Fuel Rod Failure Research Articles

Numerical Simulation Analysis of Debris Filtration in the Bottom Nozzle

According to the IAEA report, debris fretting is the second largest factor causing fuel rod cladding rupture and causing fuel rod failure, accounting for 11%. As the first structural component for fluid entering the fuel assembly, the bottom nozzle has the function of filtering debris entering the core area, effectively reducing the possibility of debris entering the fuel area causing fretting. Currently, there is no suitable numerical simulation method for calculating it. Therefore, based on the coupling method of discrete element method and computational fluid dynamics, considering the stress and contact models of typical debris in the complex flow field near the bottom nozzle under different structures, this study established the fluid solid coupling numerical simulation method between the solid debris and the flow field near the bottom nozzle. Compared with the test results, the error is less than 10%, which verifies the correctness and reliability of this method, And based on this method, typical debris filtering behavior analysis was conducted. The numerical simulation method studied in this article provides support for the design and optimization of the filtration structure of the fuel assembly's bottom nozzle for debris filtration efficiency.

Proposal and Validation of a Diagnosis Method of Fuel Failures in Pressurized Water Reactor Nuclear Power Plants

Based on research into the diagnosis methods for fuel failures in pressurized water reactor nuclear power plants and an analysis of existing problems, this paper proposes an enhanced fuel failure diagnosis method, addressing three key improvements: identification of the occurrence of fuel failure, diagnosis of the number of failed fuel rods, and the size of fuel rod failure. Other factors that may influence the diagnosis results are also identified and discussed. The proposed method is then verified using operational data from in‐service nuclear power plants in China. The verification results demonstrate that the improved diagnosis method accurately identifies fuel failures and exhibits broader applicability.

LOCA advanced failure criteria: Implementation in the mechanistic fuel performance code and practical applications

The advanced Chung-Kassner criteria for the capability of fuel rods to withstand thermal shock during LOCA reflooding and to withstand post-quench fuel handling, as well as the complementary cladding embrittlement criteria of KAERI and KIT, are compared to the traditional LOCA safety criterion of 17% ECR oxidation. The advanced failure criteria were implemented in the mechanistic thermomechanical code SVECHA/QUENCH (S/Q) and successfully applied to predict fuel rod failure behaviour in postulated LOCA scenarios and high temperature (1400 °C) pre-oxidation and quenching tests performed with fresh and irradiated fuel rod simulators. The extension of the advanced failure criteria to coated Zr alloy claddings is discussed.

Study on In-Service Inspection of Nuclear Fuel Assembly Failure Using Ultrasonic Plate Wave.

As protection for nuclear power plants is quite necessary, the nuclear fuel is sealed in zirconium alloy thin wall cladding. During service, fuel rods might be damaged caused by wall-thickness thinning, cladding corrosion and cracking, etc. This will cause the coolant to enter into the fuel rod, which may lead to the failure of the fuel assembly. However, current diagnostic methods have limitations due to the special structure of the fuel assembly and the underwater and radioactive environment. In this paper, a novel inspection method is proposed to recognize the failure of a fuel rod. The fuel rod failure can be detected based on the presence or absence of coolant inside the fuel rod by using an ultrasonic plate wave. The inspection model and process algorithm are proposed for in-service inspection. The relationship between signal and scanning position is established and analyzed. Both ultrasound field simulation and experiment have been carried out for validation. The corresponding results illustrate that the failed nuclear fuel rod of the whole fuel assembly (including the internal rods) can be effectively detected without the influence of the near-field region by using the proposed method.

Nuclear Fuel Modelling During Power Ramp

Fuel rods operating for several years in a LWR can experience fuel-cladding gap closure as a result of the phenomena due to temperature and irradiation. Local power increase induces circumferential stresses in the cladding as a result of the different expansion in the cladding and the pellet. In presence of corrosive fission products (i.e. Iodine) and beyond specific stress threshold and level of burnup, cracks may grow-up from the internal to the external cladding surface, causing fuel rod failure. The phenomenon, known as pellet cladding interaction-stress corrosion cracking PCI/SCC, or PCI, has been identified as a problem since the 70's. The PWR Super-Ramp experiment (part of OECD/NEA “International Fuel Performance Experiments (IFPE) database”) twenty eight fuel rods behaviour has been simulated using TRANSURANUS code version “v1m1j11”. Two sets (“Reference” and “Improved”) of suitable input decks modelling the fuel rods, based on the available literature are used to run the simulations. Focus is given to the main phenomena which are involved or may influence the cladding failure. Systematic comparison of the code results with the experimental data are performed for the parameters relevant for the PCI phenomenon. Sensitivity calculations on fission gas release models implemented in TRANSURANUS code are also performed in order to address the impact on the results. The results show the ability of TRANSURANUS version “v1m1j11” in conservatively predicting the rods failure due to PCI in PWR fuel and Zircaloy-4 cladding. Increased availability of experimental data would help to perform a deeper analysis.

Zy-4 LOCA cladding burst criteria computed by neural networks

This article presents new Zy-4 cladding burst criteria based on neural networks with up to five input parameters: the temperature, the engineering hoop stress, the circumferential strain, the heat-up rate and the heating mode. A large database composed of more than one thousand Zy-4 Loss Of Coolant Accident (LOCA) burst conditions has been built mainly with data from literature. The database is then used to train a series of neural networks. Many topics are investigated such as numerical technique choices for classification, input parameters normalization, training strategy or regularization. A specific evaluation function was developed to be able to compare network’s accuracy whatever the input parameter number is. First, a temperature-engineering stress model is obtained, adding the heating-rate as an input parameter strongly improves model predictions whereas adding burst strain has almost no impact. Finally, risk-informed neural networks are constructed to address various objectives like preventing from fuel rod failure (radiological consequences) or to envelop flow blockage (core reflooding). In the future, these neural networks will be implemented and used in the IRSN DRACCAR LOCA simulation code and further work on uncertainties will be performed.

Material Interactions in Severe Accidents – Benchmarking the MELCOR V2.2 Eutectics Model for a BWR-3 Mark-I Station Blackout: Part II – Uncertainty Analysis

Single case comparisons between severe accident simulations can provide detailed insights into severe accident model behavior, however, they cannot offer insights into model uncertainty, sensitivity to uncertain parameters, or underlying model biases. In this analysis, the single case benchmark comparison of the MELCOR material interaction models for a station blackout (SBO) scenario of a boiling water reactor (BWR) using representative Fukushima Daiichi Unit 1 boundary conditions is expanded to include an uncertainty analysis. As part of this uncertainty analysis, 1200 simulations are performed for each material interaction model (2400 total), with random sampling of 14 uncertain MELCOR input parameters. Input parameters are selected for their impact on models representing core degradation processes. These include candling, fuel rod failure, debris quenching and dryout. The analysis performed here is not a traditional “best-estimate” uncertainty analysis that uses best-estimate parameters or identifies best-estimate figure of merit distributions. Instead, it is an exploratory uncertainty analysis that identifies and interrogates underlying model form biases of the two material interaction models (eutectics and interactive materials models). Uniform distributions are applied to all uncertain parameters to ensure coverage of the model parameter uncertainty space. Key findings from this study include underlying model form biases exhibited by material interaction models, and notable differences in accident progression outcomes between the material interaction models. This uncertainty study extends and confirms the conclusions from the first part of this study, which compared the impact of material interaction modeling on simulation of a short-term station blackout scenario with representative Fukushima Daiichi Unit I boundary conditions. In particular, this study confirms that the eutectics model generally exhibits accelerated degradation and failure of fuel components, the core plate, and the lower head. The eutectics model also has a tendency to exhibit a greater degree of core degradation, greater debris mass formation, and larger debris mass ejection. Finally, the eutectics model exhibits higher maximum temperatures for fuel, cladding, particulate debris, oxidic molten pool, and metallic molten pool components than the interactive materials model; interactive materials model simulations exhibit a soft “limitation” on maximum temperatures that is related to the temperature at which material relocation occurs.

Material Interactions in Severe Accidents – Benchmarking the MELCOR V2.2 Eutectics Model for a BWR-3 MARK-I Station Blackout: Part I – Single Case Analysis

In this analysis, the two material interaction models available in the MELCOR code are benchmarked for a severe accident at a BWR under representative Fukushima Daiichi boundary conditions. This part of the benchmark investigates the impact of each material interaction model on accident progression through a detailed single case analysis. It is found that the eutectics model simulation exhibits more rapid accident progression for the duration of the accident. The slower accident progression exhibited by the interactive materials model simulation, however, allows for a greater degree of core material oxidation and hydrogen generation to occur, as well as elevated core temperatures during the ex-vessel accident phase. The eutectics model simulation exhibits more significant degradation of core components during the late in-vessel accident phase – more debris forms and relocates to the lower plenum before lower head failure. The larger debris bed observed in the eutectics model simulation also reaches higher temperatures, presenting a more significant thermal challenge to the lower head until its failure. At the end of the simulated accident scenario, however, core damage is comparable between both simulations due to significant core degradation that occurs during the ex-vessel phase in the interactive materials model simulation. A key difference between the two models’ performance is the maximum temperatures that can be reached in the core and therefore the maximum ΔT between any two components. When implementing the interactive materials model, users have the option to modify the liquefaction temperature of the ZrO2-interactive and UO2-interactive materials as a way to mimic early fuel rod failure due to material interactions. Through modification of the liquefaction of high melting point materials with significant mass, users may inadvertently limit maximum core temperatures for fuel, cladding, and debris components.

Accident tolerant fuel rod failure under low stress: A case study of BWR under station blackout using Bison

Accident tolerant fuels (ATFs) such as FeCrAl cladding have been designed to have improved performance in normal operation and provide increased coping time during design basis accidents (DBA) and beyond DBA. A key aspect in the evaluation of ATF designs is to determine potential increases in the coping time with respect to advancements in fuel and cladding materials. The station blackout (SBO) event has been widely recognized as one of the most severe conditions that a nuclear power plant can experience. In this comparative study, the fuel performance code Bison is utilized to determine the coping time of FeCrAl and Zircaloy claddings under a Short-Term SBO (ST-SBO) accident without any mitigation measures. In particular, four fuel rod burst criteria for Zircaloy cladding and two for FeCrAl cladding are used. It is found that none of the cladding burst criteria are fulfilled, and hoop stress is quite low in this BWR SBO design for both claddings, a stark difference from LOCA or PWR SBO events where high stress burst is common and has been demonstrated widely. Specifically, a new cladding failure mechanism for high temperature and low stress is identified in this study. The origin of the low stress at burst is also investigated.

Analysis of the upflow conversion modification and influence on selected LOCA accidents in a PWR plant

The baffle jetting phenomenon, intensified by pressure difference across baffle plates, may cause excessive fuel rod vibrations and possible damage. The downward flow direction through the baffle-barrel region opposite to the upward flow through the core in pressurized water reactors affects increase of pressure gradient. Gaps between the baffle plates widen with reactor operating time and water jets become more pronounced. During the 2015 outage the NPP Krško has undergone the upflow conversion (UFC) modification in order to minimize baffle jetting and prevent fuel rod failure. The modification consisted in altering reactor vessel internals in such a way that the coolant downflow path in the baffle-barrel region was converted to an upflow path.The impact of the UFC modification on the nuclear power plant behaviour was analyzed in the paper. The double ended guillotine break of one direct vessel injection (DVI) line was selected as a representative scenario because the DVI line break is exhibiting the highest peak cladding temperature among the small break LOCA cases. The thermal-hydraulic analysis was carried out with the RELAP5/Mod3.3 code for core average channel conditions, while additional FRAPTRAN runs were performed to evaluate peak cladding temperatures for high power and high burnup fuel assemblies. Distribution of pressure and velocity fields in the reactor pressure vessel baffle-barrel region and change in bypass flow rates due to modification were determined by the ANSYS-Fluent calculation. The accident analysis has been performed for both the downflow configuration (pre-UFC) and the upflow configuration (post-UFC) to assess the influence of the modification. Additional cases with different reactor coolant pump trip times, higher core decay heat and reduced capacity of safety systems were also analyzed to evaluate the impact on reactor safety.

Will 37Ar emissions from light water power reactors become an obstacle to its use for nuclear explosion monitoring?

37Ar is a promising candidate for complementing radioxenon isotopes as indicators of underground nuclear explosions. This study evaluates its potential anthropogenic background caused by emissions from commercial pressurized water reactors. Various 37Ar production pathways, which result from activation of 36Ar and of 40Ca, respectively, are identified and their emissions quantified. In-core processes include (1) the restart of operation and degassing of the primary cooling water after maintenance and refueling shutdown, (2) the replacement of primary coolant water for limiting its tritium concentrations, and (3) the leakage of 37Ar produced from calcium impurities in UO2 after fuel rod failures. Activation of air and of calcium in concrete within the biological shield are major out-of-core production pathways. Whereas emissions from in-core processes are transient, a rather constant 37Ar source term results from its out-of-core production. Generic atmospheric dispersion simulations indicate that already at moderate distances from the emitter, concentrations of 37Ar caused by routine reactor operations are far below its cosmogenic background in air. The only exception results from an inadvertent reactor re-start without operation of the primary cooling water degassing system for prolonged time. Such an event also causes high emissions of 41Ar which can be used for discriminating its 37Ar signal from an underground nuclear explosion.

Initial development of an RIA envelope for dispersed nuclear fuel

A reactivity insertion accident (RIA) is a design basis accident in which reactivity is rapidly injected as a result of a control rod ejection or blade drop scenario. The resultant power increase can result in fuel rod failure and subsequent release of radioactive material into the reactor’s primary system. For light water reactors, failure under RIAs is typically a result of fuel melt, pellet-cladding interaction, or rod over-pressurization. Fuel melt occurs when the fuel system is unable to transport heat out of the fuel system. Pellet-cladding interaction occurs when an aggressive fuel pellet expansion pushes on an embrittled cladding beyond its strain limit. Rod over-pressurization occurs when a fuel rod exceeds the critical heat flux and departs from nucleate boiling, causing the rod internal pressure to rapidly increase, exceeding the systems pressure, and balloon until it bursts. To prevent failure, regulators have developed safety requirements that limit the injected enthalpy based on the state of the fuel system. However, dispersed fuel could enable the removal or relaxation of regulator-imposed safety criteria. Dispersed fuel embeds fuel particles in a highly conductive metal. Dispersed particles can have a particle radius of 1 µm to greater than 100 µm. Traditional fuel systems are plagued by poor thermal conductivity, especially at higher burnups, thus reducing the ability of the fuel system to respond to an RIA. However, dispersed fuel could improve the ability of the fuel system to transport injected energy away from the fissile material and into the coolant more efficiently. Additionally, dispersing the fuel particles mitigates hard contact that typically occurs in the traditional zirconium-uranium dioxide fuel system, thus removing pellet-cladding mechanical interaction as a potential failure mechanism. This paper evaluates fuel particles dispersed in a metal matrix using the BISON fuel performance code to develop an initial failure threshold based on melt temperature of the fuel particle and/or cladding material as a function of fission density for beginning-of-life and end-of-life conditions. BISON results show that (1) reducing the size of the particles allows fission density to increase as a function of time, (2) particle-to-particle proximity must be considered to evaluate the limiting conditions, and (3) improving the fuel particle thermal conductivity improves thermal performance.

Development of fuel rod failure character analysis code for pressurized water reactors

The analysis of fuel rod failure character is the key to a real-time detection system for fuel rod failure in a pressurized water reactor (PWR). We introduce an analysis code for fuel rod failure character in PWRs that allows online analysis and offline usage. The fuel rod failure data from some nuclear reactors were used to verify the analysis code. The results show that the number of failed fuel rods calculated by the analysis code based on the isotopes of Kr and Xe in the fission gases fits very well with the actual number of failed fuel rods. The real-time detection system for fuel rod failure in PWRs and the analysis code for fuel rod failure character are of great significance for the safe operation of nuclear reactors.

DEBRIS-FRETTING TEST OF COATED AND UNCOATED ZR-1%NB CLADDING

Debris fretting is one of the most significant fuel rod failure causes. An approach to mitigate this is to apply special coatings on the surface of the cladding that are supposed to increase the wear resistance. This paper summarizes the experiments performed in Research Centre Rež on fuel cladding tube specimens. The tube segments made of uncoated and coated Zr-1%Nb alloy were exposed to fretting tests simulating the wire-debris vibrations impact on cladding during the reactor operation. The goal of these tests is to obtain results about the differences in wear resistance of uncoated and coated specimens. This paper shows the testing procedure in the air at room temperature and obtained impact markings on the specimens’ surfaces.

Simplified beam model of high burnup spent fuel rod under lateral load considering pellet-clad interfacial bonding influence

Abstract An integrated approach of model simplification for high burnup spent nuclear fuel is proposed based on material calibration using optimization. The spent fuel rods are simplified into a beam with a homogenous isotropic material. The proposed approach of model simplification is applied to fuel rods with two kinds of interfacial configurations between the fuel pellets and cladding. The differences among the generated models and the effects of interfacial bonding efficiency are discussed. The strategy of model simplification adopted in this work is to force the simplified beam model of spent fuel rods to possess the same compliance and failure characteristics under critical loads as those that result in the failure of detailed fuel rod models. It is envisioned that the simplified model would enable the assessment of fuel rod failure through an assembly-level analysis, without resorting to a refined model for an individual fuel rod. The effective material properties of the simplified beam model were successfully identified using the integrated optimization process. The feasibility of using the developed simplified beam models in dynamic impact simulations for a horizontal drop condition is examined, and discussions are provided.

Efficiency calibration of an on-line detection device for fuel rod failure in a PWR

This paper describes a method of calibrating the efficiency of an on-line detection device for fuel rod failure in a pressurized water reactor. 214Am, 137Cs, 60Co and 24Na were selected to produce 4 sets of liquid calibration sources, and a calibration system that could simulate the on-site measurement of the on-line detection device was established. Efficiency calibration was performed against γ-ray lines, which could be identified by a LaBr3(Ce) detector. After a true coincidence summing correction, an efficiency curve of the on-line detection device for fuel rod failure was obtained, with a calibration uncertainty of 3.5%.

Structural integrity of a high-burnup spent fuel rod under drop impact considering pellet-clad interfacial bonding influence

Abstract Variation of several uncertain properties of an irradiated fuel assembly would result in changes in the buckling and bending resistance that could lead to fuel rod failure in a drop accident. Dynamic analysis of the fuel rod composite is a very complicated process due to limited knowledge on important parameters relevant to design safety, such as the material properties of the irradiated cladding and the interaction between fuel pellets and cladding. In order to better understand the behavior of a spent nuclear fuel (SNF) rod in a drop impact, a detailed finite-element (FE) model for a single fuel rod was developed considering post-irradiated fuel conditions. Free-drop analyses of two bounding cases were conducted using FE analysis, with ten assumed generalized interface bonding and de-bonding configurations to demonstrate the pellet-clad interfacial bonding influence on fuel rod behavior. It is revealed that the interfacial bonding between pellet and cladding shows a significant influence on the magnitude of the maximum principal plastic strain and its location, and on the overall structural resistance. The key feature of the proposed analysis procedure is seen in its capability to consider a range of parameters that can be included in the FE model, in addition to providing a more realistic and descriptive prediction of fuel rod behavior. The procedure and methodology developed for the numerical simulation in this work can be further extended to support the development of guidelines for safe handling of SNF when augmented by realistic data obtained experimentally.

Fretting wear comparison of cladding materials for reactor fuel cladding application

Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, such grid-to-rod-fretting is a significant cause of fuel failures in U.S. pressurized water reactor power plants. Kanthal advanced powder metallurgy technology or APMT, an FeCrAl steel alloy, and a braided SiC fiber, Chemical Vapor Infiltration SiC matrix (SiC/SiC) cladding by General Atomics are possible alternatives to conventional fuel cladding in a nuclear reactor due to their favorable performance under accident conditions. Tests were performed to examine the reliability of the cladding candidates and a conventional cladding, Zircaloy-4, under dry fretting conditions at elevated temperature. The contact was simulated with a rectangular and a cylindrical specimen over a line contact area. Confocal scanning laser microscopy was used to obtain a 3D map of the surface, which was in turn used for wear and work rate calculations on the samples. The wear rate coefficient was used as a measure of the performance and wear under fretting. Additionally, Energy Dispersive Spectroscopy was performed to qualitatively describe the microchemical changes the material undergoes during fretting. While APMT steel and SiC/SiC can perform favorably in loss of coolant accident scenarios, they also need to perform well when compared to Zircaloy-4 with respect to fretting wear. Wear coefficient measurements showed that APMT steel performs favorably in comparison to Zircaloy-4 with respect to fretting wear.

Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRA-IBRAE/LM thermohydraulic code

Experiments with impulse gas injection into model coolants, such as water or the Rose alloy, performed at the Novosibirsk Branch of the Nuclear Safety Institute, Russian Academy of Sciences, are described. The test facility and the experimental conditions are presented in details. The dependence of coolant pressure on the injected gas flow and the time of injection was determined. The purpose of these experiments was to verify the physical models of thermohydraulic codes for calculation of the processes that could occur during the rupture of tubes of a steam generator with heavy liquid metal coolant or during fuel rod failure in water-cooled reactors. The experimental results were used for verification of the HYDRA-IBRAE/LM system thermohydraulic code developed at the Nuclear Safety Institute, Russian Academy of Sciences. The models of gas bubble transportation in a vertical channel that are used in the code are described in detail. A two-phase flow pattern diagram and correlations for prediction of friction of bubbles and slugs as they float up in a vertical channel and of two-phase flow friction factor are presented. Based on the results of simulation of these experiments using the HYDRA-IBRAE/LM code, the arithmetic mean error in predicted pressures was calculated, and the predictions were analyzed considering the uncertainty in the input data, geometry of the test facility, and the error of the empirical correlation. The analysis revealed major factors having a considerable effect on the predictions. The recommendations are given on updating of the experimental results and improvement of the models used in the thermohydraulic code.

Fretting Wear Behavior of APMT Steel at 350°C for Reactor Fuel Cladding Application

ABSTRACTIn support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. A combination of SEM analysis and wear & work rate calculations are performed on the samples to determine their performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.