Enhanced Verification for RELAP5-3D Parameter and Sensitivity Studies

The RELAP5-3D1 program solves a complex system of governing, closure and special process equations to model the underlying physics of nuclear power plants. For SQA (software quality assurance), the code must be verified and validated (V&V) to ensure proper performance before release to users. The physical models are validated against data from experiments and plants and verified against specifications for the computer code. In addition to physics, programs such as RELAP5-3D perform numerous other functions and processes that should also be checked to guarantee correct results. Functions include input, output, data management, and user interaction, while processes include restart, time-step backup, code coupling, and multi-case processing. Previous articles have covered the verification of the physical models, restart, and backup through extremely accurate and automated sequential verification applied on a comprehensive suite of test cases to ensure that code changes produced no unintended consequences. New developments have enabled the verification of multi-case and multi-deck processing. These features are frequently used in parameter and code sensitivity studies and therefore must be verified as working correctly. Both theory and application are presented.

Corrosion of Hot-Dip Galvanized Containment Installations: A Potential Cause for Thermal-Hydraulic Effects After LOCA in PWR?

Generic investigations regarding the influence of corrosion processes of hot-dip galvanized PWR containment installations on strainer clogging as well as on the coolant chemistry and possible resulting in-core effects are carried out within joint research projects of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), TU Dresden (TUD) and Zittau-Görlitz University of Applied Sciences (HSZG). Lab-scale experiments at HZDR and TUD are focused on elucidation of physico-chemical corrosion and precipitation processes as well as resulting clogging effects. Results of generic experiments in a lab-scale corrosion test facility suggest that there is a multi-stage corrosion process. The first stage comprises dissolution of the zinc layer in the coolant forming zinc ions and in turn affecting the coolant chemistry. During the second stage, the base material (steel) corrodes forming insoluble corrosion particles, which can subsequently lead to accelerated clogging of fiber-laden strainers within a few hours. The main influences on corrosion were identified as impact of the coolant jet onto the corroding surface, water chemistry and zinc surface / coolant volume ratio. Furthermore, retrograde solubility of zinc corrosion products in boric acid containing coolants with increasing temperature was observed. Thus, formation and deposition of solid corrosion products cannot be ruled out if zinc containing coolant is heated up during its recirculation into hot downstream components (e.g. hot-spots in core). Corrosion experiments, which included formation of corrosion products at a heated cladding tube, proved that zinc, dissolved in the coolant at low sump temperatures, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces. Due to alternating heating and cooling of the coolant during sump recirculation operation, a cycle of zinc corrosion and zinc borate precipitation may be initiated, which may eventually influence the thermal hydraulics in downstream components during the post-LOCA stage. The results obtained at lab-scale were confirmed by corresponding experiments in semi-technical test facilities of the project partner HSZG. Based on the experimental results, water chemical measures were tested to reduce corrosion and/or zinc borate precipitation effects.

Heat Transfer Coefficient Measurement for Downward-Facing Heat Transfer on Curved Rectangular Water Channel

In-Vessel Retention by External Reactor Vessel Cooling (IVR-ERVC) is method of removing the decay heat by cooling reactor vessel after corium relocation, and is also one of severe accident management strategies. Estimating heat transfer coefficients (HTCs) is important to evaluate heat transfer capability of the ERVC. In this study, the HTCs of outer wall of reactor vessel lower head were experimentally measured under the IVR-ERVC situation of Large Loss of Coolant Accident (LLOCA) condition. Experimental equipment was designed to simulate flow boiling condition of ERVC natural circulation, and based on APR+ design. This study focused on effects of real reactor vessel geometry (2.5 m of radius curvature) and material (SA508) for the HTCs. Curved rectangular water channel (test section) was design to simulate water channel which is between the reactor vessel lower head outer wall and thermal insulator. Radius curvature, length, width and gap size of the test section were respectively 2.5 m, 1 m, 0.07 m and 0.15 m. Two connection parts were connected at inlet and outlet of the test section to maintain fluid flow condition, and its cross section geometry was same with one of test section. To simulate vessel lower head outer wall, thin SA508 plate was used as main heater, and test section supported the main heater. Thickness, width, length and radius curvature of the main heater were 1.2 mm, 0.07 m, 1 m and 2.5 m respectively. The main heater was heated by DC rectifier, and applied heat flux was under CHF value. The test section was changed for each experiment. The HTCs of whole reactor vessel lower head (bottom: 0 ° and top: 90 °) were measured by inclining the test section, and experiments were conducted at four angular ranges; 0–22.5, 22.5–45, 45–67.5 and 67.5–90 °. DI water was used as working fluid in this experiment, and all experiments were conducted at 400 kg/m2s of constant mass flux with atmospheric pressure. The working fluid temperatures were measured at two point of water loop by K-type thermocouple. The main heater surface temperatures were measured by IR camera. The main heater was coated by carbon spray to make uniform surface emissivity, and the IR camera emissivity calibration was also conducted with the coated main heater. The HTCs were calculated by measured main heater surface temperature. In this research, the HTC results of 10, 30, 60 and 90 ° inclination angle were presented, and were plotted with wall super heat.

Evaluation of Sodium Pool Fire and Thermal Consequence in Two-Cell Configuration

Evaluation of accidental sodium leak, combustion, and its thermal consequence is one of the important issues to be assessed in the field of sodium-cooled fast reactor (SFR) since the liquid sodium is chemically active and might give thermal load to plant building structure due to its exothermic reaction with oxygen in air atmosphere. Therefore, many experimental investigations and numerical simulation tools development have been and still now are being carried out to understand the details of sodium fire behaviors and to contribute to the investigation and preparation of appropriate mitigation measures in the plant design. From various kinds of sodium fire situations, the present paper treats the sodium pool fire and subsequent heat transfer behavior in air atmosphere two-cell geometry both experimentally and analytically because such two-cell configuration is considered as the typical one to possess important characteristic of multi-compartment system seen in an actual plant. Main description of this paper consists of a sodium pool fire experiment that was performed in a rectangular-shaped two-cell system with an opening between the cells, and the discussion of the experimental results. Inner volume of the experimental cells is about 70 m3. The amount of used sodium and the prepared pool surface area in the experiment are about 55 kg and 2.25 m2, respectively. The experiment has provided the temperature data measured in more than 100 positions for atmospheric gas and structures other than the data of oxygen concentration and suspended sodium aerosols concentration in the cells. The analyses of the measured data clarify the basic characteristics of sodium pool combustion and consequential heat and mass transfer in the cells, for instance, suggesting several features of multidimensional thermal-hydraulic behaviors such as thermal stratification near the opening between the two cells. In the discussion, numerical analysis using a lumped-parameter based zonal model safety analysis code ‘SPHINCS’ and the comparison of its results with the experimental data are also carried out to investigate the validity and applicability of the code to this type of sodium fire situation.

Numerical Analysis of the Corium Behavior Within the Fuel Support Piece by MPS

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.

Dynamic Processes During Pulsed Gas Injection Into Liquid Column

An experimental study of dynamic processes during pulsed gas injection into quiescent liquids was performed. Both water and low melting temperature metal alloy were used as test liquids. Air and argon were used as gas phase. The test sections were vertical cylindrical columns 25 and 68 mm inner diameter. Measurements of flow parameters during gas injection were performed. Water – air experiments were performed at room temperature, the temperature of liquid metal alloy was 135 deg C. Time records of pressure in the liquid and in gas phase above the liquid were obtained. Measurements of liquid temperature and level of liquid surface were performed. It was shown that at pulse gas injection into liquid metal high amplitude pressure fluctuation may arise. Also the fluctuation variation of the free surface of the liquid may appear which are connected with the oscillations of the gas volume. Experimental data obtained may be used for verification & validation of modern CFD codes.

Analysis and Research on Natural Circulation Capacity of HFETR

Natural circulation is being used as an important circulation to remove reactor residual heat. In the core of High Flux Engineering Trial Reactor of China (HFETR), the coolant is driven by pumps normally and flows from upside to downside in the core. When HFETR is shut down or runs in low power, the natural circulation between the hot water in the core and the cold water in the reflector inside the pressure vessel is established to cool down the core. Since the natural circulation processed only in the pressure vessel, the accident pumps need to be turned on periodically to remove reactor residual heat. The inversion of flow direction in HFETR and internal natural circulation lead to a different natural circulation establishment process from traditional reactor in which coolant flows form down to top normally. In this paper the transition between the natural circulation and forced circulation is analyzed by RELAP5/MOD3 code. The results showed that the accident pump could be turned off in the power of 850kW; The time, at which the accident pump needs to be turned on to transit the natural circulation to forced circulation, is decided by the temperature of the water in top of pressure vessel, and a formula between temperature of the water in the top of pressure vessel and the reactor power was obtained. The research results have theoretical and practical value to the full use of the natural circulation ability, as well as the safety of the engineering reactors or similar test facilities.

Heat and Mass Transfer Analogy of Steam Condensation With Air Over Vertical External Surface

During the steam condensation, the presence of non-condensable gases is an important issue affecting the efficiency of the whole thermodynamic process. For this reason, many researchers investigated it by theoretical or experimental methods. A heat and mass transfer analogy model on steam condensation in presence of air over the vertical external surface based on the diffusion layer model is modified in the present paper. Based on previous authors’ experience, the suction effect at the gas-liquid interface and other analogy drawbacks are identified and overcome by supplementing it with more detailed analysis as well as targeted experiments. The experimental data obtained for condensation, outside vertical tube with an external diameter of 38 mm, of air/steam and helium/air/steam mixture, have been used to verify the present heat and mass transfer analogy formulation. By comparing against different available experimental data and previous formulations, the heat and mass transfer analogy formulation is demonstrated to be a accurate enough theoretical approximation. The deviation between predicted values of the new model and experiment results of this paper is less than 15% which has relative higher precision.

Numerical Simulation of Three Dimensional Internal Flow of a PWR Reactor

The coolant flow in the reactor pressure vessel (RPV) lower plenum is complex due to the presence of various internal structures, which has a great influence on the flow distribution at the core inlet. In order to study the thermal hydraulic characteristics in the RPV lower plenum, many scaled down test facilities have been built for different PWR reactors such as Juliette, ACOP, and ROCOM. Although the experimental study is still a main research method, it may be not economical in some situations due to the high cost and the long study period. Compared with the experimental method, Computational Fluid Dynamics (CFD) methodology can simulate three dimensional fluid flow in complex geometries and perform parametric studies more easily. The detailed and localized thermal hydraulic characteristics which are difficult to measure during experiments can be obtained. So CFD simulation has been widely used nowadays. One of the purposes of numerical simulations of the internal flow in a RPV is to get the flow distribution at the core inlet, then to make an optimization for the flow diffusor in the RPV lower plenum to improve the core inlet flow distribution homogeneity. Appropriate optimizations for the flow diffusor depends on fully understanding the flow phenomena in the RPV lower plenum. In this paper, Phenomenon Identification and Ranking Table (PIRT) is adopted to analyze the physical phenomenon that occurs in the RPV lower plenum with the typical 900MW reactor internal structures, and the importance of the various physical phenomena and the reference parameters are ranked through expert opinions and literature review. Then a preliminary three dimensional CFD simulation for the reactor vessel is conducted. The main phenomena identified by the PIRT can be observed from the simulation results.

Heat Transfer Simulation of the Fuel Transport Cask for Spherical Fuel Elements in Molten Salt Reactor

Fuel transport is an indispensable task for nuclear power plants. For pressurized water reactors (PWR) and boiling water reactors (BWR), many research projects have been completed for designing and testing the transport casks for fresh fuel as well as spent fuel [1–3]. To ensure the safety of nuclear fuel during the transportation, many aspects should be analyzed and examined for the casks with fuel inside, such as heat transfer and temperature calculation, radiation protection, nonproliferation issues, etc. The transport cask discussed in this paper is especially for new spherical fuel elements, which should be designed in accordance with the stipulations in the GB11806 Regulations for the Safe Transport of Radioactive Material [4]. The Transport Cask for spherical fuel elements used in molten salt reactor (MSR) should be designed in accordance with the safety standards for transport of radioactive material. It is necessary to evaluate the thermal performance of the transport cask separately in normal transport condition and in accident transient. The MSR fuel sphere elements cask is in a circular cylinder shape and composed of inner container and outer shell cask. The objective of the thermal analysis of the cask under hypothetical accident conditions is to demonstrate that the cask containment boundary structural components are maintained within their safe operating temperature ranges. The heat transfer process (conduction, convection, and radiation) is simulated by ANSYS-CFX in this paper and it is demonstrated that the components of cask are maintained in safe operating temperature ranges. The calculation results are below limit temperatures, indicating that the thermal design of the cask could meet the Standard Regulations. The result is also compared with the fire test, which shows the calculation model is conservative and rational.