scholarly journals Post-test Analyses of the CMMR-4 Test

2021 ◽  
Author(s):  
Takuya Yamashita ◽  
Hiroshi Madokoro ◽  
Ikken Sato
Keyword(s):  
Nuclear Power ◽  
Gas Permeability ◽  
Plasma Heating ◽  
Heating System ◽  
Boiling Water Reactors ◽  
Core Material ◽  
The Core ◽  
Post Test ◽  
Ceramic Fuel ◽  
Water Reactors

Abstract Understanding the final distribution of core materials and their characteristics is important for decommissioning the Fukushima Daiichi Nuclear Power Station (1F). Such characteristics depend on the accident progression in each unit. However, BWR accident progression involves great uncertainty. This uncertainty, which was clarified by MAAP-MELCOR Crosswalk, cannot be resolved with existing knowledge and was thus addressed in this work through core material melting and relocation (CMMR) tests. For the test bundle, ZrO2 pellets were installed instead of UO2 pellets. A plasma heating system was used for the tests. In the CMMR-4 test, useful information was obtained on the core state just before slumping. The presence of macroscopic gas permeability of the core approaching ceramic fuel melting was confirmed, and the fuel columns remained standing, suggesting that the collapse of fuel columns, which is likely in the reactor condition, would not allow effective relocation of the hottest fuel away from the bottom of the core. This information will help us comprehend core degradation in boiling water reactors, similar to those in 1F. In addition, useful information on abrasive water suspension jet (AWSJ) cutting for debris-containing boride was obtained in the process of dismantling the test bundle. When the mixing debris that contains oxide, metal, and boride material is cut, AWSJ may be repelled by the boride in the debris, which may cut unexpected parts, thus generating a large amount of waste in cutting the boride part in the targeted debris. This information will help the decommissioning of 1F.

Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

2008 ◽  
Author(s):  
James W. Morgan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Control ◽  
Nuclear Power ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Pump Speed ◽  
Water Reactors ◽  
Generator Sets

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

Non-Linear BWR Stability Analysis

2008 ◽  
Author(s):  
F. Wehle ◽  
A. Schmidt ◽  
S. Opel ◽  
R. Velten
Keyword(s):  
Stability Analysis ◽  
Nuclear Power ◽  
Reverse Flow ◽  
High Amplitude ◽  
Boiling Water Reactors ◽  
Linear Regime ◽  
Power Oscillations ◽  
Non Linear ◽  
Water Reactors ◽  
Physical Reasons

Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants already during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge oN BWR instabilities. The appropriate representation of the physical processes in the non-linear regime requires typically time domain stability analysis. The objective of this paper is to present a physical model, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. The application of this model gives a deeper insight into the physical reasons for the prevention of the uncontrolled divergence of BWR oscillations. The mechanisms that have a stabilizing effect are demonstrated.

Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injector: Part 14—Cooling Characteristics of the Particulate Core Material Debris Accumulated in Horizontal Narrow Gap

2006 ◽  
Author(s):  
Y. Hirao ◽  
G. Su ◽  
K. Sugiyama ◽  
T. Narabayashi ◽  
M. Mori ◽  
...  
Keyword(s):  
Heat Supply ◽  
Nuclear Power ◽  
High Efficiency ◽  
Hot Spot ◽  
Research Work ◽  
Lower Surface ◽  
Injection System ◽  
Core Material ◽  
Metal Debris ◽  
The Core

When LOCA occurs in proposed nuclear reactor systems, the coolability of the core would be kept by the SI core injection system and therefore the probability of the core meltdown is negligible small. In this research work, we make it clear that the coolability of the RPV bottom is secured even if a part of the core should melt and a substantial amount of debris should be deposited on the lower plenum. In this report, we examined experimentally the coolability of the RPV bottom that a Zircaloy-based loose debris layer is deposited on. We set up a heat supply section made by SUS304 on the loose debris layer and measured the heat flux released into the loose debris bed and the temperature at the lower surface of the heat supply section. In addition, we measured the temperature distribution at the bottom of the loose debris bed. It became clear in this study that the coolability depends on the amount of coolant supplied, and the hot spot would not occur when coolant is supplied. Even if a hotspot should occur in the oxidization of loose metal debris accompanied with rapid heat generation. It is found that when a small amount of coolant can be supplied, it disappears because of a high capillary force of oxidized loose debris. So it is confirmed that the soundness of RPV is basically maintained.

scholarly journals Non-Linear Stability Analysis of Real Signals from Nuclear Power Plants (Boiling Water Reactors) based on Noise Assisted Empirical Mode Decomposition Variants and the Shannon Entropy

2017 ◽  
Author(s):  
Omar A. Olvera-Guerrero ◽  
Alfonso Prieto-Guerrero ◽  
Gilberto Espinosa-Paredes
Keyword(s):  
Empirical Mode Decomposition ◽  
Shannon Entropy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Mode Decomposition ◽  
Non Linear ◽  
Water Reactors

There are currently around 78 Nuclear Power Plants (NPP) in the world based on Boiling Water Reactors (BWR). The current parameter to assess BWR instability issues is the linear Decay Ratio (DR). However, it is well known that BWRs are complex non-linear dynamical systems that may even exhibit chaotic dynamics that normally preclude the use of the DR when the BWR is working at a specific operating point during instability. In this work a novel methodology based on an adaptive Shannon Entropy estimator and on Noise Assisted Empirical Mode Decomposition variants is presented. This methodology was developed for real-time implementation of a stability monitor. This methodology was applied to a set of signals stemming from several NPPs reactors (Ringhals-Sweden, Forsmark-Sweden and Laguna Verde-Mexico) under commercial operating conditions, that experienced instabilities events, each one of a different nature

Guideline of Weld Overlay Repair Method for Primary Piping of Japanese Boiling Water Reactors

2006 ◽  
Author(s):  
Haruo Fujimori ◽  
Eisaku Hayashi ◽  
Yuichi Motora ◽  
Hitoshi Ohata ◽  
Hideo Sakurai
Keyword(s):  
Nuclear Power ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Piping System ◽  
Weld Overlay ◽  
Repair Method ◽  
Post Process ◽  
Engineering Society ◽  
Water Reactors ◽  
Service Inspection

A private guideline of a weld overlay (WOL) repair method for SCC cracked Boiling Water Rector (BWR) primary piping was provided by Thermal and Nuclear Power Engineering Society of Japan. In the WOL method, the overlaid structure of SCC resistant 308L weld metal substitutes for cracked existing piping weldments as pressure boundary. The guideline prescribes the applicability, welding conditions and essential variables, weld overlay structural design, evaluation of welding effects to the piping system, in-process control, post-process examination, and procedures of pre-service and in-service inspection.

Sign in / Sign up

Export Citation Format

Share Document