The best estimate plus uncertainty (BEPU) method is recommended by IAEA for nuclear safety analysis. Most of the existing BEPU methodologies rely on the uncertainty propagation of input parameters, while uncertainties of the constitutive models in best estimate codes tend not to be valued or even neglected. A structural method is proposed in this paper to quantify the uncertainties of the constitutive models. Different constitutive models will be classified according to the characteristics and corresponding method could be utilized for each model based on the method. Specific uncertainty quantification (UQ) methods adopted in this paper include the non-parametric curve estimation method, inverse method and design of experiment (DOE) method, and a model selection technique is adopted to opt the optimal model among all alternative models. The structural method is applied to the uncertainty evaluation of LOFT LP-02-6 experiment. Important models are identified, uncertainties of these models are quantified and propagated to the peak cladding temperature (PCT) through code calculations. Uncertainty of the PCT is quantified and the result shows that the calculated values could well envelop the experimental value.

Interfacial drag force, which indicates the momentum transfer between liquid phase and vapor phase, is a key constitutive equation in the two-fluid model. Based on the “drift-velocity approach” (utilized in RELAP5/MOD3) and the “drag coefficient approach” (utilized in RELAP5/MOD2 and CTF), three improvements are proposed, which are: 1) improved drag coefficient closure, 2) improved drag coefficient formulation approach, 3) improved bubble radius closure approach. The comparison among the two original approaches and the three improved approaches has been made with the ORNL experiment data in high pressure-low flow condition and the results have been discussed. Results indicates: 1) the EPRI correlation predicts the void fraction worse than the drag coefficient approaches; 2) the drag coefficient correlations in CTF predicts the void fraction better than improved drag coefficient formulation approach, which are the original equations in Ishii’s model; 3) the improved drag coefficient formulation approach predicts similarly with the original RELAP5/MOD2 correlations, though it gets rid of the dependence on interfacial area concentration; 4) improved bubble radius closure approach over-predicts the void fraction, however more experiment tests should be calculated before a conclusion is drawn.

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.

ATHAS is the subchannel analysis code, independently developed by Nuclear Safety and Operation Laboratory (NUSOL) of Xi'an Jiao Tong University (XJTU), the original version of the ATHAS is only suitable for pressurized water reactor. This paper is aimed at the characteristics of sodium cooled fast reactor core mentioned above, developing the analysis program AYHAS-LMR, which is applicable for LMFBR. In the design of sodium cold fast reactor, flow blockage must be considered to meet the design criteria. In view of this, a new model is proposed in this paper and added in ATHAS-LMR, considering the flow blockage which changes the flow area and the wetted perimeter. This model is based on the original grid model, we use the grid to simulate the presence of block, increase the resistance of the blockage area artificially. Then, ATHAS-LMR is able to simulate the flow blockage in LMFBR. In this paper, simulations of FFM5B blockage test were done for the high flow case with ATHAS-LMR, and we compare the simulation results with COMMIX-1. The result indicates that ATHAS-LMR is suitable for the analysis of the flow blockage.

Nuclear reactor severe accidents can lead to the release of a large amount of radioactive material and cause immense disaster to the environment. Since the Fukushima nuclear accident in Japan, the severe accident research has drawn worldwide attention. Based on the one-dimensional heat conduction model, a DEBRIS-HT program for analyzing the heat transfer characteristics of a debris bed after a severe accident of a sodium-cooled fast reactor was developed. The basic idea of the DEBRIS-HT program is to simplify the complex energy transfer process in the debris bed to a simple one-dimensional heat transfer problem by solving the equivalent thermal conductivity in different situations. In this paper, the DEBRIS-HT program code is prepared by using the existing model and compared with the experimental results. The results show that the DEBRIS-HT program can correctly predict the heat transfer process in the fragment bed. In addition, the heat transfer characteristics analysis program is also used to model the core catcher of the China fast reactor. Firstly, the dryout heat flux when all of molten core dropped on the core catcher was calculated, which was compared with the result of Lipinski’s zero dimensional model, and the error between two values is only 11.2%. Then, the temperature distribution was calculated with the heat power of 15MW.

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.

The Ransom water faucet problem has become one of the most important benchmark tests to study two-fluid/phase flow because it exists an analytical discontinuous solution. The water faucet problem is a gravity-driven wave problem and its analytical solution is derived from the liquid free fall motion. In this paper, the opposite gravity-driven wave problem named the reversed water faucet problem where the liquid admits a rising motion with reduced speed driven by gravity is studied. With assumptions of decoupled phasic pressures, approximate incompressible flow, no liquid pressure gradient, no phase change, no wall and interfacial drag, the analytical solutions of the gas volume fraction and liquid velocity distribution are derived. From the gas volume fraction analytical solution, there is a moving discontinuity which is a very important point for testing accuracy of the numerical scheme and its stability near discontinuities. Two-fluid seven-equation two-pressure model is of particular interest due to the nature of inherent well-posed advantage in all situations. What’s more, high-order accuracy schemes have attracted great increasing attention to overcome the challenge of serious numerical diffusion from 1st-order scheme for accurately simulating many nuclear thermal-hydraulics applications such as long term transient natural circulation problems. In this paper, the solution algorithms with high-order accuracy in space and time are developed for this well-posed two-fluid model and its robustness and accuracy are verified and assessed against the derived analytical solutions. The numerical results show that high-order schemes could prevent excessive numerical diffusion and are more accurate than first-order time and space schemes; the space high-order scheme could give more accurate numerical results than the time high-order scheme for discontinuous solutions.

According to the calculated axial and radial power distribution of reactor core, a numerical analysis of thermal-hydraulic behavior of coolant in supercritical water-cooled reactor of CGNPC (CGN-SCWR) was performed by using sub-channel analysis software ATHAS. The coolant flow in CGN-SCWR reactor core were in two path mode. In order to ensure the maximum cladding temperature (MCT for short) reached the requirements throughout the entire life cycle, the sub-channel analysis of the MCT in the BOC, MOC and EOC had been carried out. The temperature and density properties of coolant and moderator in the fuel assembly along the axial and radial direction were investigated. The MCT of the fuel assembly was obtained, and the selected typical sub-channels were analyzed in detail. The results show that the MCT, coolant outlet temperature and other parameters in CGN-SCWR reactor core meet the thermal-hydraulic design requirements through the whole life time.

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K) [1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.

The CHF in PWR fuel assemblies is usually predicted by the local flow correlation approach based on subchannel analysis while the effects of spacer grids, cold walls, non-uniform heat flux, etc are investigated. By using the subchannel code ATHAS to calculate each set of bundle CHF data, the local thermal-hydraulic parameters at DNB occurrence point were obtained. In present study, the minimum DNBR point method was applied to develop a new CHF correlation for PWR fuel assemblies. The so-called “three-step method” and “magnitude analysis method” were used to determine the shape and the expression of each item, respectively and the least square method was applied to determine the coefficients of the correlation. Based on the large database of CHF tests, the CHF correlation named ACC correlation has been developed to calculate the risk of DNB. The analysis and assessment results indicate that the ACC correlation can fit the experimental data well with high prediction accuracy and correct parametric trends. Coupled with subchannel code ATHAS, this correlation can simulate the thermal-hydraulics performances of PWR fuel assemblies exactly.

The spent fuel pool (SFP) is mainly used for cooling spent fuel assemblies (SFAs) discharged from the reactor core. Besides, it can also shield the radiation. The decay heat can be removed through normal operation cooling system, otherwise it can only rely on the natural circulation in the pool when the coolant pump loses power or the heat exchanger fails. Thus the pool water temperature will continue to rise until it begins to boil. During this period, if no active cooling measures are triggered, the water level will gradually decrease as water boiling. Once the water level drops to the top of the fuel assemblies, the fuels begin to be exposed in the environment. In this paper, the CPR1000 spent fuel pool was chosen as the analysis object and the best estimate system thermal hydraulic code RELAP5 was employed to investigate the process in spent fuel pool in case of loss of heat sink. The results of calculations show that when losing two sets of cooling line, the increase in water temperature in the pool from 55 °C up to 100 °C takes approximately 9.1 h, the evaporation of water volume above the SFAs takes approximately 75.4 additional hours; while in case of losing one set of cooling line, the water temperature of the pool surface reaches 76.6 °C, at which the pool water would not going to boil under the atmospheric environment condition. The results of performed analysis are useful for safety analysis and storage of the SFAs, and can be used to provide a reference for spent fuel pool engineering design.

This paper describes the preliminary safety analysis of a thermal-spectrum SCWR concept (CSR1000), which was proposed by Nuclear Power Institute of China (NPIC). The passive safety system and the design of the two-pass core concept characterize the safety performance of CSR1000. With code SCTRAN (a one-dimensional safety analysis code for SCWRs), loss of coolant flow accidents (LOFA) and loss of coolant accident (LOCA) as well as some other typical transients and accidents were analysed. The maximum cladding surface temperature (MCST) was regarded as an important criterion. The sensitivity analyses of some crucial parameters are helpful for the safety evaluation. Thus some parameters about the safety system and the actuation conditions, such as the delay time of the ADS actuation, the break area in LOCA analysis, were also involved in this paper. The analyses have shown that the proposed passive safety system is capable to mitigate the consequence of the selected abnormalities. The results will be a useful reference for the future development of CSR1000.

The radial power distribution is uneven in the Pressure Tube type SCWR bundle because of the higher fuel enrichment and the separation of light water coolant and heavy water moderator. The uneven profile will usually result in high cladding temperature. In order to optimize bundle design, the analysis of coupled neutronics/thermal-hydraulics was carried out. The neutronics code WIMS-AECL and the subchannel code ATHAS are selected to conduct the coupled neutronics/thermal-hydraulics analysis for the PT-SCWR 54-elements bundle design. The analysis of the reference case showed that the maximum cladding surface temperature of the bundles at both Beginning of the Cycle (BOC) and End of the Cycle (EOC) exceed the design limits. The optimization was carried out through (1) adopting staggered radial uranium enrichment profile and (2) optimizing the pitch-circle diameters of element rings. The results showed that the optimized bundle can achieve lower maximum cladding surface temperature, which is well below the limiting criteria, and meet the criteria of neutronics demands.

Safety analysis is one of the chief difficulties during the research and design of SCWRs. Currently, the development of the SCWR safety analysis code is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs, has been developed based on RETRAN-02, the best estimate code used for safety analysis of Light Water Reactors. The ability of SCTRAN to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with the APROS and RELAP5-3D codes, respectively. The results show that the SCTRAN code developed in this study is capable of performing safety analysis for SCWRs, and the results are reliable. SCTRAN developed in this study is of great value and engineering significance, and could be utilized in the Chinese SCWR research and development.

Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assembly with wire wrap can achieve a more uniform coolant temperature profile than the grid spaced assembly, which will result in a lower peak cladding temperature; (2) The pressure drop in a wire wrapped assembly is less than that in a grid spaced assembly, which can reduce the operating power of pump. (3) The wire wrap pitch has significant effect on the flow in the assembly. Smaller H wire /D rod will result in stronger cross flow and a more uniform coolant temperature profile, and also a higher pressure drop.

Supercritical Water Reactors (SCWRs) are essentially water reactors operating at pressure and temperature above critical point. The heat transfer coefficient is relative low when the bulk temperature is above the pseudo-critical point due to the properties of vapor-like fluid. To obtain better heat transfer characteristics, increasing the fluctuation using obstacles is the conventional method. Heat transfer characteristic in vertical tube with different obstacles is numerically investigated under supercritical condition. Numerical simulation is carried out with commercial CFD code Fluent 6.1 and adaptive grid. The results show that The RNG k-ε model with enhanced wall treatment can obtain a reliable result; the blockage ratio and the local temperature have large influence on the heat transfer enhancement. The influence region and decay trend of obstacles are also studied and compared with existing correlations.

CANDU supercritical water reactor (SCWR) offers advantages in the areas of sustainability, economics, safety and reliability and proliferation resistance. However, there is still a big deficiency in understanding and prediction of heat transfer behaviour in supercritical fluids. In this paper, heat transfer is numerically investigated on supercritical water for three-dimensional horizontal flows. Three ε-type turbulence models are tested and the numerical results are compared with experimental data. Based on the result, the standard k-ε turbulence model with enhanced wall treatment is recommended. The effect of the buoyancy and heat transfer deterioration is also analyzed, and the criteria for onset of buoyancy effects is evaluated. The quantity Gr/Re 2.7 recommended by Jackson et al. (1975) gives a capacity to predict the buoyancy.

Prediction of the Critical Heat Flux (CHF) for upward flow of water in uniformly heated vertical round tube is studied with Artificial Neuron Networks (ANNs) method utilizing different neuron number in hidden layers. This study is based on thermal equilibrium conditions. The neuron number in hidden layers is chosen to vary from 5 to 30 with the step of 5. The effect due to the variety of the neuron number in hidden layers is analyzed. The analysis shows that the neuron number in hidden layers should be appropriate, too less will affect the prediction accuracy and too much may result in abnormal parametric trends. It is concluded that the appropriate neuron number in two hidden layers should be [15 15] in the article.