Nuclear Heat Research Articles

Cross sections, penetrating range and stopping power for charge particles emission from n + 52Cr

In a fusion reactor, DT reactions produce huge amount of neutrons which hit the structure material and produce heat and damages. To understand the production of nuclear heating and material damages, study of double differential cross section (DDX), energy differential cross section (EDX), penetrating range of incident particles and stopping power of target element was calculated for 52Cr(n, p) and 52Cr(n, a) reactions. The calculated data have been compared with the suitable experimental data collected from EXFOR and other. The contribution of compound, pre-equilibrium and direct reaction processes has been analyzed for the entire cross section calculation. The emission of charged particles at different angles has been discussed based on the reaction processes. Parameter adjustment has been applied to fit the experimental data.

Three-dimensional thermal-hydraulics/neutronics coupling analysis on the full-scale module of helium-cooled tritium-breeding blanket

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

Post-irradiation examination of the Sirius-1 nuclear thermal propulsion fuel test

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high-temperature gas-cooled fission reactor to provide thermal energy to heat the coolant, which also serves as the propellant. The fuel system consists of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, which has been irradiated in Idaho National Laboratory's (INL) Transient Reactor Test Facility (TREAT). This enabled evaluation of these materials under representative nuclear heating rates (∼95 K/s) and peak surface temperatures (∼2527 K). These surface conditions were indicative of achievement of the target peak internal temperature of approximately 2850 K. These tests named Sirius-1 (UN–W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced minor, stable cracks in the fuel specimen and spalling of surface material. Volatilized uranium ‘soot’ was found deposited on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and the production of a laminar microstructure on the edge of the fuel system. Overall, the specimen was stable and remained within a coolable geometry upon conclusion of multiple thermal cycles to prototypical operating temperatures.

Thermal-hydraulic safety analysis of radioisotope production in HANARO using MCNP6 and COMSOL multiphysics: A feasibility study

The High-flux Advanced Neutron Application Reactor (HANARO) produces radioisotopes (RIs) (131I, 192Ir, etc.) through neutron irradiation on various RI production targets. Among them, 177Lu and 166Ho are particularly promising owing to their theranostic characteristics that facilitate simultaneous diagnosis and treatment. Prior to neutron irradiation, evaluating the nuclear heating of the RI production target is essential for ensuring the thermal-hydraulic safety of HANARO. In this study, the feasibility of producing 177Lu and 166Ho using irradiation holes of HANARO was investigated in terms of thermal-hydraulic safety. The nuclear heating rates of the RI production target by prompt and delayed radiation were calculated using MCNP6. The calculated nuclear heating rates were used as an input parameter in COMSOL Multiphysics to obtain the temperature distribution in an irradiation hole. The degree of temperature increase of the 177Lu and 166Ho production targets satisfied the safety criteria of HANARO. The nuclear heating rates and temperature distribution obtained through the in silico study are expected to provide valuable insight into the production of 177Lu and 166Ho using HANARO.

Structure optimization of helium-cooled blanket for fusion reactor based on three-dimensional full-scale thermal-hydraulic analysis

Fusion energy is widely considered as the most promising method to solve the energy crisis, while the blanket is of vital importance in the fusion device due to its main function of thermal extraction and energy conversion. The high heat flux from plasma and the internal nuclear heat deposition are imposed on fusion blanket. Slice model or two-dimensional model is usually adopted for traditional design of fusion blanket, in which most side effects are neglected, introducing relatively large uncertainty for thermal removal capability of the blanket. In this study, the integral three-dimensional full-scale model is established based on the split-zone method for mesh generation. Based on the validated numerical method, the preliminary thermal-hydraulic analysis of the full-scale model shows that inadequate heat transfer and significant flow distribution difference are encountered for the initial blanket design, in which the full-scale effect is not considered. After that, the specific structure modifications are proposed for different blanket components to solve the abovementioned problems. The maximum temperature of first wall is reduced from 671°C to 639°C, which is under the material temperature limit of 650°C; the inverse V-shaped flow distribution in the cap is modified as more even flow in cap channels; the maximum temperature of the function zone in blanket is reduced from 999°C to 759°C, which is an important structure modification for the blanket. The integral three-dimensional full-scale thermal-hydraulic analysis is performed after the local structure modification. The results show that the function zone is cooled down obviously by the coolant in the side wall. Compared to the temperature limit of 900°C for lithium orthosilicate, there is a more rigorous limit for the beryllium material, since they share the common cooling capability layer by layer while the temperature limit for the latter is only 650°C. The three-dimensional full-scale thermal-hydraulic analysis in our research is valuable to evaluate the integral reliability of the fusion blanket.

Numerical study of MHD mixed convection flow in the EU DEMO WCLL breeding blanket

The Water-Cooled Lithium Lead (WCLL) breeding blanket is a proposed candidate option for European DEMO nuclear fusion reactor. The utilization of PbLi as the predominant material occupying the bulk region of the breeding blanket has received significant attention. The interaction between Lorentz forces (Ha≈9800), intense buoyancy forces (Gr≈1.77×1010), and pressure-driven flow (Re≈236.77), plays a crucial role in the magnetohydrodynamics(MHD) and heat transfer phenomena of the region. To replicate the dynamics of PbLi, taking into account the impact of the strong magnetic field used for plasma confinement and the non-uniform heat source generated by nuclear heating, a 3-D DNS numerical simulation has been employed. Owing to the influence of buoyancy forces, several medium-sized circulations that are elongated in the direction of the magnetic field occupy the plenum region which is situated between the cooling pipes. Moreover, the electrical interaction between the liquid metal and the conducting wall gives rise to high-velocity flow jets, which exhibits peak velocities that are increased with the magnitude of the magnetic field. To characterize the heat transfer occurring in different regions of the system, local Nusselt and Grashof numbers have been determined. Optimization strategies for the BZ cooling system geometrical configurations are proposed based on the estimation of heat transfer and prediction of tritium permeation.

Thermal evaluation of different integration schemes for solar-nuclear hybrid systems

Solar-nuclear hybrid systems have been proposed to meet the growing energy demand and mitigate the effects of global warming caused by fossil fuels. To achieve optimal design for these systems, an appropriate integration scheme is crucial. This article proposes three integration schemes to couple the solar system with the nuclear power plant and evaluates their performance based on energy and exergy evaluations. A reference design of the hybrid system is used, consisting of a 77.04 MWth solar tower system coupled with a 330 MWth System-Integrated Modular Advanced Reactor. Numerical results show that heating partial feedwater and saturated steam with solar energy is the best-integrated scheme, resulting in the maximum thermal efficiency and exergy efficiency for the solar-nuclear hybrid system. Comparing the standalone reactor power system to the solar-nuclear hybrid system reveals that using solar energy to heat the feedwater and saturated steam can greatly improve the energy efficiency of the power plant, increasing it to 34.97%. The solar heat to nuclear heat ratio is 25.01%, while the solar power to nuclear power ratio is as high as 44.42%, indicating effective utilization of solar heat contribution. Exergy loss distributions of the solar-nuclear hybrid system reveal that exergy losses mainly occur within the nuclear reactor and solar field, taking possession of 70% of all exergy losses.

Analysis of the nuclear loads on Chromium monoblock divertor target for DEMO

The Plasma Facing Components (PFCs) of DEMO divertor play a fundamental role for the heat removal and particle exhaust functions, during fusion plasma. This implies a very harsh operation environment characterized by the combination of complex loading and stress conditions, intense particle bombardment, high heat flux deposition and a significant neutron irradiation. In the framework of the European DEMO divertor project (WPDIV), several target design concepts alternative to the baseline ITER-like technology are under evaluation, such as Chromium block design. It consists of chromium block, tungsten armor tile and copper alloy cooling pipe, and it is very attractive from an activation reduction, waste management and safety point of view. Therefore, the assessment of the most impacting nuclear loads for the PFC design is a pivotal aspect for DEMO divertor R&D program and this study is fully devoted to the comprehension of the spatial distributions of the main nuclear quantities (radiation damage, He-production and nuclear heating density) on the overall divertor targets. Neutronics analyses have been performed with MCNP5 Monte Carlo code using both DEMO MCNP Water Cooled Lithium Lead (WCLL) and Helium Cooled Pebble Bed (HCPB) blanket models, with a fully heterogeneous representation of the divertor cassette and Chromium PFCs geometry. The results of this analysis show that nuclear heating density and radiation damage distributions are higher with WCLL blanket than HCPB, except for He-production which seems not significantly affected by the blanket configuration. For each nuclear quantity, the peak values are achieved in correspondence of the baffle region and especially on the external units of the targets, as expected, being this region the most exposed to irradiation.

Expert opinions on strengths, weaknesses, opportunities, and threats of utilizing nuclear reactor waste heat for water desalination

Using waste heat generated by nuclear reactors for non-electric applications is vital for reducing environmental costs and thermal discharges. Waste heat is a significant portion of the reactor's overall energy and could reach up to two-thirds of the output power depending on the reactor design and efficiency. Despite the benefits of using waste heat, especially for essential applications like water desalination in areas with limited freshwater resources, the potential utilization is not yet applied. This study examines the strengths (S), weaknesses (W), opportunities (O), and threats (T) of using nuclear reactor waste heat for desalination in the Arabian Gulf region across four main areas: technical, environmental, socioeconomic, safety, and regulatory. The identified SWOT factors are combined with the Analytic Hierarchy Process (AHP) to prioritize them based on input from relevant experts. The resulting integrated SWOT-AHP analysis identified global factor priorities and concluded with five consistent measures to support decision-makers in evaluating the feasibility of utilizing nuclear reactor waste heat for seawater desalination. Overall, the study revealed that regulation and relevant safety measures present the primary obstacle in this application, while enabling research and development was emphasized as the key factor to address these issues and increase confidence in decision-making.

The investigation of the neutronic responses of components in the K-DEMO

The neutronic response of consisting materials of the Korean fusion DEMOnstration reactor (K-DEMO) in its fusion neutron environment is a crucial consideration factor from the conceptual design stage of the K-DEMO. Especially in the design of in-vessel components (IVCs) of the K-DEMO that are placed in the most extreme neutron irradiation field, neutronic damage of constituent components is a major limiting factor that determines the lifetime of IVCs. From the analysis of the neutronic response of IVCs of the K-DEMO, the neutron wall loading (NWL) related to the tritium breeding ratio and nuclear heating of IVCs can be quantified to assess the self-sufficient supply of tritium, and thermal energy transferred from fusion neutrons, respectively. The calculated NWL shows that the harmonizing design of the cooling configuration of each blanket segment with the corresponding NWL is critical thermos-hydraulic design issue for the efficient utilization of thermal energy in the blanket. Another finding is that the double null magnetic field configuration and related blanket configuration with a water-cooled pebble bed of the K-DEMO make a self-sufficient tritium supply challenging. The implicated lifetime of the first plasma-facing tungsten wall of the K-DEMO is around 2 full power years (FPY) in the severely neutron-irradiated region. For the reduced activation ferritic/martensitic steel layers in the blanket, the lifetime of it is estimated around 4 FPY in the inboard region. Based on the response analysis results of this study, optimization of the design of the K-DEMO will continue iteratively in the future.

Process design of a thermochemical cycle for hydrogen production compatible with nuclear fusion heat sources

We quantitatively design a hydrogen-producing copper–chlorine thermochemical cycle that is thermally combinable with nuclear fusion reactors. The mass and heat balances throughout the cycle, including the hydrolysis, pyrolysis, and electrolysis reaction processes, accompanied by multiple separation steps, are numerically investigated. Through the process design, the feasibility of the practical operation of the hydrogen production cycle is presented, and the thermal and electric power required for the operation is evaluated. As we first design a process with straightforward employment of the exact reaction condition from an earlier experimental study of the electrolysis process, the heat required for the condensation of CuCl2 solution is found enormous. By modifying the process condition with an increased electrolyte concentration, to reduce the amount of H2O to be evaporated in the CuCl2 condensation, the total heat for the cycle significantly decreases from 1270 GJ·h−1 (20.3 MJ·mol-H2−1) to 204 GJ·h−1 (3.26 MJ·mol-H2−1). This value is still larger than the heat required for hydrogen production by H2O electrolysis, 48.2 GJ·h−1 (0.772 MJ·mol-H2−1), but an extrapolation towards the saturated CuCl concentration achieves 46.0 GJ·h−1 (0.736 MJ·mol-H2−1). For energy-cost performance improvement of the thermochemical cycle, the development of the electrolytic cell operable with a high concentration of CuCl aqueous solution is thus found to be of primary effectiveness.

Thermal radiation impact on chemical reactive flow of micropolar nanomaterial subject to Brownian diffusion and thermophoresis phenomenon

Background and objectiveEntropy generation (a novel prospective) in different thermodynamical processes has usefulness in polymer process. Importance of entropy optimization is noticed in nuclear reactions, heat exchangers, biological system, electronic cooling, combustion, porous media, thermal systems, turbo machinery, turbine systems and many others. Entropy minimization has been utilized for assessment of thermal energy through thermodynamical processes. Entropy minimization rate is employed to boosts up efficiency of thermodynamical system. Therefore, present investigation is aimed to investigate optimized radiative flow of micropolar nanofluid due to exponential stretched curved surface. Heat expression with heat generation and radiation is explored. Moreover, interesting aspects of homogenous and heterogeneous chemical reactions are also presented. For nanofluid characteristics the Buongiorno's model is incorporated. Velocity and thermal slip conditions are explored. Mathematical modeling of flow expressions are presented through curvilinear coordinates. Entropy rate is modeled employing second law of thermodynamics. MethodologySuitable relations have been employed to get the dimensionless systems. Resultant dimensionless systems are computed through ND-solve technique. ResultsGraphical analysis for sundry variables on liquid flow, entropy generation, temperature, microrotation velocity and concentration are addressed. An opposite scenario for microrotation and velocity holds against material parameter. Heat transport rate augments in presence of slip constraint. Change in radiation constant and thermophoresis variable correspond to higher temperature. Entropy generation has enhancing trend for radiation parameter. Concentration has opposite scenario for random and thermophoresis diffusion variables. Larger thermal ratio parameter enhances entropy generation.

Numerical Investigation of Fluid Flow and Heat Transfer Characteristics over Double Backward-Facing Step with Obstacles

Many engineering device, including nuclear reactors, heat exchangers, and electronic circuit boards, exhibit flow separation and reattachment. To reduce thermal stress impacts, the components of these devices, which resemble stepped channel designs and are susceptible to high heat flux, must be properly cooled. The present work numerically investigates the fluid flow and heat transfer characteristics through a double backward-facing step with elliptic obstacles located after each step. For this purpose, equations governing fluid flow and heat transfer are solved in a Cartesian framework using an in-house code based on streamline upwind/Petrov-Galerkin finite element method. The effect of various parameters (axis ratios (0.25, 0.5, 0.75, 1), vertical location of obstacles (0.38, 0.577, 0.769), Reynolds number (300, 500, 800, 1000)) on fluid flow and heat transfer characteristics in the channel with and without obstacles are compared and quantified. The recirculation region length of the double backward-facing stepped channel with obstacles decreases after the first and second steps when compared with no obstacles. The local Nusselt number distribution along the corner regions of the stepped wall is enhanced due to obstacles. The axis ratio of obstacles has less influence on the convective heat transfer enhancement than the vertical location of the obstacles.

Study on neutronics modeling with 22.5° model using ANSYS for CFETR

The detailed China Fusion Engineering Test Reactor (CFETR) 22.5° computer aided design (CAD) model is very difficult to convert into Monte Carlo N Particle Transport Code (MCNP). Manually writing MCNP input data is complicated, which is not only time-consuming but also cannot guarantee accuracy. Therefore, in order to improve the efficiency and accuracy of model transformation, modeling with CAD using CATIA is introduced, and MCNP files are converted by ANSYS. This is because ANSYS has a function that converts CAD “stp” format to MCNP input in the geometry section. Meanwhile, ANSYS can also reverse the converted MCNP input file to inspect which module has the problem. Compared with the software platform that can automatically cut, although the CATIA-to-ANSYS method is inferior in terms of automatic operation, it has advantages in accuracy and quickly dealing with error modules. Moreover, it can also perform parametric modeling in CATIA, which facilitates the optimization of the blanket structure. In this paper, the detailed CFETR 22.5° model was developed, and then parametric modeling of the blanket based on CATIA was performed. Finally, a detailed neutronics model is obtained by ANSYS transformation and inspection. Some representative models were initially validated by comparing volume changes before and after conversion. Then, the final neutronics model was used to calculate the nuclear analyses, including the neutron wall loading, fast neutron flux, and nuclear heating on the inboard side. The results show that the volume of the transformed model is basically consistent with the original model, and the error of results is small.

Front-Face RH Compatible Structure Design and Thermal Analysis for CFETR Divertor Dome

The China Fusion Engineering Test Reactor (CFETR) is a device developed to verify the engineering feasibility of a fusion reactor. For CFETR, the divertor is an important plasma-facing component, whose main function is to exclude impurities and remove plasma heat. In addition, the requirement for remote handling (RH) maintenance must be satisfied because of the level of radioactivity in the vacuum vessel after shutdown. The dome is an important component of the divertor, whose main function is to isolate impurity particles as well as to improve the ability of excluding particles. In the optional dome design, a hybrid divertor-blanket concept, a front-face RH compatible structure in plasma-facing units (PFUs), and a RH maintenance scheme for the main bolt are proposed. The vulnerable targets can be replaced directly and thus reduce the RH maintenance time. The dome needs to withstand the heat flux of 10 MW/m2 and nuclear heat in the condition of 1.5 GW of fusion power in the engineering design requirements. Because of the RH compatible structure, higher requirements are demanded for the design of the dome cooling system. In this study, the cooling system and the customized heat transfer structure of dome PFUs are designed to guarantee the maximum heat removal level. The steady-state thermal analysis shows that the cooling system fulfills the design requirements. The concept of the hybrid divertor-blanket and the front-face RH compatible structure for the divertor target have certain reference significance and value for the engineering design and RH maintenance research for the fusion reactor divertor in the future.

Neutronic response of the neutral beam after EAST NBI upgraded

In order to improve the heating effect of neutral beam injector, the neutral beam injector (NBI) of the Experimental Advanced Superconducting Tokamak (EAST) is planned to be upgraded. After the EAST NBI is upgraded, the neutron emission intensity will increase. Neutron leaking from the neutral beam port penetrate the beamline, resulting in the material radiation damage. The neutronic response has been investigated by using the Monte Carlo particle transport code MCNP with the two dimensional RZ model of the EAST NBI. The results demonstrate that a 40% increase of the neutron emission rate has been attained with the upgraded EAST NBI. However, the simulations for the activation of different materials, displacement damage and nuclear heating of the neutral beam still satisfy the requirement of material safety performance. The calculations of nuclear response show that the materials can meet the requirements of the NBI steady state operation after the EAST NBI updated.

Construction of hot helium leak test facility for ITER blanket shield block

The ITER Blanket Shield Block (SB) is one of the main in-vessel components that provides nuclear shielding and removes nuclear heating for the Vacuum Vessel (VV) and external vessel components. The SB, which contains internal cooling channels, is manufactured from 316L(N)-IG stainless steel, and the internal channels are designed to provide the required thermal-hydraulic performance while aiming for a water/steel ratio that results in good nuclear shielding with headers and welded cover plates. Therefore, the soundness of all SB products' welded joints must be confirmed to eliminate the risk of leakage during manufacturing and the Factory Acceptance Test (FAT). However, a test facility capable of testing large components at high temperatures has never been constructed anywhere. The ITER Korea Domestic Agency (KODA) has therefore successfully completed the construction and commissioning of two Hot Helium Leak Test (HHLT) facilities, and all SB series products are being entirely tested under RT and an elevated temperature of 250 °C with a maximum operational pressure of 5.0 MPa. This paper provides information on the design features of the two HHLT facilities and lessons learned from the commissioning test.

Capacity and operation optimization of a low-temperature nuclear heating system considering heat storage

Integrating heat storage devices with the pool-type low-temperature heating reactor (PLTHR) can improve the flexibility of PLTHR for load following. A novel low-temperature nuclear heating system (HST-NHS) is proposed, which combines the DHR-400 heating reactor system, a heat storage tank (HST), and a gas boiler (GB) to achieve real-time matching between heating supply and demand. A mixed integer nonlinear programming (MINLP) optimization model with the annual cost (AC) as the optimization objective is established. An operation strategy that considers the device priority is proposed. The capacity and operation of HST-NHS are simultaneously optimized using the CPLEX solver. The optimization results indicate that an HST with a rated capacity of 1324.14 MWh and a GB with a rated capacity of 201.69 MW can provide the operational flexibility needed for the HST-NHS throughout the heating period. When compared with comparison schemes 1 (DHR-400 and HST), 2 (DHR-400 and GB), and 3 (pure gas boiler heating system), HST-NHS achieves significant annual cost savings. Finally, a sensitivity analysis of device price and fuel price is conducted, revealing that the unit price of the DHR-400 heating reactor system and natural gas price are the main factors affecting the AC and dynamic payback period (DPP) of HST-NHS. This work offers theoretical guidance for the subsequent engineering application of the DHR-400 heating reactor system.

Research Progress on the Influence of Thermo-Chemical Effects on the Swelling Pressure of Bentonite

The swelling pressure of bentonite changes dramatically due to diffused nuclear radiation heat and underground osmosis, causing the failure of the buffer isolation layer in deep geological repositories for the disposal of high-level radioactive waste. A detailed overview of the relevant research results on the swelling pressure variation of bentonite under thermo-chemical effects is presented in this paper. The results showed that the values of the swelling pressure obtained by different test methods are dissimilar. The swelling pressure of bentonite decreased with the increasing pore solution concentration; nevertheless, the effect of temperature on the swelling pressure is still controversial. At the micro-level, crystal layer swelling and double- layer swelling are generally considered to be the main factors affecting the swelling pressure; the pore structure and water distribution of bentonite will change owing to thermo-chemical effects. At the macro-level, involving intergranular stress, a mechanical parameter was proposed to explain the mechanism of the changes in the swelling pressure of bentonite. Finally, future research directions for the study of the evolution of bentonite swelling properties under thermo-chemical effects are proposed, based on the current research results.

Computationally efficient methodology of 3D thermal-hydraulic analysis of Diagnostic Shielding Module #2 of ITER Equatorial Port #11

The process of design justification of Diagnostic Shielding Modules (DSMs) of port plugs of the International Thermonuclear Experimental Reactor (ITER) includes a stage of thermal-hydraulic computational analysis. The ITER tokamak operates in a pulsation mode that means the nuclear heating affecting DSMs is unsteady, and transient analysis is required. Three-dimensional (3D) approaches for fully coupled thermal-hydraulic analysis, which imply simultaneous computations for fluid and solid parts of a 3D DSM model, lead typically to “heavy” numerical models being very expensive from the computational point of view, especially for transient simulations. In order to significantly reduce the consumption of computational resources, the present 3D thermal-hydraulic analysis of DSM #2 of ITER Equatorial Port #11 was performed using one-way coupling. Firstly, steady-state CFD computations of turbulent flow were performed for the front part of DSM cooling system, which is a high-density network of channels of complex geometry. Obtained in CFD analysis map of wall shear stress on surface of channels was then converted into a map of local heat transfer coefficient using a simple and conservative approach based on known empirical correlations for turbulent flow in a circular pipe. For the back part of DSM cooling system, which basically consist of many relatively long channels, an individual (single) value of heat transfer coefficient for each of the channels was directly evaluated using empirical correlation. The heat transfer coefficient maps obtained for both parts of DSM cooling system were used finally as a boundary condition in transient thermal analysis of the solid part of the DSM model. Details and justifications of the described computationally-efficient methodology are presented.