Hot Channel Factors Research Articles

Conceptual design of narrow rectangular channel test section for research reactor application

Conceptual design was carried out on test section for studying heat transfer characteristics of narrow rectangular channel for research reactor application. Thermal hydraulic design requirements were specified through literature study on mid- to high-powered research reactors currently under operation or design. Engineering design was then carried out in streamlined fashion with help of in-house design code and computational fluid dynamics analysis. Considering various constraints and uncertainties in terms of engineering hot channel factors, sensitivity studies were performed and realistic test section design was obtained. For final design, stochastic operable heater power map is generated by merging individual operable ranges from each constraint, with random sampling of hot channel factors.

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Evaluation of hot channel factor for sodium-cooled fast reactors with multi-physics toolkit

The evaluation of hot channel factor (HCF) is of great significance to the quantification of safety margins for reactor designs. In this paper, HCFs for a sodium-cooled fast reactor (SFR) are evaluated with the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit, which is developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of DOE for multi-physics reactor performance and safety simulations. The high-fidelity neutronics and thermal hydraulics solvers PROTEUS and Nek5000 in the SHARP toolkit are coupled to perform the multi-physics simulations for HCF evaluation. The HCFs induced by cladding manufacturing tolerance, fissile content mal-distribution, wire orientation and uncertainties on the cladding, coolant, and fuel properties are evaluated for a reference core SFR design (AFR-100). The HCFs calculated with the SHARP toolkit are compared to legacy HCFs for similar reactor types. The comparison demonstrates the reduction or elimination of modeling uncertainties in the calculation of HCFs using high fidelity advanced modeling and simulation tools without the need of expensive experiments. Moreover, the reduction of the uncertainties on HCFs evaluation allows an increase in nominal parameters and safety margin, which in turn improves the economic competitiveness of the SFR.

Impact of the homogenization level, nodal or pin-by-pin, on the uncertainty quantification with core simulators

This paper assesses the impact of the homogenization level (nodal or pin-by-pin) on the uncertainties predicted by core simulators, highlighting the need of calculations at pin level for a reliable estimate of uncertainties in hot channel factors. In order to perform the analysis, two methodologies for nuclear data uncertainty quantification in stand-alone neutronics calculations have been implemented in the core simulator COBAYA. The first one is based on first-order perturbation theory and allows sensitivity/uncertainty analysis in the multiplication factor. The second one is based on random sampling and allows the uncertainty propagation in all responses computed by the code. Both methodologies are used in conjunction with SCALE system, which provides the capabilities to propagate the nuclear data uncertainties into the few-group constants required for the diffusion calculations. These methodologies are validated using a simplified pin-cell and a fuel assembly, and then applied to a full 3D core in the context of the OECD/NEA UAM benchmark (Uncertainty Analysis in Best-Estimate Modeling for Design, Operation and Safety Analysis of LWRs).

Thorium-based mixed oxide fuel in a pressurized water reactor: A burnup analysis with MCNP

A three-dimensional MCNP model of a Westinghouse-type 17×17 PWR was used to assess the feasibility of enhancing the current fuel supply with a thorium-plutonium mixed oxide fuel (ThMOX). A conventional UO2-fueled configuration was compared to a configuration which incorporated ThMOX fuel. The safety of the ThMOX configuration was compared to that of the UO2 configuration at several time steps of interest within all three cycles (beginning of cycle, peak excess reactivity of cycle, and end of cycle) using the following metrics: axial and radial nuclear hot channel factors FZNandFRN, moderator and fuel temperature coefficients (MTC and FTC), delayed neutron fraction (βeff), and shutdown margin (SDM). Additionally, the performance of the ThMOX configuration was assessed by tracking cycle lengths, the amount of plutonium destroyed, and fission product poison concentration.The ThMOX configuration’s FZN (1.455±0.222) was slightly less favorable than that of the UO2 configuration (1.515±0.246), though its FRN was statistically identical (3.623±0.179 compared to 3.430±0.066 for UO2). The delayed neutron fraction was another area of concern for the ThMOX configuration. It was lower than that of the UO2 configuration, but the lowest βeff occurred at the beginning of Cycle 3 (3.324E−03±8.919E−07). The βeff of the UO2 configuration was 46% higher (6.177E−03±1.510E−06) at this time step. Similarly, the ThMOX configuration’s SDMs were less favorable than those of the UO2 configuration, though even the lowest (3971±43pcm) remained well within the NRC prescribed limit of 1300pcm. However, the MTC of the ThMOX configuration was more favorable for all time steps except the beginning of Cycle 3, and even then it was negative as required by the NRC (−2.325±0.00098pcm/°F). The FTC results were also more favorable for the ThMOX configuration. Its highest FTC value (−6.718±0.001pcm/°F at the peak of Cycle 1) was more negative than all but the lowest FTC of the UO2 configuration (−7.335±0.002pcm/°F at the end of Cycle 3).Overall, incorporating ThMOX fuel into the fuel supply for a conventional PWR seems feasible from a neutronics standpoint. However, this analysis would benefit from a more detailed burnup simulation and optimization of the ThMOX fuel composition. Also, a full thermal hydraulic analysis in conjunction with a neutronics analysis is required.

Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10MW with engineering hot channel factors included. Thus, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.

Development of a transverse mixing model for large scale impulsion phenomenon in tight lattice

Tight-lattice is widely adopted in the innovative reactor fuel bundles design since it can increase the conversion ratio and improve the heat transfer between fuel bundles and coolant. It has been noticed that a large scale impulsion of cross-velocity exists in the gap region, which plays an important role on the transverse mixing flow and heat transfer. Although many experiments and numerical simulation have been carried out to study the impulsion of velocity, a model to describe the wave length, amplitude and frequency of mixing coefficient is still missing. This research work takes advantage of the CFD method to simulate the experiment of Krauss and to compare experiment data and simulation result in order to demonstrate the feasibility of simulation method and turbulence model. Then, based on this verified method and model, several simulations are performed with different Reynolds number and different Pitch-to-Diameter ratio. By fitting the CFD results achieved, a mixing model to simulate the large scale impulsion phenomenon is proposed and adopted in the current subchannel code. The new mixing model is applied to some fuel assembly analysis by subchannel calculation, it can be noticed that the new developed mixing model can reduce the hot channel factor and contribute to a uniform distribution of outlet temperature.

CFD investigations of helical wire-wrap fuel pin bundle and its comparison with straight wire bundle

Computational Fluid Dynamics (CFD) investigations of a fast reactor fuel pin bundle wrapped with helical and straight spacer wires have been carried out and the advantages of using helical spacer wire have been assessed. The flow and temperature distributions in the fuel pin bundle are obtained by solving the statistically averaged 3-Dimensional conservation equations of mass, momentum and energy along with high Reynolds number k-ε turbulence model using a customized CFD code CFDEXPERT. It is seen that due to the helical wire-wrap spacer, the coolant sodium not only flows in axial direction in the fuel pin bundle but also in a transverse direction. This transverse flow enhances mixing of coolant among the sub channels and due to this, the friction factor and heat transfer coefficient of the coolant increase. Estimation of friction factor, Nusselt number, sodium temperature uniformity at the outlet of the bundle and clad hot spot factor which are measures of the extent of coolant mixing and non-homogeneity in heat transfer coefficient around fuel pin are paid critical attention. It is seen that the friction factor and Nusselt number are higher (by 25% and 15% respectively) for the helical wire wrap pin bundle compared to straight wire bundle. It is seen that for 217 fuel pin bundle the maximum clad temperature is 750 K for straight wire case and the same for helical wire is 720 K due to the presence of transverse flow. The maximum temperature occurs at the location of the gap between pin and wire. The ΔT between the bulk sodium in the central sub-channel and peripheral sub-channel is 30 K for straight wire and the same for helical wire is 18 K due to the presence of secondary transverse flow which makes the outlet temperature more uniform. The hotspot factor and the hot channel factors predicted by CFD simulation are 10% lower than that used in conventional safety analysis indicating the conservatism in the safety analysis.

Neutronics and sub-channel thermal-hydraulics analysis of the Iranian VVER-1000 fuel bundle

In the present research, sub-channel thermal-hydraulic analysis of the Bushehr Nuclear Power Plant (BNPP) core, a Russian VVER-1000 type, at hot full power (HFP) and steady state BOC conditions is presented herein using modified COBRA-EN code. Required power distribution and hot channel factors were computed by our past neutronics calculations using MCNP-5 code. Due to increasing temperature, modification on the continuous-energy cross sections in the MCNP libraries are carried out using NJOY code. Maximum and average fuel temperature, enthalpy, void fraction, coolant temperature and density, coolant mass flow rate and pressure drop are calculated using different models. Thermal-hydraulics calculations of the most rated channel (hottest sub-channel), which is determined based on coupled neutronics-thermal hydraulics calculations are investigated and results in temperature, enthalpy, critical heat flux and MDNBR of the hottest sub-channel are found. Finally, our results are compared with analytical approaches and the reactor FSAR.

Safety assessment of U–Mo fuel mini plates irradiated in HANARO reactor

Neutronic and thermal hydraulic characteristics of U–Mo fuel mini plates irradiated in the HANARO reactor were analyzed for the safety assessment of these plates. A total of eight fuel plates were double-stacked; each stack contained three 8.0gU/cc U–7Mo fuel plates and one 6.5gU/cc U–7Mo fuel plate. The neutronic and thermal hydraulic analyses were carried out using the MCNP code and TMAP code, respectively. The core status used in the study was the equilibrium core, and four Control Absorber Rod (CAR) locations were considered: 350mm, 450mm, 550mm, and 650mm away from the bottom of the core. For the fuels in the lower stack, the maximum heat flux was found at the CAR located at 450mm. For the fuels in the upper stack, the maximum heat flux was found at the CAR located at 650mm. The axial power distributions for the upper and lower stacks were selected on the basis of thermal margin analyses. A mock-up irradiation target assembly was designed and tested at the out-of-pile test facility to measure the flow rate through the irradiation site, given that the maximum flow rate through the irradiation site at the HANARO reactor is limited to 12.7kg/s. For conservative analyses, measurement and correlation uncertainties and engineering hot channel factors were considered. During normal operation, the minimum ONB temperature margins for the lower and upper stacks are 41.6°C and 31.8°C, respectively. This means that boiling does not occur. However, boiling occurs during the limiting accidents. Nevertheless, the fuel integrity is maintained since the minimum DNBR are 1.96 for the Reactivity Insertion Accident (RIA) and 2.10 for the Locked Rotor Accident (LRA). From the neutronic and thermal hydraulic analyses of the U–Mo fuel mini plates irradiated in the HANARO reactor, it is concluded that boiling does not occur during normal operation and that the fuel integrity is maintained during limiting accidents.

Thorium-based mixed oxide fuel in a pressurized water reactor: A beginning of life feasibility analysis with MCNP

Thorium is an asset the nuclear industry does not use, and plutonium is a liability that much of the world would like to be rid of. By incorporating a thorium–plutonium mixed oxide fuel (Th-MOX) into the fuel cycle, pressurized water reactors could provide a means for the United States to address both of these issues – but only if key reactor safety parameters are not affected.The feasibility of utilizing Th-MOX fuel in a pressurized water reactor is examined under steady-state, beginning of life conditions. With a three-dimensional MCNP model of a Westinghouse-type 17×17 PWR, many possibilities for replacing one-third of the UO2 assemblies with Th-MOX assemblies were considered. The excess reactivity, critical boron concentration, and centerline axial and radial flux profiles for several configurations and compositions of a one-third Th-MOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5wt% PuO2 was determined to be the best candidate for further analysis. Therefore, this configuration was compared to a 100% UO2 core using the following parameters: delayed neutron fraction (βeff), temperature coefficient, shutdown margin (SDM), and axial and radial nuclear hot channel factors (FZN and FRN).The one-third Th-MOX configuration showed an undesirable reduction in βeff from 0.00716±4.60E−07 for the 100% UO2 configuration to 0.00607±4.30E−07. The reduction in βeff would perhaps be ameliorated by the one-third Th-MOX configuration’s temperature coefficient of reactivity, which at −2.05±0.02pcm°F−1 is more favorable than the corresponding value of −1.42±0.02pcm°F−1 for the 100% UO2 configuration. The SDM of the one-third Th-MOX configuration is estimated to be 4079±7pcm, which is 28% lower than value of the 100% UO2 configuration. The FZN for the two cores were virtually identical. However, FRN for the one-third Th-MOX configuration (1.67±0.28) was 20% higher than the corresponding value for the 100% UO2 configuration (1.39±0.23).These preliminary results are encouraging. However, additional investigations are required to study the impact of thermal fluid feedback and the effect of burnup and poison buildup.

The max–min ant system and tabu search for pressurized water reactor loading pattern design

An automatic loading pattern (LP) design tool for a pressurized water reactor is developed. The design procedure consists of two steps: first, a LP is generated using max–min ant system (MMAS) and then tabu search (TS) is adopted to search the satisfactory LP. The MMAS is previously developed and the TS process is newly-developed. The heuristic rules are implemented to generate the candidate LP in TS process. The heuristic rules are comprised of two kinds of action, i.e., a single swap in the location of two fuel assemblies and rotation of fuel assembly. Since developed TS process is a local search algorithm, it is efficient for the minor change of LP. It means that a proper initial LP should be provided by the first step, i.e., by MMAS. The design requirements such as hot channel factor, the hot zero power moderator temperature coefficient, and cycle length are formulated in the objective function. The results show that the developed tool can obtain the satisfactory LP and dramatically reduce the computation time compared with previous tool using ant system alone.

Development of thermal hydraulic and margin analysis code for steady state forced and natural convective cooling of plate type fuel research reactors

Thermal hydraulic and Margin Analysis code for Plate type fuel assembly, TMAP-1.0, was developed to analyze the thermal hydraulic characteristics and thermal margins for steady state forced convective and natural circulation cooling in research reactors facilitated with plate type fuel assemblies. Forced convective cooling during normal operation at nominal power, and natural convective cooling during a training operation at low power, were considered for the model. Since the fuel cooling channels in plate type fuel assemblies are independent such that there is no cross flow from one channel to another, a one-dimensional analysis is suitable to predict the thermal hydraulic characteristics through the fuel cooling channels. The flow directions considered in the present model were downward during forced convective cooling and upward during natural convective cooling. The most unfavorable condition in the core was simulated by introducing engineering hot channel factors as well as radial and axial power peaking factors. Thermal margins such as Onset of Nucleate Boiling (ONB) temperature margin and Departure from Nucleate Boiling Ratio (DNBR) were estimated for reactor safety and design purposes. Furthermore, coolant, cladding outer surface, and fuel centerline temperatures were estimated along the axial direction with and without the engineering hot channel factors applied. For forced convective cooling, the prediction resulting from the present analyses were compared with experimental data taken by Sudo et al. (1984) and predictions by a subchannel analysis code, MATRA. For natural convective cooling, coolant velocities resulting from the balance of buoyancy and friction forces predicted by the present analyses were compared with predictions by RELAP5/MOD3. The comparisons showed good agreements for both forced and natural convective cooling.

Thermal hydraulic analyses of JRR-3: Code-to-code comparison of COOLOD-N2 and TMAP

A code-to-code comparison study was performed to investigate the thermal hydraulic characteristics of the JRR-3 reactor. The COOLOD-N2 and TMAP codes were used to analyze the JRR-3 reactor under downward forced convective and natural convective flows. The geometry and operational conditions of JRR-3 used in the present study were provided by the NEA (Nuclear Energy Agency) data bank. The total power peaking factors such as the radial and axial power peaking and the engineering hot channel factors were implemented to simulate hot and average channels. Thermal hydraulic characteristics, i.e., coolant and cladding wall surface temperatures, and the flow velocity and thermal margins, i.e., minimum ONB (Onset of Nucleate Boiling) temperature margin and minimum DNB (Departure from Nucleate Boiling) ratio were estimated at a nominal power of 20 MW for forced convection cooling and at 0.2 MW for natural convection cooling. As a result, a comparison study between the COOLOD-N2 and TMAP codes showed a comparability of the thermal hydraulic analysis results except for the ONB temperature margin and the coolant velocity in the hot channel during natural convection cooling. The differences resulted from the present analyses were discussed.

Automatic multi-cycle reload design of pressurized water reactor using particle swarm optimization algorithm and local search

An automatic multi-cycle core reload design tool, which searches the fresh fuel assembly composition, is developed using particle swarm optimization and local search. The local search uses heuristic rules to change the current search result a little so that the result can be improved. The composition of the fresh fuel assemblies should provide the required cycle energy and satisfy the constraints, such as the hot zero power moderator temperature coefficient and the hot channel factor. Instead of designing loading pattern for each FA composition during search process, two fixed loading patterns are used to calculate the core status and the better fitness function value is used in the search process. The fitness function contains terms which reflect the design objectives such as cycle energy, constraints, and fuel cost. The results show that the developed tool can achieve the desire objective.

Multi-layer fuel assembly design proposal for supercritical water cooled reactor

In the current SCWR fuel assembly design, there exist some challenges for thermal hydraulic and neutron-physical behavior. For the thermal SCWR design, it is of great importance to reduce the hot channel factor and the maximum cladding temperature as far as possible. The challenging task in the fast spectrum SCWR design is to achieve sufficiently large negative void reactivity coefficient and increase the conversion ratio. To satisfy the requirements and to solve challenges mentioned above, two structures of multi-layer fuel assembly for both thermal and fast SCWR core are proposed in this paper. For the thermal core fuel assembly, the main idea is to axially divide the active zone into several sub-layers, between which inactive layers are introduced, where fluid from the previous active layer mixes well with each other and enters the next active layer with a well homogenous distribution of fluid temperature. For the fast core fuel assembly, the main idea is to introduce the axial blanket (depleted UO2) regions between the divided MOX seed regions, to achieve a higher conversion ratio, lower temperature reactivity coefficient.Both thermal hydraulic and neutron-physical performance of the proposed multi-layer fuel assembly are investigated by a subchannel code coupled with 3-D neutronics analysis. The results obtained so far have shown that the multi-layer concept is feasible and promising.

Coupled analysis for new fuel design using UN and UC for SCWR

All-ceramic bi-material coated fuel pellet design is proposed and analyzed for thermal Super Critical Water Reactor (SCWR). Uranium mono nitride (UN) with Zirconium Carbide (ZrC) coating and Uranium mono carbide (UC) with Silicon Carbide (SiC) coating are analyzed. Carbide and nitride ceramic fuels offer the advantage of high thermal conductivity as compared to UO2. Use of coating can solve the problems of harder spectrum and reactive nature for UN and UC which arise when these fuels are used in light water thermal reactors. Larger heavy metal density of UN and UC can lead to equivalent or even larger values of heavy metal loading using coated pellet concept. Coating can give the extra advantage of working as yet another barrier against release of radioactive fission gases in accidental scenarios. Due to large coolant density variation along active height of fuel, coupled neutronics/thermal hydraulic analyses are performed using MCNP4c/SACoS coupled system. Design proposed by Shanghai Jiao Tong University (SJTU) for thermal SCWR is chosen to perform the calculations. Coupled analyses show that coated fuel pellets lead to more uniform radial power density which leads to smaller value of hot channel factor. A significant decrease in fuel centerline temperature is seen due to high thermal conductivity of carbide and nitride ceramic fuels. A staggering difference of more than 500 °C is seen between the maximum average centerline temperatures for UO2 and coated fuel pellets. Slightly smaller values of average clad surface temperature are obtained using coated fuel pellets as compared to UO2. A maximum coating temperature of less than 800 °C is observed. Effect of varying coating thickness is studied by performing analyses for different coating thickness values. Keeping the same fuel pin radius and pitch, increasing the thickness of coating materials can lead to increased value of thermal neutrons but will have a negative effect on the burnup due to smaller heavy metal loading. Reactivity coefficients, i.e. Doppler, coolant and moderator, show negative values. Considering the favorable neutronic and thermal hydraulic properties along with a potential increase in cycle length, all-ceramic bi-material coated fuel pellets can be a very good fuel option for SCWR.

Automatic pressurized water reactor loading pattern design using ant colony algorithms

An automatic core reload design tool was developed for a pressurized water reactor (PWR). A loading pattern (LP) was searched for using three different algorithms: the rank-based ant system (RAS), max–min ant system (MMAS), and Ant-Q which are variants of the ant colony algorithm. The fuel assemblies (FAs) were permuted in a one eighth core position and then the LP was copied to the other one eighth core with mirror symmetry, to form a quarter core LP, which was extended to a full core LP with rotational symmetry. Heuristic information was implemented to reduce search space and thus computation time. Safety requirements, such as the hot channel factor FΔH and moderator temperature coefficient (MTC), which must be satisfied, were formulated as penalty terms of the quality function. The search procedure contained two steps. The first step was to place the FA so that FΔH and MTC might be slightly violated, and the second step was to rotate the FA, which would improve the FΔH and MTC and the fuel cycle length. When the LP was designed, the SIMULATE-3 code calculated the FΔH, MTC, and cycle length, which were used to update the pheromone. The results demonstrated that the developed tool can obtain a LP which possesses the desired cycle length and also satisfies safety requirements.

Statistical calculation of hot channel factors

It is a conventional practice in the design of nuclear reactors to introduce hot channel factors to allow for spatial variations of power generation and flow distribution. Consequently, it is not enough to be able to calculate the nominal temperature distributions of fuel element, cladding, coolant, and central fuel. Indeed, one must be able to calculate the probability that the imposed temperature or heat flux limits in the entire core are not exceeded. In this paper, statistical methods are used to calculate hot channel factors for a particular heterogeneous geometry, Material Testing Reactor (MTR). Tehran Research Reactor (TRR) is used as a case for the present parametric study. The obtained results from this method are compared with the other statistical methods. It is shown that among the statistical methods available, the semi-statistical method is the most reliable one.

CORE AND SUB-CHANNEL EVALUATION OF A THERMAL SCWR

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermalhydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.