As pool scrubbing plays an important role in fission product (FP) decontamination, a reliable model is needed. Despite the needs, mechanism of FP transfer from air-water from the swarm is not explained exactly which means that the evaluation of physical model used in pre-existing model couldn’t be done enough. Existing model for pool scrubbing is predicted in the MELCOR code. Inside the code, a simple model of bubbly jet divided in 3 regions is used: 1) Globule region where the gas including FP enter and collapse, 2) Swarm rise region where the bubble rises up stationary after the collapse is done and, 3) Entrainment region where the bubble pop out to the atmosphere. In each region, the decontamination factor (DF), the particle density ratio before and after each region, is calculated. On these region, flow and physical force inside the gas phase is predicted to be one of the driving force which cause the FP transfer. Therefore, our study aims at the particle behavior on the gas phase. As to understand the physical phenomenon individually, the study focuses on flow behavior and inner flow of a single rising bubble. As an approach, comparison of bubble containing aerosol and no aerosol has been done for each parameter of size, aspect ratio, velocity. Compared with existing equations, the rising speed of clean bubble condition and aspect ratio of CSI condition agreed well to the MELCOR code [1] . On the other hand, many difference were also measured in other condition. Application of parameters obtained from experiment were done against the MELCOR model. Calculation of velocity inside the oil droplet using the experimental parameters obtained from visualization measurement was done. The local gravitational sedimentation and centrifugal velocity took a higher value in clean bubble and OX50 condition compared to CSI condition. On the other hand, Brownian diffusion velocity had an opposite trend. PIV measurement were performed by a silicone oil to visualize the inner flow clearly and compared with the calculation. Seen from the results, the local diffusion velocity took a lower value compared to the calculation using the MELCOR model.

The subject of heat transfer behaviors of steam generator is of great interest for better heat transfer efficiency and safety in industry. On most previous study, the models of SG based on RELAP5 used heat structures as the power generation, which is different from heating by primary side. This paper investigates the distinctions of different heat transfer methods on the thermal-hydraulic parameters by RELAP5. Two different heat transfers are selected, case 1 is uniformly heated by heat structure and case 2 is heated by the primary side of SG. The results show that the onset position of critical heat flux for case 2 is ahead of that for case 1. Besides, the temperature of tube wall corresponding to position of critical heat flux increases abruptly in case 1, and the wall temperature in superheated region in case 1 increases faster than that in case 2 when the water becomes fully superheated vapor. Then instability distinctions for two heat transfer cases are studied in this investigation. It is detected the fluid flow heated by second method is more stable than the first one, and the system has a higher ultimate thermal power when density wave oscillation occurs for case 2.

In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the temperature distribution in the fuel pin bundle is of central interest. In particular, the use of lead or lead-bismuth eutectic (LBE) as coolant for the new generation fast reactors is one of the most promising choices. Due to the high density and high conductivity of lead or LBE, a detailed analysis of the thermo-fluid dynamic behavior of the heavy liquid metal (HLM) inside the sub-channels of a fuel rod bundle is necessary in order to support the Front-End Engineering Design (FEED) of GEN. IV/ADS prototypes and demonstrators. In this frame, the synergy between numerical analysis by CFD and data coming from large experimental facilities seems to be crucial to assess the feasibility of the components. At ENEA-Brasimone R.C., large experimental facilities exist to study HLM free, forced and mixed convection in loops and pools: e.g. NACIE-UP is a large scale LBE loop for mixed convection experiments. The MYRRHA-19 like Fuel Pin Bundle Simulator installed in the NACIE-UP facility allows to make non-uniform and dissymmetric tests with only a few pins heated. This technical feature of the FPS is very interesting for CFD validation and this kind of data tests in HLM fuel bundles are not so common in the literature. In the present paper, a post-test validation is made by a detailed CFD model of the test section. Experimental data, statistically treated by the error propagation theory, are briefly presented and a preliminary comparison with CFD results using different models/turbulent Prandtl numbers are shown. Three monitored section at different levels are compared both for wall and bulk temperatures. This post-test comparison with this experimental configuration is unique and represents a further step towards the validation of the CFD models and methods in fuel bundle geometries cooled by HLM.

Two-fluid model is a common method to simulate the subcooled flow boiling heat transfer, in which the wall boiling model is mainly used for the partition of wall heat flux and the mass transfer between two phases on the wall. The model determines the amount of vapor phase and predicts the cross-sectional void fraction in the channel, nucleate site density and bubble departure diameter play an important role in the accurate prediction of wall boiling model. Eulerian two-fluid model coupled with Rensselaer Polytechnic Institute (RPI) wall boiling model is employed to simulate the heat transfer characteristics and boiling phenomena in vertical narrow rectangular channels by using FLUENT code. Based on the experimental data of subcooled boiling in vertical narrow rectangular channel, different combinations of nucleate site density and bubble departure diameter correlations are used to calculate under different conditions of heat flux and inlet subcooling. Comparing the calculated heat transfer coefficients along the vertical height with experimental results, it can be found that these two parameters have a significant effect on the subcooled boiling heat transfer in narrow rectangular channels. Different parameter combinations lead to differences in wall heat flux distribution, different heat flux and inlet subcooling also have different effects on these models, which eventually lead to different evaporative heat flux, thus affecting the prediction of void fraction.

Computational Fluid Dynamics (CFD) methods have been developed into effective fluid simulation means to be used on the hydraulic design in the field of nuclear reactor. However, it is difficult to generate suitable mesh and select turbulence model for simulation because of the complex geometry structure and flow behavior in the pressure vessel. Based on CFX software, a hydraulic computation model of typical pressurized water reactor is established and the flow distribution at core inlet is analyzed. The simplified geometry model consists cold legs, downcomer, lower plenum, secondary support component, core support plate, and lower core plate. The computation model is divided into three parts for mesh generation, including the part of inlet and downcomer, the part of lower plenum and core inlet section, and the part of core. In order to reach the independence of grid several methods of mesh generation which contains different mesh density at local key parties are investigated to screen out the suitable mesh scheme. The k-ε, k-ω, and SST k-ω turbulence model are respectively used for simulation and the sensitivity of turbulence model at different locations of flow field is analyzed. The results show that the mass flow rate of the near wall flow field, computed by using k-ω turbulence model, is consistent with SST k-ω model, while the mass flow rate of central flow field computed by using k-ε turbulence agrees with SST k-ω model. The result computed by using k-ε turbulence model shows relatively uniform flow distribution at core inlet, which is more consistent with the measured data with the average difference of 3.1%. By using the k-ε model, the probability distribution of the difference between the calculated results and the experimental values follows the law of Normal distribution. The final coolant flow distribution at each orifice is evaluated, and the maximum normalized flow flux is found at center orifice while the flow rate at the edge of core is relatively lower.

Monte Carlo (MC) burnup calculation method, implemented through coupling neutron transport and point depletion solvers, is widely used in design and analysis of nuclear reactor. Burnup calculation is generally solved by dividing reactor lifetime into steps and modeling geometry into numbers of burnup areas where neutron flux and one group effective cross sections are treated as constant during each burnup step. Such constant approximation for neutron flux and effective cross section will lead to obvious error unless using fairly short step. To yield accuracy and efficiency improvement, coupling schemes have been researched in series of MC codes. In this study, four coupling schemes, beginning of step approximation, predictor-corrector methods by correcting nuclide density and flux-cross section as well as high order predictor-corrector with sub-step method were researched and implemented in RMC. Verification and comparison were performed by adopting assembly problem from VERA international benchmark. Results illustrate that high order coupled with sub-step method is with notable accuracy compared to beginning of step approximation and traditional predictor-corrector, especially for calculation in which step length is fairly long.

The concept of multi-dimensional heterogeneous resonance integral tables is proposed. The new type of resonance integral is designed for different fuel pins appearing in one lattice with two extra dimension of optical radius and number density ratio in the fuel. Numerical results show that this treatment improves the accuracy of embedded self-shielding method on irregular lattices.

Monolithic fuel is a fuel form that is considered for the conversion of high performance research reactors. In order to qualify this new fuel system, the fuel plates should meet the safety standards and perform well in reactor. The fuel system must maintain its mechanical integrity, sustain a geometric stability and should have stable and predictable irradiation behavior. The requirement to maintain mechanical integrity under normal operating conditions is primarily demonstrated by a successful testing of fuel plates up to the limiting conditions defined by the fuel performance envelope, including an adequate margin. Although large number of plates have been tested with satisfactory thermo-mechanical performance, post-irradiation examination of plates from previous RERTR-12 experiments have revealed that pillowing occurred in several plates, rendering performance of these plates unacceptable. To address such failures, efforts are underway to define the mechanisms responsible for the in-reactor pillowing, and suggest improvements to the fuel plate design and operational envelope. For this purpose, selected plates from previous experiments were simulated to understand the thermo-mechanical response of the plates to the fission density and thermally induced stresses. Simulation results were then comparatively evaluated with post-irradiation examinations of selected plates. The simulation results and experimental observations established a possible correlation between failure by plate pillowing, high porosity and a presence of tensile stress state. The study has implied that porosity leading to degradation of material properties, accompanied by a sufficiently large tensile stress state can lead to a pillowing-type failure at reactor shutdown. This paper presents these findings, discusses such failure modes, and the influence of fuel burn-up and power on the magnitude of the shutdown-induced tensile stresses.

The isotope Xe-135 has a large thermal neutron absorption cross section and is considered to be the most important fission product. A very small amount of such neutron poison may significantly affect the reactor reactivity since they will absorb the neutrons from chain reaction, therefore monitoring their atomic density variation during reactor operation is extremely important. In a molten reactor system, Xe-135 is entrained in the liquid fuel and continuously circulates through the core where the neutron irradiation functions and the external core where only nuclei decay occurs, at the same time, an off-gas removal system operates for online removing Xe-135 through helium bubbling. These unique features of MSR complicate the Xe-135 dynamic behaviors, and the calculation method applied in the solid fuel reactor system is not suitable. From this point, we firstly analytically deduce the nuclei evolution law of Xe-135 in the flowing salt with an off-gas removal system functioning. A study of Xe-135 dynamic behavior with the core power change, shutdown, helium bubbling failure and startup then is conducted, and several valuable conclusions are obtained for MSR design.

As one of the significant equipment in CEFR operation system, cold trap plays an important role in purifying the coolant sodium and ensuring the safe and steady operation of the reactor. After several years’ of operation, the cold trap of CEFR can’t work very well as it has been designed previously, which results in the low impurity aggradation efficiency, being easily blocked and frequent replacement. Analyzing the structure of the current cold trap, there are two possible reasons causing the problems mentioned above. The first one is that the sodium flow path is improper in the cold trap resulting in the blockage in specific area. The second one is that the density and layout of wire mesh didn’t take the zoning scheme, which would make the precipitation of impurities become easier in exterior area and the flow path is thus blocked. Based on the two reasons, a modified scheme is proposed from the view of structure and hydraulics to improve the impurities precipitation mode so that the lifetime of the cold trap can be prolonged. The modified scheme is to block about 1/3 orifices on interior cylinder wall in wire mesh zone and to open two rows of diversion holes on the upper support plate of wire mesh so that the velocity distribution can be improved. Meanwhile, in order to make impurities precipitate from the bottom of the cold trap, the wire mesh is arranged to be in upper, middle and lower areas using different wire mesh density. By 3-D modeling and numerical analysis and then comparing the results to that of current cold trap of CEFR, it can be concluded that the velocity distribution in modified scheme is more reasonable and the new layout of the wire mesh can meet the impurities precipitation requirements better. Further comparative analysis with the results under the actual operational condition will be carried on after the completion of the manufacture for modified cold trap.

In the concrete cask, the canister is sealed with lids by welding, and has high sealing performance. But considering long-term storage, there is a concern about loss of the sealing performance due to stress corrosion cracking (SCC). In the concrete cask, unlike the metal cask, it is not mandatory to constantly monitor helium pressure between the lids. However, it is useful from the viewpoint of improving safety during the long-term storage to install a helium leak detector in the canister inside the concrete cask. Currently, we are developing the leak detector utilizing the phenomenon that the surface temperature of the canister changes when helium leaks out of the canister. As part of developing the leak detector of the canister, leak tests were performed using a small canister model as a pressurized vessel and a 1/4.5 scale cask model of the actual cask including the canister. This leak detector utilized the phenomenon that canister bottom temperature (TB) increases and canister lid temperature (TT) decreases when the internal pressure of the canister decreases. In computational fluid dynamics (CFD) calculation, focused on this phenomenon, the influence of the internal pressure and physical properties of internal gas in the canister were examined by calculating conditions of three kinds of pressure and two types of gas (air and helium). The main purpose of the CFD calculation was to confirm the results of the experiment, and we grasped the phenomenon occurring in the canister and elucidated its mechanism. For the CFD calculation, a commercial CFD software, STAR-CCM+ ® (ver.12.06.010) by Siemens PLM Software Company, was used. A CAD file used for the calculation simulated also the shape inside the canister (e.g. basket, fuel rods). A polyhedral mesh was used for a calculation mesh. In the small canister model, a mesh of its ambient air was not generated, and heat transfer between the canister surface and the ambient air was calculated from a heat transfer correlation equation. On the other hand, in the 1 / 4.5 scale cask model, the mesh of its ambient air was generated, so that the heat transfer on the surface of the canister was calculated according to the actual heat transfer phenomenon. The internal gas and the ambient air of the canister were ideal gas, and buoyancy due to density change was taken into consideration. A realizable k-epsilon model was used for a turbulence model, and a DO model was used for a radiation model.

Series of experiments are conducted in a single microchannel, where subcooled water flows upward inside a transparent and vertical microchannel. The cross section of the channel is rectangle with the hydraulic diameter of 2.8mm and the aspect ratio of 20. The working fluid is 3–15K subcooled and surface heat flux on the channel is between 0–3.64 kW/m 2 , among which two-phase instability at low vapor quantity may occur. By using a novel transparent heating technique and a high-speed camera, visualization results are obtained. The parameters are acquired with a National Instruments Data Acquisition card. In the experiments, long-period oscillation and short-period oscillation are observed as the primary types of instability in a microchannel. Instability characteristics represented from signals correspond well with the flow pattern. Moreover, effects of several parameters are investigated. The results indicate that the oscillating period generally increases with the heat flux density and decreases with inlet subcooling, while the effects of inlet resistance are more complex.

An experimental investigation was conducted in a natural circulation (NC) loop to study the characteristics of two-phase flow instability under low pressure condition. A 3 × 3 rod bundle channel was used as the test section. The effects of heating power, inlet subcooling degree and system pressure on the two-phase NC flow instability types and stable boundaries were studied. The experimental results show that three typical flow conditions can occur in rod bundle channel under NC condition, which are single-phase NC flow, subcooled boiling NC flow oscillation and density wave oscillations (DWO). The oscillation amplitude and period of DWO can be enlarged by increasing the heat flux. Increasing the inlet subcooling degree can increase the marginal heating power of flow instability in NC system. The occurrence of DWO can be suppressed by increasing the system pressure. The flow instability boundary presented by the subcooling number and phase change number was also obtained in present work.

In a de-nitration conversion processes of the nuclear fuel cycle, mixed oxides (MOX) are produced from the reprocessing solution (plutonium and uranium mixed nitrate solution) of used nuclear fuel. The microwave heating (MH) method has various advantages as one of de-nitration conversion techniques, i.e., this can complete rapidly processes, reduce the waste liquid, and operated remotely and easily by the waveguide. Fine crystal powders are thereby generated, and the manufacture of high-density and high-quality pellets is significant advantage. The MH method is accompanied with transient boiling phenomena such as overflow and flashing. From the viewpoint of enhancing mass productivity and cost efficiency, in the future, scaling up for the size of de-nitration vessel and shortening processing time are desired. In addition, the safe design of device and the appropriate conditions in microwave irradiation process are required. Hence the extensive understanding of transient boiling phenomena induced by microwave heating is important, and the detailed mechanism of flushing and overflow should be clarified. The aim of this study is to clarify the transient boiling phenomena and detailed mechanism of flashing and overflow. Flashing are affected by physical factors such as solution properties, vessel characteristics, and input energy. Distilled water and different dielectric constant solution as working fluid are used. Dielectric constant was adjusted by the concentration of potassium chloride aqueous solution. A cylindrical vessel kept enough clean by ultrasonic washing machine before experiments was used. The influence of vessel diameter and initial water depth on flashing is examined. As input energy, amount of power supplied and position of the object to be heated in an oven are varied. A high-speed video camera was installed to observe boiling phenomena, and thermography and fiber optic thermometer were to measure the temperature on vessel walls and in a solution, respectively. As a result, the higher the dielectric constant solution, the less risk of flashing phenomena. In no flashing case, the surface temperature became over 100 °C and the precipitation of potassium chloride was existed. When the high dielectric constant solution was heated by microwave irradiation, the potassium chloride which was precipitated on the inner surface of the vessel was heated to high temperature. On the other hand the temperature of water was lower than that of the deposition of the potassium chloride. As described above, the influence of the high dielectric constant solution on the flashing phenomenon by microwave heating was evaluated. And the mechanism of flashing phenomena was assumed by detailed observation of bubble nucleation. Eventually the mechanism of flashing phenomena was examined.

In vessel retention (IVR) of molten core debris via water cooling at the external surface of the reactor vessel is an important severe accident management feature of advanced passive plants. During postulated severe accidents, the heat generated due to the molten debris relocation to the lower reactor pressure vessel head needs to be removed continuously to prevent vessel failure. Besides the local critical heat flux (CHF) of outer wall surface which is the first importance, a stable feature of natural circulation flow and an effective natural circulation capability within the external reactor vessel cooling (ERVC) channel tend to be rather crucial for the success of IVR. Under this circumstance, a full-height ERVC test infrastructure for large advanced pressurized water reactor (PWR) IVR strategy engineering validation, namely reactor pressure vessel external cooling II test facility (REPEC-II), has been designed and constructed in Shanghai Jiao Tong University (SJTU). And therefore, a brief introduction to the SJTU REPEC II facility as well as the experimental progress to date, is hereby given in the paper. During test campaign on the REPEC II facility, the one-dimensional natural circulation boiling flow characteristics during IVR-ERVC severe accident mitigation are investigated, with the experimental observation and measurement on natural circulation flow characteristics within the REPEC II test facility. Based on the abundant results acquired in the test campaign, it is attempted, in this paper, to summarize and evaluate the ERVC performances and trends under various practical engineered conditions. The main evaluation results includes: influence on ERVC flow characteristics of various non-uniform heat load distribution cooling limits, the observed sinusoidal oscillation is suggested to be flashing-induced density wave oscillations and the oscillation period correlated well with the passing time of single-phase liquid in the riser. It is expected that these conclusions may help designers to have a reliable estimate of the impact of some related engineered factors on real IVR-ERVC performance.

Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped. This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5 × 5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40°C, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.

For the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel or steel is solidified into debris particles which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes debris relocated and flattened, which called debris relocation. The thickness of debris bed has great influence to the cooling ability of fuel debris in low plenum. To ensure the effective implementation of the in-vessel retention (IVR), it’s very necessary to evaluate the transient changes of shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a debris relocation model based on COMMEN code was developed in this paper. The debris relocation model was established based on the extrapolation of the shear strength mechanism, which was originally proposed and widely applied in soil mechanics filed. Shear strength is a function of the particles’ density and position. Debris bed is fluidized only when the shear stress in particle unit is larger than shear strength of debris particles. By integrating the debris relocation model into the COMMEN code, the transition process of the bed in depressurization experiments was simulated and compared against the experimental results. Good agreement shows that the debris relocation model presented in this paper can reasonably simulate the relocation behavior.

V&V constitutes a powerful framework to demonstrate the capability of computational tools in several technological areas. Passing V&V requirements is a needed step before applications. Let’s focus hereafter to the area of (transient) Nuclear Thermal-hydraulic (NTH) and let’s identify V1 and V2 as acronyms for Verification and Validation, respectively. Now, V1 is performed within NTH according to the best available techniques and may not suffer of important deficiencies if compared with other technological areas. This is not the case of V2. Three inherent limitations shall be mentioned in the case of Validation in NTH: 1. Validation implies comparison with experimental data: available experimental data cover a (very) small fraction of the parameter range space expected in applications of the codes; this can be easily seen if one considers data in large diameter pipe, high velocity and high pressure or high power and power density. Noticeably, the scaling issue must be addressed in the framework of V2 which may result in controversial findings. 2. Water is at the center of the attention: the physical properties of water are known to a reasonable extent as well as large variations in values of quantities like density or various derivatives are expected within the range of variation of pressure inside application fields. Although not needed for current validation purposes (e.g. validation ranges may not include a situation of critical pressure and large heat flux) physically inconsistent values predicted by empirical correlations outside validation ranges, shall not be tolerated. 3. Occurrence of complex situations like transition from two-phase critical flow to ‘Bernoulli-flow’ (e.g. towards the end of blow-down) and from film boiling to nucleate boiling, possibly crossing the minimum film boiling temperature (e.g. during reflood). Therefore, whatever can be mentioned as classical V2 is not or cannot be performed in NTH. So, the idea of the present paper is to add a component to the V&V. This component, or step in the process, is called ‘Consistency with Reality’, or with the expected phenomenological evidence. The new component may need to be characterized in some cases and is indicated by the letter ‘C’. Then, the V&V becomes V&V&C. The purpose of the paper is to clarify the motivations at the bases of the V&V&C.

Passive residual heat removal system (PRHRS) is of great significance for reactor shutdown safety. The PRHRS of a small modular reactor, such as the integral pressurized water reactor (iPWR) and the modular high temperature gas-cooled reactor (MHTRG), is composed of the primary loop (PL), intermediate loop (IL) and air-cooling loop (AL). The AL is a density-difference-driven natural circulation caused by the difference of air temperature at the inlet and outlet of the air-cooling tower. Thus, it is possible to adopt the air flow in AL to generate electricity for post-shutdown reactor monitoring. In this paper, a novel residual heat electricity generation system (RHEGS), which is composed of the PRHRS and a vertical wind generator installed in the air-cooling tower, is proposed for the power supply of post-shutdown monitoring instruments. To verify the feasibility of practical implementation, the dynamical model of this newly designed RHEGS including the dynamics of PRHRS, windmill, rotor as well as doubly-fed induction generator (DFIG) are all given. Then, both steady-state and transient verification for the RHEGS of a nuclear heating reactor NHR200-II plant with a rated thermal power of 200 MW th is carried out, which shows that the output active power of RHEGS can be 20∼30kW which is about 1% the residual heat flux and can fully meet the power requirements of post-shutdown monitoring instruments.

The Super Fast Breeder Reactor (Super FBR) utilizes supercritical light water as coolant, which changes from liquidlike high density state to gas-like low density state continuously in the core without phase change. In the preceding study (Noda et al., 2017), new concept of axially heterogeneous core with multi-axial fuel shuffling was proposed. The core consisted of two layers of mixed oxide (MOX) fuel and two layers of blanket fuel with depleted uranium (DU), which were arranged alternatively in the axial direction. The study showed that, with independent fuel shuffling in the upper part and lower part of the core, breeding performance could be improved by increasing the upper blanket fuel batch number while keeping the fuel batch number of the rest of the core unchanged, because of increased neutron flux in the upper blanket. However, the study did not consider influence of different coolant density histories in the different axial level of the core on the core neutronics. Hence, this study aims to reveal influence of the different coolant density histories through design and analyses of the multi-axial fuel shuffling core with two MOX layers and three blanket layers. The three levels correspond to the coolant density below, around, and above the pseudo-critical temperature. The neutronics calculations are carried out with SRAC 2006 code and JENDL-3.3 nuclear data library. Unit cell burnup calculations based on collision probability method are carried out for 5 different coolant density histories to consider influence of different neutron spectrum on breeding performance of the core. Influence of instantaneous coolant density changes on the core neutronics are considered by coupling core burnup calculations with thermal-hydraulics calculations based on single channel model. Influence of independent fuel shuffling of the upper blanket on the core neutronics (breeding performance and void reactivity characteristics) is investigated, followed by a similar investigation on the lower blanket. The differences between the two schemes are investigated since coolant density histories are greatly different between the upper blanket and the lower blanket.