, a variable step length method is adopted to further improve the coupling optimization speed of this code, and several leading design objectives are set as the optimization criteria instead of taking all the possible practical engineering constrains into account. Finally, this code has been applied to optimize the radially arranged zones of a conceptual design scheme of Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder blanket, which adopts mixed pebble beds as both the tritium breeder and neutron multiplier, under both the normal and critical conditions.

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module (TBM) program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development (R&D) is focused on two types of breeding blanket concepts: lead–lithium ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB) blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic R&D program for DEMO relevant technology development. In the HCCB concept Li2TiO3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept (case-1), the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept (case-2), the pebble bed is vertically (poloidal–radial) orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations. Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.