On the Factors Affecting the Fretting-Wear Risk of PWR Fuel Assemblies

2006 ◽  
Author(s):  
Pablo R. Rubiolo ◽  
Michael Y. Young
Keyword(s):  
Fretting Wear ◽  
Structural Damping ◽  
Linear Vibration ◽  
Key Factors ◽  
Pressurized Water ◽  
Wear Performance ◽  
Factors Affecting ◽  
Vibration Model ◽  
Fuel Assemblies ◽  
Water Reactors

An analysis of the factors affecting the fretting-wear risk of fuel assemblies operated in Pressurized Water Reactors (PWR) is presented. In this work the effect of rod-to-grid gap size, spring relaxation, assembly grids misalignments, rod structural damping and stiffness, and friction coefficients are investigated by performing Monte Carlo simulations with a non-linear vibration model of the fuel rod. The goal of the study is to identify key factors that have to be included in the assessment of the fretting-wear performance of core fuel assemblies.

Fretting Wear of the Zr Fuel Rod Cladding

2014 ◽  
Author(s):  
Ladislav Pecinka ◽  
Jaroslav Svoboda ◽  
Vladimír Zeman
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactors ◽  
Fretting Wear ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Assemblies ◽  
Flow Induced Vibrations ◽  
Water Reactors ◽  
Fretting Damage

Fretting wear is a particular type of wear that is expected to occur in fuel assemblies of pressurized water cooled nuclear reactors. Fretting damage of fuel rods may cause Nuclear Power Plant (NPP) operations problems and are very expensive to repair. As utilities and fuel vendors adopt higher utilization of uranium and improved thermal margins plants, burned fuel rods will be loaded at core the periphery as part of the margin mechanisms. Pressurized Water Reactors (PWRs) have experienced fuel rods fretting wear failures due to flow induced vibrations. This study describes basic results of theoretical analysis and describes experiments to predict thinning of the Zr cladding wall thickness performed.

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

AP1000 Plant CILC/CIPS Risk Assessment Using Advanced TH Methodology

2012 ◽  
Author(s):  
G. Wang ◽  
P. Sapienza ◽  
R. J. Fetterman ◽  
M. Y. Young ◽  
J. R. Secker ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Localized Corrosion ◽  
Pressurized Water ◽  
Assembly Design ◽  
Power Shift ◽  
Fuel Assemblies ◽  
Water Reactors ◽  
The Relationship

Similar to many existing Pressurized Water Reactors (PWR), the AP1000® cores will undergo sub-cooled nucleate boiling in the upper grid spans of some fuel assemblies at normal operating conditions. Sub-cooled nucleate boiling may increase crud deposits on the fuel cladding surface which may increase the risk of Crud Induced Power Shift (CIPS) and/or Crud Induced Localized Corrosion (CILC). A CIPS/CILC risk assessment has been performed to support the AP1000 fuel assembly design finalization. In this paper, the advanced thermal-hydraulic (TH) methodology used in the AP1000 plant CIPS/CILC risk assessments are summarized and discussed, and the relationship between the CIPS/CILC mechanisms, fuel reliability, and plant operating conditions is also presented. Finally, acceptable AP1000 core CIPS/CILC risk assessment results are summarized and suggestions that specifically target reducing CIPS/CILC risks for AP1000 plants are described.

An Introduction to RCC-C: An AFCEN Code for Design and Construction of PWR Fuel Assemblies

2013 ◽  
Author(s):  
Christine Vauglin ◽  
Marc Ton-That ◽  
Morello Sperandio ◽  
Denis Buisine
Keyword(s):  
French Society ◽  
Pressurized Water ◽  
Technical Content ◽  
Fuel Assemblies ◽  
Core Components ◽  
Mechanical Components ◽  
Water Reactors ◽  
Design Characteristics ◽  
Service Inspection ◽  
Design Construction

AFCEN is the French society which publishes codes for design, construction and in-service inspection rules for Pressurized Water Reactors. The fields covered by theses codes are: mechanical components, in-service surveillance of mechanical components, electrical equipments, nuclear fuel, and more recently, civil works and fire protection. After a brief global presentation of AFCEN history and current position, we set out the technical content of one of these codes: RCC-C. This code is dedicated to fuel assemblies and associated core components. It sets forth minimum generic requirements to be met by the supplier and by the manufacturer for the design justifications and for the manufacturing and inspection operations of PWR fuel assemblies and rod cluster control assemblies. We go over the different chapters of RCC-C: product and part characteristics; manufacturing and inspection processes and methods and associated qualifications; inspection requirements for the different items; design characteristics and requirements which the design justification shall meet; methods for demonstrating that design requirements are met.

Monte Carlo Simulation of Fretting Wear Performance of Fuel Rods

2005 ◽  
Author(s):  
Pablo R. Rubiolo
Keyword(s):  
Monte Carlo ◽  
Fretting Wear ◽  
Fuel Rods ◽  
Wear Performance ◽  
Wear Rates ◽  
Fuel Rod ◽  
Cell Clearance ◽  
Vibration Model ◽  
Support Conditions ◽  
Wear Damage

The effect of the diverse parameters affecting the fretting-wear performance of nuclear fuel rods is investigated by performing Monte Carlo simulations with a fuel rod vibration model. The study is focused on the analysis of the effect of the grid parameters, including the cell clearance and the grid/support misalignments, on the support preload forces distribution, the rod dynamic response and the overall wear damage. In the present approach, the fuel rod and grids are modeled as a beam constrained at a finite number of axial positions and a non-linear vibration model is used to predict the rod motion and the wear rates. The results of the analysis suggest that an important fraction of the variability of the assembly wear damage distribution can be explained by the local variations of the rod-support conditions.

scholarly journals Burnable Poison Selection and Neutronics Analysis of Plate Fuel Assemblies

2021 ◽  
Vol 9 ◽  
Author(s):  
Shikun Xu ◽  
Tao Yu ◽  
Jinsen Xie ◽  
Lei Yao ◽  
Zhulun Li
Keyword(s):  
Fuel Assembly ◽  
Critical Role ◽  
Pressurized Water Reactors ◽  
Effective Control ◽  
Pressurized Water ◽  
Burnable Poison ◽  
Fuel Assemblies ◽  
Long Life ◽  
Water Reactors ◽  
Selection Of

Burnable poisons play a critical role in long-life pressurized water reactors. Plate fuel elements have good application prospects in long-life pressurized water reactors. In long-life pressurized water reactors with large initial residual reactivity in the core, a reasonable selection of burnable poisons can suppress the large residual reactivity at beginning of lifetime and can achieve a long burnup depth at end of lifetime. Therefore, the selection of burnable poisons is a crucial factor to be considered in the design of long-life pressurized water reactors. In this study, the selection of burnable poisons and neutronics characteristics of long-life PWR plate fuel assembly were studied. The transport-burnup calculations of different burnable poison fuel assemblies were carried out. Some candidate BPs are selected to realize the effective control of reactivity. The results show that when the enriched isotopes 157Gd, 167Er and B4C are used as burnable poisons, there is almost no reactivity penalty; when PACS-J and 231Pa are used as burnable poisons, due to their own characteristics, not only does not cause reactivity penalty at end of lifetime, but also the fuel assembly life is extended, the fuel utilization rate is improved. The combination of PACS-J and the slow-burnup burnable poisons can obtain a better reactivity curve. The results of this article show that the plate fuel assemblies can be selected with enriched isotope 157Gd, enriched isotope 167Er, B4C, 231Pa and PACS-J as burnable poisons, and the combinations of burnable poisons can be selected with two combination schemes, PACS-Er and PACS-Pa.

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