Analysis Contour Plots in RELAP/SCDAPSIM/MOD 3.4 and MOD 4.1

RELAP/SCDAPSIM is a nuclear reactor simulator and accident analysis code that has been used in the nuclear energy industry for many years. Currently, Innovative Systems Software is developing a new tool that will show the behavior of the core components during a simulation of an accident. The addition of contour plots for the SCDAP components showing different properties: temperature, hydrogen production, etc. will allow a better understanding of core behavior during a severe accident in a nuclear reactor. The SCDAP components are fuel rods, electrically heated simulator rods, such as those used in the CORA experiments, control rods, a shroud and a BWR blade/box. This paper describes the progress in the development of the contour plot tool based on the OpenGL and FORTRAN90 libraries. The purpose of this tool is help to the user analyze the simulation of an accident and to debug an input file.

Numerical Investigation of Corium Coolability in Core Catcher: Sensitivity to Modeling Parameters

To avoid settling of molten materials directly on the vessel wall in severe accident sequences, the implementation of a ‘core catcher’ device in the lower plenum of sodium fast reactor designs is considered. The device is to collect, retain and cool the debris, created when the corium falls down and accumulates in the core catcher, while interacting with surrounding coolant. This Fuel-Coolant Interaction (FCI) leads to a potentially energetic heat and mass transfer process which may threaten the vessel integrity. For simulations of severe accidents, including FCI, the SIMMER code family is employed at KIT. SIMMER-III and SIMMER-IV are advanced tools for the core disruptive accidents (CDA) analysis of liquid-metal fast reactors (LMFRs) and other GEN-IV systems. They are 2D/3D multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics codes coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. However, the experience of SIMMER application to simulation of corium relocation and related FCI is limited. It should be mentioned that the SIMMER code was not firstly developed for the FCI simulation. However, the related models show its basic capability in such complicate multiphase phenomena. The objective of the study was to preliminarily apply this code in a large-scale simulation. An in-vessel model based on European Sodium Fast Reactor (ESFR) was established and calculated by the SIMMER code. In addition, a sensitivity analysis on some modeling parameters is also conducted to examine their impacts. The characteristics of the debris in the core catcher region, such as debris mass and composition are compared. Besides that, the pressure history in this region, the mass of generated sodium vapor and average temperature of liquid sodium, which can be considered as FCI quantitative parameters, are also discussed. It is expected that the present study can provide some numerical experience of the SIMMER code in plant-scale corium relocation and related FCI simulation.

Validation of Film Evaporation Model in GASFLOW-MPI

A dynamic film model is developed in the parallel CFD code GASFLOW-MPI for passive containment cooling system (PCCS) utilized in nuclear power plant like AP1000 and CAP1400. GASFLOW-MPI is a widely validated parallel CDF code and has been applied to containment thermal hydraulics safety analysis for different types of reactors. The essential issue for PCCS is the heat removal capability. Research shows that film evaporation contributes most to the heat removal capability for PCCS. In this study, the film evaporation model is validated with separate effect test conducted on the EFFE facility by Pisa University. The test region is a rectangle gap with 0.1m width, 2m length, and 0.6m depth. The water film flowing from the top of the gap is heated by a heating plate with constant temperature and cooled by countercurrent air flow at the same time. The test region model is built and analyzed, through which the total thermal power and evaporation rate are obtained to compare with experimental data. Numerical result shows good agreement with the experimental data. Besides, the influence of air velocity, wall temperature and gap widths are discussed in our study. Result shows that, the film evaporation has a positive correlation with air velocity, wall temperature and gap width. This study can be fundamental for our further numerical study on PCCS.

Safety Analysis Model of DUCG Based on FMEA/FTA

Failure Modes and Effects Analysis (FMEA) is a useful tool to find possible flaws, to reduce cost and to shorten research cycle in complex industrial systems. Fault Tree Analysis (FTA) has gained credibility over the past years, not only in nuclear industry, but also in other industries like aerospace, petrochemical, and weapon. Both FMEA and FTA are effective techniques in safety analysis, but there are still many uncertain factors in them that are not well addressed until now. This paper combines FMEA and FTA based on Dynamic Uncertain Causality Graph (DUCG) to solve this issue. Firstly, the FMEA model is mapped into a corresponding DUCG graph. Secondly, FTA model is mapped into a corresponding DUCG graph. Thirdly, combine the above DUCG graphs. Finally, users can modify the combined DUCG graph and calculations are made. This paper bridges the gap between FMEA and FTA by combining the two methods using DUCG. And additional modeling power and analytical power can be achieved with the advantages of the combined DUCG safety analysis model and its inference algorithm. This method can also promote the application of DUCG in the system reliability and safety analysis. An example is used to illustrate this method.

Data-Driven Fault Diagnosis for Nuclear Power Plant: The Implicit Model Approach

This paper proposes a data-driven fault detection approach for nuclear power plant. The approach starts from input and output (I/O) data obtained from operating data of industrial process. Due to the model is not explicitly appeared, the proposed approach is named as implicit model approach (IMA). Residual generator is obtained directly from I/O data rather than from the mechanism, based which the algorithm of IMA-based fault detection is proposed. The main advantage of IMA-based fault detection is that it can circumvent complicated model identification. The approach generates parameterized matrices of residual signal inspired by subspace relevant technology without any prior knowledge about mechanisms of the plant. Fault information has been injected to a simulating platform of a compact reactor in the simulation part, by which we verify the effectiveness of IMA-based fault detection.

OECD-NEA Expert Group on Multi-Physics Experimental Data, Benchmarks and Validation

High-fidelity, multi-physics modeling and simulation (M&S) tools are being developed and utilized for a variety of applications in nuclear science and technology and show great promise in their abilities to reproduce observed phenomena for many applications. Even with the increasing fidelity and sophistication of coupled multi-physics M&S tools, the underpinning models and data still need to be validated against experiments that may require a more complex array of validation data because of the great breadth of the time, energy and spatial domains of the physical phenomena that are being simulated. The expert group on Multi-Physics Experimental Data, Benchmarks and Validation (MPEBV) of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) was formed to address the challenges with the validation of such tools. The work of the MPEBV expert group is shared among three task forces to fulfill its mandate and specific exercises are being developed to demonstrate validation principles for common industrial challenges. This paper describes the overall mission of the group, the specific objectives of the task forces, the linkages among the task forces, and the development of a validation exercise that focuses on a specific reactor challenge problem.

Preliminary Design of Unloading Device for MNSR LEU Conversion

In order to solve the problem that the current unloading operation will destroy the sealing performance of Miniature Neutron Source Reactor (MNSR) reactor vessel and the tightness can’t be restored, and to meet the application requirements that the original reactor vessel will be reloaded and operated after MNSR LEU conversion, the new unloading device is designed, which can be used without separation of reactor vessel. There has only one fuel assembly in MNSR. When the fuel assembly are unload for MNSR LEU conversion, the cover plate of the pool is removed, the cadmium string is put in, and the neutron detector is placed at first. After removing the drive mechanism and the control rod, and opening the small cover plate at the top of reactor vessel, the fuel assembly can be grabbed and unloaded by unloading tool only through the opening of the small top cover plate. The MNSR spent fuel has very high radioactivity. The auxiliary mechanical device can be used with unloading tools to realize operation in a long distance by lifting and level motion, which is convenient to shield and can reduce the works’ irradiation dose level effectively. Through calculation and analysis, the results show that the structure strength of unloading device is much larger than the actual load to ensure operation safety and reliability. The unloading device is easy to process and operate, and can be used in the practical operation of MNSR LEU conversion or decommissioning at home and abroad to simplify the operation steps and improve the working efficiency.

Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions

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.

Muon Tomography for Measuring Amount of Nuclear Materials in Fuel Debris

A method to measure an amount of nuclear materials in fuel debris by using muon tomography has being developed for proceeding with decommissioning of Fukushima Daiichi nuclear power plant. As a result of the Fukushima Daiichi nuclear disaster, the molten fuels were mixed with reactor structures and accumulated as fuel debris in the reactor buildings. There is still a large amount of fuel debris remained in each reactor. Fuel debris removal is planned in the near future and the debris will be taken out in this process. The debris need to be inspected from a viewpoint of nuclear material control. Since the debris is a mixture of fuel and other structures, it is hard to quantitate nuclear materials in debris by existing measurement method. Muons are cosmic-ray particles which have high energies, therefore, they are highly penetrative. This feature makes muon tomography sensitive to find heavy materials such as uranium or plutonium. We conducted a simulation study of applying muon tomography to measure fuel debris by using a Monte-Carlo method. A simulation model which includes muon detectors, shielding container and fuel debris was constructed to reproduce a measurement situation at the site. In conclusion, muon tomography quantitate the nuclear materials, therefore, this method should be useful for the fuel debris removal of Fukushima Daiichi reactors.