A Study of GOTHIC 8.0 Code Application to AP1000 Containment Response

The AP1000 containment model has been developed by using WGOTHIC version 4.2 code. Condensation heat and mass transfer from the volumes to the containment shell, conduction through the shell, and evaporation from the shell to the riser were all calculated by using the special CLIMEs model. In this paper, the latest GOTHIC version 8.0 code is used to model both condensation and evaporation heat and mass transfer process. An improved heat and mass transfer model, the diffusion layer model (DLM), is adopted to model the condensation on the inside wall of containment. The Film heat transfer coefficient option is used to model the evaporation on the outside wall of containment. As a preliminary code consolidation effort, it is possible to use GOTHIC 8.0 code as a tool to analysis the AP1000 containment response.

Full Implicit Integrate Solution to the Coupled Neutronics/Thermal-Hydraulics Problems

The traditional solution of the coupled neutronics/ thermal-hydraulics problems has typically been performed by solving the individual field separately and then transferring information between each other. In this paper, full implicit integrate solution to the coupled neutronics/ thermal-hydraulic problem is investigated. There are two advantages compared with the traditional method, which are high temporal accuracy and stability. The five equations of single-phase flow, the solid heat conduction and the neutronics are employed as a simplified model of a nuclear reactor core. All these field equations are solved together in a tightly coupled, nonlinear fashion. Firstly, Newton-based method is employed to solve nonlinear systems due to its local second-order convergence rate. And then the Krylov iterative method is used to solve the linear systems which are from the Newton linearization. The two procedures above are the so-called Newton-Krylov method. Furthermore, in order to improve the performance of the Krylov method, physics-based preconditioner is employed, which is constructed by the physical insight. Finally, several Newton-Krylov solution approaches are carried out to compare the performance of the coupled neutronics / thermal-hydraulic equations.

The Human Reliability Analysis in Fire PSA Using Spar-H Method

According to existing research results, fire risk makes a significant contribution to the total risk of a nuclear power plant (NPP). So fire probabilistic safety analysis (PSA) for NPPs is becoming more and more important in recent years. How to perform human reliability analysis (HRA) which is an essential part of PSA is therefore being paid more and more attention in fire PSA. This paper describes the characteristics and special considerations of HRA in fire PSA, and demonstrates in fire PSA how to use SPAR-H method which is so-called an advanced second-generation HRA method and is being widely used in PSA for Chinese NPPs. The study results can be a reference for other HRA analysts to use SPAR-H method in fire PSA models or other PSA models in Chinese NPPs or the world-wide nuclear industry.

The Modelling and Coupling Methodology of ANSYS CFX Using Porous Media for PB-AHTR

This paper introduces a first-principle steady-state coupling methodology using the Monte Carlo Code RMC and the CFD code CFX which can be used for the analysis of small and medium reactors. The RMC code is used for neutronics calculation while CFX is used for Thermal-Hydraulics (T-H) calculation. A Pebble Bed-Advanced High Temperature Reactor (PB-AHTR) core is modeled using this method. The porous media is used in the CFX model to simulate the pebble bed structure in PB-AHTR. This research concludes that the steady-state coupled calculation using RMC and CFX is feasible and can obtain stable results within a few iterations.

Predicting Thermal Mixing and Fatigue Inside Control Rod Guide Tubes

The cracked control rods shafts found in two Swedish NPPs were subjected to thermal fatigue due to mixing of cold purge flow with hot bypass water in the upper part of the top tube on which the control rod guide tubes rests. The interaction between the jets formed at the bypass water inlets is the main source of oscillation resulting in low frequency downward motion of hot bypass water into the cold purge flow. This ultimately causes thermal fatigue in the control rod shaft in the region below the four lower bypass water inlets. The transient analyses shown in this report were done to further investigate this oscillating phenomenon and compare to experimental measurements of water temperatures inside the control rod guide tube. The simulated results show good agreement with experimental data regarding all important variables for the estimation of thermal fatigue such as peak-to-peak temperature range, frequency of oscillation and duration of the temperature peaks. The results presented in this report show that CFD using LES methodology and the open source toolbox OpenFOAM is a viable tool for predicting complex turbulent mixing flows and thermal loads.

Three-Dimension Thermal Hydraulic Analysis of Stationary Target Plate for Highly Intensified Neutron Generator (HINEG)

Highly intensified neutron generator (HINEG) is a D-T neutron generator tritium target system; it can be used in researching fusion energy and advanced fission energy. The heat flux at the target plate is extraordinarily high, cooling the target plate effectively, limiting the working temperature, are the keys to keep the normal working of HINEG. This paper has simulated the 3D values of some different kind of target plate systems, using the computational fluid dynamics software and static analysis software, presented the stationary analysis results of temperature and fluid fields, as well as stress field. The flow and temperature distributions provide important data for advanced design and performance evaluation of HINEG.

Application of RSE-M2010 on In-Service Inspection of Taishan EPR Project

The latest edition of French In-service Inspection Rules RSE-M2010, incorporating the up-to-date upstream French regulations, orders and requirements for pressure equipments, and taking into account both of the radioactive risk and industrial risk in nuclear power plant (NPP), has been adopted as the applicable rule for in-service inspection (ISI) of EPR units. In RSE-M2010, the previously used benchmark for classification Safety Class has been replaced by the Nuclear Pressure Equipments Class (ESPN Class), and the category of pressure equipments has been introduced to monitor the industry risks of NPP pressure equipments, making it much more precise and convenient to define the scope of equipments which subjected to ISI and corresponding ISI requirements on frequency and methods. This paper described the main differences of the ISI requirements in RSE-M2010 and previous edition of RSE-M, also introduced practices of applying RSE-M2010 when preparing the ISI program of Taishan EPR units. Based on the application practice of RSE-M2010 on Taishan EPR project, some proposals for future improvement of this code are presented. Preliminary thinking for future implementation of EPR ISI activities has also been described.

Thermal-Hydraulic Simulations of Guard Containment in Depressurization Accidents of a Gas-Cooled Fast Reactor

A gas-cooled fast reactor is designed as an advanced nuclear reactor in next generation in the EU. In depressurization accident scenarios, pressurization caused by a release of helium from the primary system with a higher pressure into the guard containment would endanger the integrity of the containment. In the design stage, the released source term is analyzed theoretically, and is applied as a boundary condition in the 3D CFD code simulation to the transient pressurization process. The simulation results supply a reference value about the design pressure of the containment.

ALWR PRA Standard Development

In 2012 a review and comment ballot containing requirements for use of the Standard on ALWRs was issued. The ballot is based on the ASME/ANS PRA Standard (i.e., RA-Sa-2009). The ballot was developed by the Joint Committee on Nuclear Risk Management (JCNRM) working group (WG) on ALWRs. Based on the charter of the working group and stakeholder interest, Parts 1 through 5 of the Standard have been addressed. Thus, address internal events at power, internal flooding at power, internal fires at power, and seismic at power have been addressed. Other external events started in 2012 and will be balloted in 2013. Low power and shutdown (LP/SD) modes, and Level 2 PRA and Level 3 PRA Standards (when they are available) will then be considered. A formal ballot for Parts 1 through 5 is planned for 2013. This paper provides the approach and results. The mandatory appendix is important because an ALWR plant in the preoperational cannot meet the standard as written, and the US NRC expects that the PRA will meet the Standard. Thus changes are needed to the standard to address the differences in preoperational and operational plants and the differences in current generation and ALWR plants. This will assist vendors and licensees in successfully developing PRAs to meet US NRC requirements.

Experimental Analyses by SIMMER-III on Duct-Wall Failure and Fuel Discharge/Relocation Behavior

This paper describes experimental analyses using SIMMER-III/IV, which are two/three-dimensional multi-component multi-phase Eulerian fluid-dynamics codes, for the purpose of the code validation. Two topics of key phenomena in core disruptive accidents were presented in this paper: duct-wall failure and fuel discharge/relocation behavior. To analyze the duct-wall failure behavior, the SCARABEE BE+3 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall event progression; which was characterized by coolant boiling, clad melting, fuel failure, molten pool formation, duct-wall failure, etc.; observed in the experiment. The CAMEL C6 experiment investigated the fuel discharge and relocation behavior through a simulated control rod guide tube, which is important in evaluating the neutronic reactivity. SIMMER-IV well simulated fuel-coolant interaction, sodium voiding, fuel relocation behavior observed in the experiment. These experimental analyses indicated the validity of the SIMMER-III/IV computer code for the duct wall failure and fuel discharge/relocation behavior.