Main Evolutions of the RCC-C Design and Construction Code for Fuel Assemblies Since 2015

2018 ◽  
Author(s):  
Marc Ton-That ◽  
Christine Vauglin ◽  
Gilbert Trillon
Keyword(s):  
Nuclear Fuel ◽  
Electric Utility ◽  
Original Text ◽  
Nuclear Operator ◽  
Pressurized Water ◽  
Development Organization ◽  
Fuel Assemblies ◽  
Core Components ◽  
Mechanical Components ◽  
Water Reactors

AFCEN is a French Standard Development Organization 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, civil works and fire protection. AFCEN was initially founded by electric utility EDF and nuclear steam supply system manufacturer FRAMATOME. AFCEN has more than 60 institutional members, representing more than 650 experts who contribute to the development and continuous improvement of codes. The RCC-C code, which is dedicated to PWR fuel assemblies and associated core components, set forth generic requirements to be fulfilled by the suppliers and by the manufacturers for the design justifications and for the manufacturing and inspection operations of PWR fuel assemblies and rod cluster control assemblies. The RCC-C is intended to be used in the frame of contractual relations between a customer (nuclear operator) and a nuclear fuel supplier. The first edition was published in 1981. Over the years, many changes have been made to the original text but the structure hasn’t been much modified. Because of this, the text was becoming less coherent for the users and was lacking also minimal explanations. A redesign of the code was scheduled for the 2015 edition to address those problems. With the involvement of fuel vendors FRAMATOME, WESTINGHOUSE, and French nuclear operator EDF, the text was restructured and clarified. New requirements were implemented and the set of both design and manufacturing rules was strengthened to reflect fuel vendors’ practices and operator expectations. This article explains the main modifications that were implemented since the 2015 edition, and also outlines the prospects for future changes taking into account the latest regulatory requirements and evolutions of the industrial practices.

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.

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.

Identification of Non-Linear Vibration Parameters of a Nuclear Fuel Rod

2018 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Kostas Karazis ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Harmonic Balance Method ◽  
Resonance Phenomenon ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Non Linear

In Pressurized Water Reactors (PWR), fuel assemblies are composed of fuel rods, long slender tubes filled with uranium pellets, bundled together using spacer grids. These structures are subjected to fluid-structure interactions, due to the flowing coolant surrounding the fuel assemblies inside the core, coupled with large-amplitude vibrations in case of external seismic excitation. Therefore, understanding the non-linear response of the structure and, particularly, its dissipation, is of paramount importance for the choice of safety margins. To model the nonlinear dynamic response of fuel rods, the identification of nonlinear stiffness and damping parameters is required. The case of a single fuel rod with clamped-clamped boundary conditions was investigated by applying harmonic excitation at various force levels. Different configurations were implemented testing the fuel rod in air and in still water; the effect of metal pellets simulating nuclear fuel pellets inside the rods was also recorded. Non-linear parameters were extracted from some of the experimental response curves by means of a numerical tool based on the harmonic balance method. The axisymmetric geometry of fuel rods resulted in the presence of a one-to-one internal resonance phenomenon, which has to be taken into account modifying accordingly the numerical identification tool. The internal motion of fuel pellets is a cause of friction and impacts, complicating further the linear and non-linear dynamic behavior of the system. An increase of the equivalent viscous-based modal damping with excitation amplitude is often shown during geometrically non-linear vibrations, thus confirming previous experimental findings in the literature.

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.

Comparison of Nuclear Fitness for Service Codes: ASME Section XI and RSE-M AFCEN Codes

2014 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactors ◽  
Pressurized Water ◽  
Term Operation ◽  
Mechanical Components ◽  
Water Reactors ◽  
Test Requirements ◽  
Service Inspection

Two major Codes are used for Fitness for Service of Nuclear Power Plants: one is the ASME B&PV Code Section XI and the other one is the French RSE-M Code. Both of them are largely used in many countries, partially or totally. The last 2013 RSE-M covers “Mechanical Components of Pressurized Water Reactors (PWRs): - Pre-service and In-service inspection - Surveillance in operation or during shutdown - Flaw evaluation - Repairs-Replacements parts for plant in operation - Pressure tests The last 2013 ASME Section XI covers “Mechanical components and containment of Light Water Reactors (LWRs)” and has a larger scope with similar topics: more types of plants (PWR and Boiling Water Reactor-BWR), other components like metallic and concrete containments… The paper is a first comparison covering the scope, the jurisdiction, the general organization of each section, the major principles to develop In Service Inspection, Repair-Replacement activities, the flaw evaluation rules, the pressure test requirements, the surveillance procedures (monitoring…) and the connections with Design Codes… These Codes are extremely important for In-service inspection programs in particular and essential tools to justify long term operation of Nuclear Power Plants.

Experiments on Nonlinear Vibrations of a Nuclear Fuel Rod Supported by Spacer Grids in Air and Submerged in Water

2019 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Giulio M. Franchini ◽  
Francesco Giovanniello ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Vibration Response ◽  
Nonlinear Phenomena ◽  
Internal Resonances ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Stiffness And Damping

Abstract For safety reasons, the nuclear fuel assemblies of Pressurized Water Reactors (PWR) must be able to withstand external excitations ranging from large amplitude seismic motions of the reactor to flow-induced vibrations from the surrounding coolant water. A nuclear fuel assembly is composed of long slender tubes, most of them filled with uranium pellets, maintained in a square array by spacer grids. The spacer grids provide a nonlinear flexible boundary condition with friction and micro-impacts that complicates the nonlinear dynamics. In order to improve safety margins in the design of nuclear fuel assemblies, it is of great interest to understand the influence of the spacer grids, as it relates to the overall structural stiffness and damping properties. In particular, the evolution of the vibration amplitude with increasing excitation forces is still undetermined. In order to understand the nonlinear vibration response of a zirconium fuel rod filled with nuclear fuel pellets and supported by spacer grids, experiments were carried out in water and in air. They consisted of measuring the vibration response of the rod under a step-sine harmonic excitation at different force amplitude levels in the frequency neighborhood of the fundamental mode. If the excitation is large enough, the response of the rod displays nonlinear phenomena such as the shift of the resonant frequencies, multiple solutions with instabilities (jumps) and hysteresis, and one-to-one internal resonances. These experiments were carried out on zirconium tubes filled with axially unconstrained as well as axially blocked metallic pellets, which simulate the nuclear fuel. The zirconium tubes were tested both in air and immersed in water. The experimental data will be processed in the future by means of an identification procedure to extract the nonlinear stiffness and damping parameters of the system. An increase of the equivalent viscous damping with the excitation amplitude level is expected.

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