Verification of Control Rod Assembly Homogenization for LMFBR With a New Lattice Physics Code GALAXY-H

2013 ◽  
Author(s):  
Yohei Kamiyama ◽  
Hiroki Koike ◽  
Kazuki Kirimura ◽  
Kazuya Yamaji ◽  
Shinya Kosaka ◽  
...  
Keyword(s):  
Cell Model ◽  
Severe Accident ◽  
Assembly Model ◽  
Sph Method ◽  
Control Rod ◽  
Super Cell ◽  
Fuel Assemblies ◽  
Transport Calculation ◽  
Nuclear Constants ◽  
Calculation Code

A new FBR lattice physics code GALAXY-H has been developed by Mitsubishi Heavy Industries, Ltd. (MHI). GALAXY-H is a hexagonal version of GALAXY, which is a two dimensional transport calculation code for PWR assembly. GALAXY-H generates assembly nuclear constants used in the FBR core calculation code. The methodology of flux calculation for GALAXY-H is based on the method of characteristics (MOC) as well as GALAXY. The fuel assemblies of Japanese demonstrated and commercial FBRs are intended to contain the inner duct called FAIDUS where molten fuel is removed to prevent re-critical at severe accident. One of the objectives for developing GALAXY-H is to treat the inner duct and wrapper tube configurations exactly. In this paper, the method generating nuclear constants of control rod assembly is developed with multi-assembly model to exclude the super-cell model that has been used in FBR design so far. Besides, GALAXY-H employs the SPH method for reduction of homogenization error, which is popular method in LWR design. From this, the advanced nuclear constants calculation method for FBR control rod assembly is developed and the basic applicability of FBR nuclear design by using GALAXY-H is confirmed.

Enhancement of Plant Availability by Outage Shortening in European Advanced Boiling Water Reactor (EU-ABWR)

2013 ◽  
Author(s):  
Kazuhiro Kamei ◽  
Kazuyoshi Kataoka ◽  
Kazuto Imasaki ◽  
Noboru Saito
Keyword(s):  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Plant Availability ◽  
Control Rod ◽  
Water Reactor ◽  
Spray System ◽  
Maintenance And Inspection ◽  
Reactor Pool ◽  
Safety Systems

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash, severe accident mitigation systems, the N+2 principle in safety systems, the diversity principle and a large output of 1600 MWe. These features enable EU-ABWR’s design objectives and principles to be consistent with the requirements in the Finnish utility and the safety requirements of Finnish YVL guide. By adopting Scandinavian outage processes, the Plant Availability is aimed to be greater than 95%. ABWRs have an excellent design potential to acheive short outage duration (e.g., shortening of maintenance and inspection duration by applying Fine Motion Control Rod Drive and Reactor Internal Pump). In addition, the EU-ABWR applies following key design improvements to reduce a refueling outage duration; a) Direct Reactor Pressure Vessel (RPV) Head Spray System, b) Self-standing Control Rods and c) Water shielding reactor pool. In this paper, coolability of RPV due to application of the Direct RPV Head Spray System is also verified with numerical evaluations by Computation Fluid Dynamics (CFD) analysis.

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

2016 ◽  
Author(s):  
Ronghua Chen ◽  
Lie Chen ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Eutectic Reaction ◽  
Flow Behavior ◽  
Three Dimensional ◽  
Severe Accident ◽  
Guide Tube ◽  
Melt Flow ◽  
Control Rod ◽  
Mps Method ◽  
Moving Particle ◽  
Accident Conditions

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.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

scholarly journals Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

2017 ◽  
Vol 2017 ◽  
pp. 1-25 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Light Water Reactors ◽  
Control Rod ◽  
Light Water ◽  
Severe Accidents ◽  
Fuel Bundle ◽  
Water Reactors

The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

Study on Eutectic Melting Behavior of Control Rod Materials in Core Disruptive Accidents of Sodium-Cooled Fast Reactors: Thermophysical Properties of Eutectic Mixture Containing of High Concentration Boron in a Solid State

2020 ◽  
Author(s):  
Toshihide Takai ◽  
Tomohiro Furukawa ◽  
Hidemasa Yamano
Keyword(s):  
Stainless Steel ◽  
Solid State ◽  
Boron Carbide ◽  
Specific Heat ◽  
Thermophysical Properties ◽  
Crystalline Phase ◽  
Eutectic Mixture ◽  
Severe Accident ◽  
Eutectic Melting ◽  
Control Rod

Abstract In a core disruptive accident scenario, boron carbide, which is used as control rod material, may melt below the melting temperature of stainless steel due to the eutectic reaction with it. Produced eutectic mixture is assumed to relocate widely in the degraded core, and this behavior plays an important role to reduce the neutronic reactivity of the degraded core materials significantly. However, these behaviors have never been simulated in the severe accident computer codes, and reducing the uncertainty is important for reasonable assessment. To contribute improvement of the core disruptive accident analysis code to handle these eutectic melting and relocation behavior, authors had been carried out the evaluation of the thermophysical properties of stainless steel containing boron carbide, which needed as a basic data for cord improvement. Since the solubility range of boron against iron is expected to be wide, the crystalline phase of eutectic mixture may change according to boron concentration in the eutectic mixture. And this may affect the thermophysical properties themselves. In this work, the density and specific heat of stainless steel containing 17 mass% boron carbide in a solid state are obtained and compared with these of stainless steel containing 0 and 5 mass% boron carbide. By adding 17 mass boron carbide to stainless steel type 316L, the density decreased approximately 24% and the specific heat increased approximately 25% at 293 K. The density of stainless steel containing boron carbide tended to decrease almost linearly depending on the amount of boron carbide added, none the less for difference of crystalline phase. On the other hand, increasing trend of the specific heat of stainless steel containing 17 mass% boron carbide accompanying elevating temperature showed different behavior from that of stainless steel containing 0 and 5 mass% boron carbide. This difference in the trend of the specific heat was considered to be caused the difference in the crystalline phase.

A THEORETICAL STUDY OF POINT DEFECTS IN ZIRCONIA – SILICON INTERFACES

2006 ◽  
Vol 16 (01) ◽  
pp. 389-396
Author(s):  
Masashi Nakatomi ◽  
Koichi Yamashita
Keyword(s):  
Point Defects ◽  
Density Functional ◽  
Theoretical Study ◽  
Gate Dielectrics ◽  
Cell Model ◽  
Interstitial Oxygen ◽  
Super Cell ◽  
Defect Levels ◽  
Silicon Interface ◽  
Oxygen Atoms

We present a theoretical study on the point defects in ZrO 2–silicon interfaces using molecular dynamics (MD) calculations. A super-cell model that contains 9 atomic layers of silicon and 9 atomic layers of ZrO 2 was used for the simulation. Three atomic layers containing 17 oxygen atoms, eight silicon atoms, and nine Zr atoms were used to simulate the ZrO 2–silicon interface. We then performed density functional theory (DFT) with plane-wave basis to calculate the interface band structure. Results demonstrate that the stretched Zr – O bonds at the interface would produce some defect levels in the band gap. Particularly, the defect levels originated from the interstitial oxygen atoms are located close to the bottom of the ZrO 2 conduction band and hence it will affect the electrical properties of the gate dielectrics.

Lingao II Source Term Analysis of a Mitigated Severe Accident Scenario

2008 ◽  
Author(s):  
Yaodong Chen ◽  
Yongzhong Wu ◽  
Hui Zhang
Keyword(s):  
Severe Accident ◽  
Primary System ◽  
Start Up ◽  
System Collapse ◽  
Fuel Assemblies ◽  
Level 1 ◽  
Dry Out ◽  
Power Recovery ◽  
Term Analysis ◽  
Typical Scenario

A typical scenario, Station-Black-Out, which makes up an important part of contribution to core degradation frequency of three loop PWR Lingao phase II according to the PSA level 1 studies, was selected in this paper for Severe Accident Analysis with the application of combination of MELCOR and COCOSYS. Of the scenario, concurrent with postulated power recovery before degradation proceeded to failure of lower head, the evolution of in_vessel phase was studied; a detailed phenomena covering heat-up and dry-out of primary system, collapse and melt down of fuel assemblies, thermal hydraulic response inside the containment compartments, release and accumulation of H2 and fission product (FP) source terms, their leakage into environment were simulated and investigated. The effectiveness of preventive and mitigative measures, such as dedicated core depressurization system (DCDS), passive autocatalytic recombiner (PAR) system, are somehow validated through the calculation; the results also indicated that with the power recovered at 6 hour to start up ECCS, the corium could be retained above support plate.

A Method for RCCA Drop Analysis

2013 ◽  
Author(s):  
Tong Liu ◽  
Heng Huang ◽  
Peng Li ◽  
Wei Xu ◽  
Yuemin Zhou ◽  
...  
Keyword(s):  
Computer Program ◽  
Equations Of Motion ◽  
Control Rod ◽  
Structure Interaction ◽  
Fluid Resistance ◽  
Fuel Assemblies ◽  
Control Assembly ◽  
Drop Time ◽  
The Impact ◽  
Fluid Interaction

The control rod drop time is one of the most important parameters for safety analysis. The calculation accuracy of control rod drop time will be affected if ignoring the transverse vibration, which is mainly caused by the fluid-structure interaction between the RCCA (rod cluster control assembly)s, guide tubes and the fluid interaction. A new method for RCCA drop analysis is presented in this paper. The transverse equations of motion are established considering fluid-elastic structure interaction, assuming the fluid is incompressible laminar flow. And the vertical equations of motion are established considering the gravity force of RCCAs, the fluid resistance, and the friction force led by collision. A computer program based on this method is used to calculate the control rod drop time, the impact and the dynamic response of RCCA. The analysis results are compared with those of the in-house code, which is used for commercial design. It shows the computer program using the new models provides a useful tool in the design of RCCAs and fuel assemblies.

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