For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

Tritium control is one of the most critical issues for the development of Molten Salt Reactor (MSR). The point kinetic model has been added into the TAPAS (Tritium trAnsPort chAracteristicS code) to research the dynamic responses of tritium transport characteristics during transients. Two typical transient accident conditions for Transportable Fluoride-salt-cooled High-temperature Reactor (TFHR) are simulated, including unprotected reactivity insertion (URIA) and unprotected overcooling (UOC). The distributions of the key variables in the primary loop and variations with time are obtained including TF, T 2 and Cr 2+ . Numerical results show that during transients, tritium has quite different transport characteristics compared with steady state, ranging from tritium production and speciation, absorption by graphite, Cr corrosion and deposition to tritium permeation. The results are discussed in comparison with earlier findings in detail. This study may provide a valuable reference for the tritium control in MSR.

In-vessel-retention (IVR) has become an important subject of severe accident mitigation strategy. Up to now, many experimental and numerical investigations have been performed on the natural convection characteristics in melt pools with volumetric heating. But these studies are limited to the melt pools under static condition. As floating nuclear reactors become increasingly popular among both commercial and military ships, for successful application of IVR in this occasion, research should be done on the heat transfer characteristics of melt pool under moving conditions. Currently, the specially-designed facility is under construction in Xi'an China for the relevant experiment and numerical studies are performed beforehand. In this paper, a hemisphere with an inner radius of 0.5m, similar to LIVE experimental facility, is chosen to simulate the melt pool. Its flow behaviors under periodic rolling condition are simulated by means of CFD calculation. This work may cast a light on the melt pool characteristics under moving conditions and could be further evaluated by future experimental data.

In order to enhance the inherent safety of sodium-cooled fast reactors, innovative hydraulically suspended absorber rod (HSR) passive shut-down system have been proposed for China demonstration fast reactor. In this study, based on the functional and performance requirements, a full-scale experimental setup has been designed and fabricated for the analysis of the HSR as applied to the prototype reactor. The main characteristic of the test facility is the actuation of the mobile safety rod is triggered by coolant flow rate decrease in the primary loop below half the nominal value and then the rod inserts into the stationary sleeve by gravity. The objective is to investigate the dynamic performance of HSR and establish the laws of its movement at lowering the flow rate modeling the coastdown of primary circulating pump. A series of tests have been performed, including start-up, steady-state operation, loss of flow accident, sensitivity analysis and reliability test. This study also focused on the effect of various factors on scram time, the effect of pump coasting time, rod weight, gap between rod and guide tube, bypass holes, cone angle of rod, flow rate and fluid temperature are analyzed. The experimental results demonstrate the functionality and reliability of the HSR, which would lay foundation for further optimization design.

Thermal stratification phenomena occurring in the upper plenum during a scram transient have an important influence on the structural integrity and the passive safety of sodium-cooled fast breeder reactor (SFR). A two-dimensional thermal-hydraulic analysis code was developed under cylindrical coordinate based on conservation laws of mass, momentum and energy. Block-structured grids were generated to resolve the problems with complicated geometric properties. A second-order scheme based on midpoint rule was applied for the discretization of convection and diffusion terms. Two RANS-type turbulent models, i.e. the standard k–ε model (SKE) and the realizable k–ε model (RKE), are available in this code. A sodium test with scaled model, characterized by large aspect ratio, of a Japanese prototype SFR was used for the validation, mainly from the viewpoints of vertical temperature profiles and rising characteristics of the stratification interface. Results showed that this code could reproduce overall basic behaviors of thermal stratification. The sodium with higher temperature stayed largely stagnant in the upper region under buoyancy effect. Due to the high heat conductivity of sodium, momentum transportation made its leading function. Thus, the RKE model which accounts for the mean deformation rate gave better outcomes than the SKE model.

Multiphase flow widely exists in the nature and engineering. The two-phase flow is the highlight of the studies about the flow in the vessel and steam explosion in nuclear severe accidents. The Moving Particle Semi-implicit (MPS) method is a fully-Lagrangian particle method without grid mesh which focuses on tracking the single particle and concerns with its movement. It has advantages in tracking complex multiphase flows compared with gird methods, and thus shows great potential in predicting multiphase flows. The objective of this thesis is to develop a general multiphase particle method based on the original MPS method and thus this work is of great significance for improving the numerical method for simulating the instability in reactor severe accident and two-phase flows in vessel. This research is intended to provide a study of the instability based on the MPS method. Latest achievements of mesh-free particle methods in instability are researched and a new multiphase MPS method, which is based on the original one, for simulating instability has been developed and validated. Based on referring to other researchers’ papers, the Pressure Poisson Equation (PPE), the viscosity term, the free surface particle determination part and the surface tension model are optimized or added. The numerical simulation on stratification behavior of two immiscible flows is carried out and results are analyzed after data processing. It is proved that the improved MPS method is more accurate than the original method in analysis of multiphase flows. In this paper, the main purposes are simulating and discussing Rayleigh-Taylor (R-T) instability and Kelvin-Helmholtz (K-H) instability. R-T and K-H instability play an important role in the mixing process of many layered flows. R-T instability occurs when a lower density fluid is supported by another density higher fluid or higher density fluid is accelerated by lower density fluid, and the resulting small perturbation increases and eventually forms turbulence. K-H instability is a small disturbance for two different densities, such as waves, at the interface of the two-phase fluid after giving a fixed acceleration in the fluid. Turbulence generated by R-T instability and K-H instability has an important effect in applications such as astrophysics, geophysics, and nuclear science.

Due to great deal of operation experience and technology accumulation, sodium cooled fast reactor (SFR) is the most promising among the six Generation IV reactors, which has advantages of breeding nuclear fuel, transmuting long-lived actinides and good safety characteristics. Thermal-hydraulic computer codes will have to be developed, verified, and validated to support the conceptual and final designs of new SFRs. However, work on developing thermal hydraulic analysis code for SFR is very limited in China, while the common software RELAP5 MOD3 is unable to analyze liquid metal systems. So the modified RELAP5 MOD3.2 is being considered as the thermal-hydraulic system code to support the development of the SFRs. The thermodynamic and transport properties of sodium liquid and vapor have been implemented into the RELAP5 MOD3.2 code, as well as the specific heat transfer correlations for liquid metal. The sodium liquid properties use polynomial equations based on data obtained from Argonne National Laboratory, and the vapor is assumed to be perfect gas. The property equations are acceptably accurate for analysis of SFR, especially for single-phase liquid. New files are added to the fluids directory to generate property tables for new working fluid, which are similar to the table interpolation subroutines for light and heavy water in the original file directory. The method of code modifications are universal for other working fluids and will not affect the code original performance. Some basic verification work for the modified code are carried out. The steam generator of CEFR is analyzed to verify the modified code. The calculated results show that all the water will boil off in the evaporator and the calculated results are in good agreement with the design values. By using modified RELAP5 to model the primary loop of EBR-II fast reactor, the SHRT-17 PLOF test was analyzed. The results show that the natural circulation can be established in the EBR-II primary system after main pumps off to remove the core decay residual heat effectively, and the peak temperature under the safety limits. Moreover, the results computed in this work compared well with the test experimental data for the steady state condition. During the transients, the changing trends of temperature and pressure are similar to experimental data. The discrepancies between calculation and experiment are considered acceptably which need to be improved in the future work. Our work could demonstrate the capability and reliability of the modified RELAP5 for the analysis of SFRs further.

Since the Fukushima accident, an increasing number of studies have considered Accident Tolerant Fuels and cladding properties that can improve the safety performance during an accident in a nuclear reactor. In a research project funded by the US Department of Energy, a group at the University of Wisconsin-Madison focused on modeling the fuel performance for beyond design basis accident conditions. The MELCOR 1.8.6-YR program was used to model the SURRY power station for Zircaloy compared to FeCrAl alloy cladding. The model used an oxidation model for FeCrAl developed by Oak Ridge National Lab (ORNL). The ORNL oxidation model was also used to analyze the fuels’ performance, and these results were compared with the results obtained with the Idaho National Lab (INL) oxidation model for FeCrAl. Results showed that the ORNL model predicted a larger amount of hydrogen generated than the INL model. More experimental data may be needed to better understand the FeCrAl oxidation kinetic mechanisms and perfect oxidation models for this alloy.

In order to investigate the thermal-hydraulic characteristics of LBE in Accelerator Driven System (ADS), the Natural Circulation Capability Loop (NCCL) test facility was designed and constructed at Xi’An Jiaotong university in 2015. NCCL is a middle-scale experimental loop designed for investigating the natural circulation capacity, gas-lift pump enhancing circulation capacity and heat-transfer characteristic of LBE. For the natural circulation capability experiment, the loop is filled with argon gas at 0.2 MPa before filling LBE from store tank. The maximum temperature of LBE is 500 C°, while the normal operation temperature was maintained at 400 C°. In this paper, the LBE natural circulation characteristics were investigated with experiments in NCCL test facility. The study includes measurements on start-up of natural circulation and capability of natural circulation. Significant natural convection flow was observed in the experiments. It was found that the natural circulation was quickly established and stabilized due to LBE high thermal expansion property. It took only a few minutes to have a stable natural circulation prevailing from cold conditions. At the same time, the temperature difference between heating section and cooling section increase quickly and reach to the maximum value. And in the range of 10 minutes, a steady circulation can be performed. The natural circulation flowrate depends on the loop resistance, and the temperature difference between the hot leg and the cold leg, as determined by the power level and the heat sink capacity. The experiments show that the maximum flowrate for the natural circulation is 0∼0.81 kg/s.

The occurrence of Steam-Water Co-Current Flow (SWCOCF) in the AP1000 Pressurizer Surge Line (PZR SL) is an important phenomenon during the depressurization stage of the Small Break Loss-of-Coolant Accidents (SB-LOCAs), and the predictions of SWCOCF pressure drop in PZR SL are important issues in the AP1000 reactor safety analyses. To investigate the pressure drop characteristics of SWCOCF in the PZR SL assembly of AP1000, a PZR SL test model made of stainless steel is designed with scaling ratio of 1:9.15 , and a test loop is also established to supply SWCOCF conditions for the experiments. The test section is mainly composed of a long inclined tube with a short acrylic visual section, a whole surge line tube, and a vertical section. The SWCOCF pressure drop experiments are conducted with pressure ranging from 0.1MPa to 0.24MPa and vapor mass fraction from 0.02 to 0.5 . The research results show that: the whole surge line two-phase friction multiplier increases with the increasing of steam fraction, and decreases with the increasing of pressure. A prediction model of two-phase friction multiplier for the PZR SL is developed based on the Martinelli-Nelson method, which can predict the tests data very well.

In the typical boiling water reactor (BWR), each control rod guide tube supports four fuel assemblies via an orificed fuel support piece in which a channel is designed to be a potential corium relocation path from the core region to the lower head under severe accident conditions. In this study, the improved Moving Particle Semi-implicit (MPS) method was adopted to analyze the melt flow and ablation behavior in this region during a severe accident of BWR. A three-dimensional particle configuration was constructed for analyzing the melt flow behavior within the fuel support piece. Considering the symmetry of the fuel support piece, only one fourth of the fuel support was simulated. The eutectic reaction between Zr (the material of the corium) and stainless steel (the material of the fuel support piece) was taken into consideration. The typical melt flow and freezing behaviors within the fuel support piece were successfully reproduced by MPS method. In all the simulation cases, the melt discharged from the hole of the fuel support piece instead of plugging the fuel support piece. The results indicate that MPS method has the capacity to analyze the melt flow and solidification behavior in the fuel support piece.

Natural convection in a volumetrically heated fluid is of significant importance related to the In-Vessel Retention (IVR) issue. The COPRA facility is designed to simulate the lower plenum of reactor vessel at 1:1 scale for the Chinese advanced PWR, which is a two-dimensional 1/4 circular slice structure with an inner radius of 2.2 m. Water was used as simulant in COPRA-L1 tests. Due to the full scale geometry, The Rayleigh numbers within the corium pools could reach to 10 16 , matching those in the prototypical situation during severe accident. Due to the high Rayleigh numbers and high turbulence in the melt pool, we choose the Large-Eddy Simulation (LES) method to do the numerical calculations. The LES method can better capture the heat transfer characteristic in a high-Rayleigh number volumetrically heated liquid pool. The numerical results of pool temperature and heat flux distribution were compared with those from COPRA-L1 tests.

As the conventional core analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code for motion conditions (SACROM) is developed. To evaluate the effect of different motion conditions on coolant flow, the model of additional forces is established. To check the accuracy of the models, the code has been verified by test data, a commercial subchannel code and a CFD code. In the steady-state verification, the ISPRA data were used and the predicted results agree well with the test data. For the transient simulations without motion conditions, the code COBRA-EN was chosen and the results from SACROM fit the results from COBRA-EN well. And CFX code was used to verify the accuracy of the model of additional forces for motion conditions. The results show that the code can be used in the thermal hydraulic characteristics of the reactor core under motion conditions.

An system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II space reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop components models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original coolant loop radiator, is also modeled, including the two-dimensional heat pipe analysis model, the fin model and the coolant transport duct model. The start-up procedure of the improved TOPAZ-II system are calculated. The results show that the code can be used to obtained the thermal characteristics of the system start-up process.

Xi’an Jiaotong University (XJTU) has developed a containment code called ATHROC (Analysis of Thermal Hydraulic Response of Containment) for the design of a new advanced PWR (Pressurized Water Reactor) in China. In the present work, the well-known NUPEC (NUclear Power Engineering Corporation) tests are simulated to validate the code. NUPEC test M-7-1 and M-4-3 were separately simulated to investigate the mixing effect of hot gas and steam being injected at a low elevation with or without spray cooling in the dome. An analytical model consisting of 26 cells and 67 flow paths was established. In addition, 62 distributed heat structures and 1 lumped heat structure were also modeled in the simulation. The predictions of both simulated tests agree relatively well with the experiment data. The maximum relative errors of predicted temperature for M-7-1 test and M-4-3 test were about 10% and 3%, respectively. The peak pressures of the two tests were about 3.8% and 6.7% overestimated, respectively, and the peak helium concentrations were about 25% and 20% overestimated, respectively. On the other hand, thermal and helium stratification within the containment were well predicted.

Thermal hydraulic characteristics of liquid sodium flowing in an annulus are experimentally studied. The annulus is 1100 mm in length, 6 mm as inside diameter and 10 mm as outside diameter. The heat flux in the experiment is from 50 to 210 kW/m 2 , with Re number from 0 to 18000 and average fluid temperature from 200 °C to 500 °C. Experimental data show that the flow regime of liquid sodium flowing in the annulus can be divided into three regions including laminar flow (Re<2000), transition flow (2000<Re<4000) and turbulent flow (Re>4000). The effects of heat flux, Re number and average fluid temperature of the test section on the heat transfer coefficient are investigated separately. For different regions, correlations for the friction coefficient and for the Nu number are obtained from the experimental data.

In order to gain more insights into the system depressurization and entrainment behavior after actuation of the fourth-stage (ADS-4) valves during a loss-of-coolant-accident, the ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) scaled to AP1000 was constructed to simulate the accident scenario with air-water and steam-water. A brief scaling analysis with emphasis on related thermal hydraulic processes was presented. Entrainment phenomena at vertical up tee branch were observed and analyzed. Preliminary test data of onset of entrainment and entrainment rate were collected with air-water tests and relevant conclusions were obtained.