Detailed CFD Analysis of Coolant Mixing in VVER-440 Fuel Assembly Heads Performed With the Code CFX-5.5

2004 ◽  
Author(s):  
Ga´bor Le´gra´di ◽  
Attila Aszo´di
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Power Plants ◽  
Sensitivity Study ◽  
Temperature Fields ◽  
Head Model ◽  
Detailed Knowledge ◽  
Mixing Processes ◽  
Significant Information ◽  
Fuel Assemblies

3D modeling of the thermal hydraulical processes in a fuel assembly head means a great challenge for the CFD technique due to the complexity of its structure and the flow domain. On the other hand, this field is of great importance since detailed knowledge on mixing processes in the assembly heads and calculations on the signals of the thermocouples positioned just above the heads would give very significant information for the safety analyses connected to the power upgrading of nuclear power plants. Therefore development of a complex fuel assembly model was started in the Institute of Nuclear Techniques of the Budapest University of Technology and Economics in the near past. In this paper, the fuel assembly head model, the sensitivity study of it, calculations and results are presented. The main goal of our work is investigating the signal of the thermocouples which are placed just above the fuel assemblies. The calculations were performed with consideration of four kinds of different fuel assemblies. The inlet velocity and temperature fields were calculated by the COBRA subchannel code of the Paks Nuclear Power Plant of Hungary. With all kind of fuel assemblies, calculations were performed with assumptions of normal symmetrical and highly asymmetrical heat source profiles of inner assemblies and assemblies positioned beside absorber elements.

Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique

2018 ◽  
Author(s):  
Xing Li ◽  
Sichao Tan ◽  
Zhengpeng Mi ◽  
Peiyao Qi ◽  
Yunlong Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Great Influence ◽  
Reactor Core ◽  
Transfer Characteristic ◽  
Index Of Refraction ◽  
Temperature Fields ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Combined Use

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.

scholarly journals Temperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode

2021 ◽  
Vol 7 (1) ◽  
pp. 9-13
Author(s):  
David A. Hakobyan ◽  
Victor I. Slobodchuk
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Cooling Mode ◽  
Temperature Conditions ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK reactors have not been fully resolved so far. For this reason, nuclear power plants are forced to search for new options for the disposal of spent fuel, which can provide at least temporary SNF storage. One of the possible solutions to this problem is to switch to compacted SNF storage in reactor spent fuel pools (SFPs). As the number of spent fuel assemblies (SFAs) in SFPs increases, a greater amount of heat is released. In addition, no less important is the fact that a place for emergency FA discharging should be provided in SFPs. The paper presents the results of a numerical simulation of the temperature conditions in SFPs both for compacted SNF storage and for emergency FA discharging. Several types of disturbances in normal SFP cooling mode are considered, including partial loss of cooling water and exposure of SFAs. The simulation was performed using the ANSYS CFX software tool. Estimates were made of the time for heating water to the boiling point, as well as the time for heating the cladding of the fuel elements to a temperature of 650 °С. The most critical conditions are observed in the emergency FA discharging compartment. The results obtained make it possible to estimate the time that the personnel have to restore normal cooling mode of the spent fuel pool until the maximum temperature for water and spent fuel assemblies is reached.

scholarly journals A modern data measurement system to study and test thermionic heat to electricity converters

2018 ◽  
Vol 4 (3) ◽  
pp. 179-183
Author(s):  
Andrey Kirillov ◽  
Valeriy Yarygin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Test System ◽  
Temperature Fields ◽  
Test Facility ◽  
Measurement Systems ◽  
Data Measurement ◽  
Functional Capabilities ◽  
Thermionic Nuclear Power ◽  
A Current

Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS jointly with the measured test section and experimental test facility temperature fields. The structure and the characteristics of a DMS based on products from ICP DAS Co., Ltd are presented. A developed VAC measurement program providing the operator with a convenient graphic interface and enabling adjustment of the measurement parameters has been considered. The VAC recording errors in the process of measurements have been determined using TIC simulators. The error in the VAC diffusion portion on a simulator (with a current of less than 3 A) is not more than 1%. Thanks to the use of modern components, the developed DMS offers extended functional capabilities for measuring the thermocouple signals in an experimental electrophysical test facility. The DMS structure provides for the convenience of scaling (through a larger number of measuring channels) and makes it possible to add modules from other manufacturers. The experience of operating this DMS will be used to develop the DMS for an in-pile test system designed for similar functions.

scholarly journals Approaches to Safety Justification for Loading of VSC-VVER Containers in ZNPP DSFSF

2019 ◽  
pp. 82-87
Author(s):  
Ya. Kostiushko ◽  
O. Dudka ◽  
Yu. Kovbasenko ◽  
A. Shepitchak
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Spent Fuel ◽  
Fuel Assemblies ◽  
Fuel Storage ◽  
The Government

The introduction of new fuel for nuclear power plants in Ukraine is related to obtaining a relevant license from the regulatory authority for nuclear and radiation safety of Ukraine. The same approach is used for spent nuclear fuel (SNF) management system. The dry spent fuel storage facility (DSFSF) is the first nuclear facility created for intermediate dry storage of SNF in Ukraine. According to the design based on dry ventilated container storage technology by Sierra Nuclear Corporation and Duke Engineering and Services, ventilated storage containers (VSC-VVER) filled with SNF of VVER-1000 are used, which are located on a special open concrete site. Containers VSC-VVER are modernized VSC-24 containers customized for hexagonal VVER-1000 spent fuel assemblies. The storage safety assessment methodology was created and improved directly during the licensing process. In addition, in accordance with the Energy Strategy of Ukraine up to 2035, one of the key task is the further diversification of nuclear fuel suppliers. Within the framework of the Executive Agreement between the Government of Ukraine and the U.S. Government, activities have been underway since 2000 on the introduction of Westinghouse fuel. The purpose of this project is to develop, supply and qualify alternative nuclear fuel compatible with fuel produced in Russia for Ukrainian NPPs. In addition, a supplementary approach to safety analysis report is being developed to justify feasibility of loading new fuel into the DSFSF containers. The stated results should demonstrate the fulfillment of design criteria under normal operating conditions, abnormal conditions and design-basis accidents of DSFSF components.  Thus, the paper highlights both the main problems of DSFSF licensing and obtaining permission for placing new fuel types in DSFSF.

scholarly journals ANALYTICAL SIMULATION OF A PWR USING MELCOR

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
Maritza Rodríguez Gual ◽  
Nathalia N. Araújo ◽  
Marcos C. Maturana
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Normal Operation ◽  
Detailed Knowledge ◽  
Pressurized Water ◽  
Severe Accidents ◽  
Reference Plant

After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

Neutronic Analysis for VVER-440 Type Reactor Using PARCS Code

2018 ◽  
Author(s):  
Luca Ratti ◽  
Guido Mazzini ◽  
Marek Ruščák ◽  
Valerio Giusti
Keyword(s):  
Czech Republic ◽  
Cross Sections ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Models ◽  
Reactor Core ◽  
Sensitivity Study ◽  
The Czech Republic ◽  
State Office ◽  
Scale Models

The Czech Republic National Radiation Protection Institute (SURO) provides technical support to the Czech Republic State Office for Nuclear Safety, providing safety analysis and reviewing of the technical documentations for Nuclear Power Plants (NPPs). For this reason, several computational models created in SURO were prepared using different codes as tools to simulate and investigate the design base and beyond design base accidents scenarios. This paper focuses on the creation of SCALE and PARCS neutronic models for a proper analysis of the VVER-440 reactor analysis. In particular, SCALE models of the VVER-440 fuel assemblies have been created in order to produce collapsed and homogenized cross sections necessary for the study with PARCS of the whole VVER-440 reactor core. The sensitivity study of the suitable energy threshold to be adopted for the preparation with SCALE of collapsed two energy-group homogenized cross sections is also discussed. Finally, the results obtained with PARCS core model are compared with those reported in the VVER-440 Final Safety Report.

Flow-Induced Vibration Analysis of an Integral Economizer Once-Through Steam Generator

1989 ◽  
Vol 111 (4) ◽  
pp. 501-506
Author(s):  
M. K. Au-Yang ◽  
B. Brenneman
Keyword(s):  
Fuel Assembly ◽  
Steam Generator ◽  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
First Generation ◽  
Second Generation ◽  
Flow Induced Vibration ◽  
Operational Test

The integral economizer once-through steam generator is a second-generation steam generator used in B&W’s 205-fuel assembly nuclear power plants. Besides having an integral economizer, this steam generator differs from the first generation units, sixteen of which have been operating with B&W’s 177 fuel assembly nuclear power plants for more than ten years, in having a much higher flow rate. This higher flow rate induces a correspondingly higher fluid-dynamic load on all of the steam generator internal components, particularly the tube bundle. This paper describes the flow-induced vibration design analysis of this second-generation nuclear steam generator. The three most commonly known flow-induced vibration phenomena were considered: fluid-elastic instability, turbulence-induced vibration and vortex-induced vibration. To minimize uncertainties in the many experimentally determined input parameters such as damping ratios, Connors’ constant and the dynamic pressure power spectral densities, a parallel analysis was carried out on the operating first-generation steam generator, and the results compared. The analytical results were verified by the recent start-up of B&W’s first 205-fuel assembly nuclear plant. No vibration problems were encountered during either the pre-operational test or in several months of full power operations.

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