The effect of fuel rod supporting conditions on fuel rod vibration characteristics and grid-to-rod fretting wear

The grid-to-rod fretting wear-induced fuel rod failure observed in PWRs may be caused by excessive fluid-induced vibration and inadequate fuel rod support by the spacer grid spring. In order to simulate in-reactor grid-to-rod fretting wear behaviors, the grid-to-rod fuel rod supporting conditions as a function of time were predicted by taking into account cladding creep rate, initial spacer grid spring deflection, spacer grid spring force relaxation, etc. Based on these grid-to-rod supporting conditions, the fuel rod vibration modes and natural frequencies were calculated with the help of the ANSYS code, while the fuel rod vibration amplitudes were estimated by the Paidoussis’ empirical formula. With these vibration characteristics that depend upon the grid-to-rod supporting conditions, the in-reactor fretting wear axial profile observed on the fuel rod surface are found to be simulated quite well. In addition, key design guidelines for the fuel assembly and the spacer grid are proposed to minimize the grid-to-rod fretting wear that may be utilized to develop an advanced fuel design against fretting wear.     

The effect of fuel rod loading speed on spacer grid spring force

The burnup-dependent grid-to-rod gap combined with the fluid-induced vibration may generate grid-to-rod fretting wear-induced fuel failure for some fuel assemblies in a certain burnup range. The grid-to-rod gap is dependent on initial spacer grid spring force, spring force relaxation and cladding creepdown. It is found that the initial spring force is reduced during the fuel rod loading into the fuel assembly skeleton. The extent of the initial spring force loss is strongly dependent on the fuel rod loading speed. Based on the initial spring force loss data obtained from two kinds of fuel rod loading speeds of 0.18 and 0.33 m/s, it can be said that the higher rod loading speed generates the larger initial spring force loss. This is because the higher speed generates the larger overshooting of spring deflection during the fuel rod loading. The extent of overshooting may be affected by axial misalignment of SG cells, spring-to-fuel rod end plug contact angle, ballooning of FR end plug weld region and the extent of gravity-induced FR bowing, combining with the fuel rod loading speed. The rod loading speed of 0.33 m/s is found to produce some spacer grid cells less than a minimum initial spring force requirement of 12 N against the grid-to-rod fretting wear-induced failure. In order to produce initial spacer grid spring force meeting the minimum spring force requirement, it is recommended that the lower rod loading speed be used, combined with axially aligned spacer grid cells and lower contact angle of spring-to-fuel rod end plug.     

Development of an advanced PWR fuel for OPR1000s in Korea

The advanced PWR fuel for the OPR1000s in Korea, PLUS7, has been developed to enhance thermal performance, high burnup capability and fuel reliability against grid-to-rod fretting wear and debris. The outstanding design features of PLUS7 include mixing vane mid-grids for increasing thermal performance and minimizing vibration-induced fretting wear, optimized fuel dimensions and advanced zirconium alloys for high burnup capability of 72,000 MWD/MTU, and an optimized fuel rod diameter for reducing pressure drop and improving neutron economy. The fuel assembly and its components performances have been verified through a wide spectrum of mechanical, thermal hydraulic, vibration and fretting wear tests. Based on the verification test results and the evaluations with the help of the KNF design code system, it is found that the PLUS7 fuel will maintain its integrity up to the envisaged burnup of 72,000 MWD/MTU. In addition, the PLUS7 fuel performances were evaluated to be considerably improved in comparison with the current fuel used in the OPR1000s.     

The study on grid-to-rod fretting wear models for PWR fuel

The fretting wear is found to be generated at grid-to-rod contact areas by flow-induced vibration. This flow-induced grid-to-rod fretting wear may be initiated at a certain critical grid-to-rod gap that strongly depends on the extent of flow-induced vibration and grid spring designs. Three fretting wear excitation mechanisms acting on the grid-to-rod fretting wear are summarized. In order to examine the impact of grid spring designs on the fretting wear rate, the fretting wear tests for three kinds of grid spring designs were carried out for 500 h, simulating the reactor flow conditions. In parallel, three grid-to-rod fretting wear models that include constant work rate model, constant work density rate model and linear work density rate model have been developed. The three fretting wear models were used to predict the fuel rod perforation times with the use of the fretting wear test results. It is said that the constant work density rate model or the linear work density rate model is quite effective in predicting the grid-to-rod fretting-induced rod failure time observed in commercial nuclear power plants.     

A review on the clad failure studies

In this paper, an attempt has been made to systematically organize the research investigations conducted on clad tube failure, so far. Before presenting the review on the clad failure studies, an introduction to different clad materials has been added, in which the effect of alloying elements on the material properties have been presented. The literature on clad failure has been broadly categorized under the headings LOCA and RIA. The failure mechanisms like creep, corrosion and pellet-clad interaction have been discussed in details. Each subsection of the review has been provided with summary table, in which the studies are arranged in the chronological order. A small section on acceptance criteria for ECCS has also been included. The last section of the review has been dedicated to the core-degradation phenomena.     

Enhancing TRU burning and Am transmutation in Advanced Recycling Reactor

This paper presents about conceptual designs of Advanced Recycling Reactor (ARR) focusing on enhancement in transuranics (TRU) burning and americium (Am) transmutation. The design has been conducted in the context of the Global Nuclear Energy Partnership (GNEP) seeking to close nuclear fuel cycle in ways that reduce proliferation risks, reduce the nuclear waste in the US and further improve global energy security. This study strives to enhance the TRU burning and the Am transmutation, assuming the development of related technologies in this study, while the ARR based on mature technologies was designed in the previous study. It has followed that the provided TRU burning core is designed to burn TRU at 28 kg/TW_thh, by adding moderator pins of B₄C (Enriched B-11) and the Am transmutation core will be able to transmute Am at 34 kg/TW_thh, by locating Am blanket of AmN around the TRU burning core. It indicates that these concepts improve TRU burning by 40-50% than the previous core and can transmute Am effectively, keeping the void reactivity acceptable.     

Capabilities of TRANSURANUS code in simulating power ramp tests from the IFPE database

Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. OECD NEA sets up the “International Fuel Performance Experiments (IFPE) database”, a public domain database on nuclear fuel performance experiments with the purpose of model development and code validation. The objective of the activity (performed in the framework of the IAEA CRP FUMEX-III project) is to investigate the pellet-clad interaction mechanism and the capability of TRANSURANUS code in simulating the phenomena, processes occurring in the fuel rod during the power ramps, with focus on the parameters influencing the cladding failures. The experimental database adopted is the Studsvik PWR Super-Ramp subprogram, part of the IFPE database, which consists of 28 pressurized water reactor fuel rods power ramped at burnup from 28 to 45 MWd/kgU. Relevant results by TRANSURANUS are presented in connection with the experimental evidences. Focus is given on the PCI/SCC failure, demonstrating that the failure threshold, available in TRANSURANUS, results conservative both in case of KWU and W rods.     

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.