Pellet-cladding Mechanical Interaction Research Articles

Fuel pin failure modes in the UK's advanced gas cooled reactors

The current paper presents a review of key failure modes in AGR fuel, a description of failure mechanisms and the post failure effects that follow.The main failure modes of AGR fuel are found in the following proportions: ratchetting (axial clad straining and fuel-stacking with resulting fuel-stack gap formation and clad collapse) 3%; over pressure 4%; pellet clad mechanical interaction 7%; pin-brace interaction/rotating pins (fuel pin and support structure interaction, also known as fretting) 37%; clad restructuring/microstructural effects (including clad grain growth reducing ability of clad to withstand pellet clad interaction) 41%; other (including chipped pellets, endcap welds, and non-fuel component damage) 8%.A description of each failure mode is provided detailing the characteristics of the failure mode and a summary of the factors contributing to the failure mode.An overview of post-failure effects in AGR fuel is provided. Observations of failed fuel in hot-cells reveal a range of post-failure effects, in addition to the characteristics which contributed to the failure. Understanding post failure effects is important in order to isolate features which may have occurred prior to failure, and to inform on the behaviour of fuel between failure and discharge from reactor.

Thermomechanical analysis of SiC-based duplex claddings with varying thickness ratio for accident-tolerant nuclear fuel systems

SiC-based duplex claddings, consisting of monolithic SiC and SiC/SiC fiber composite, are emerging as a promising candidate for accident-tolerant fuel (ATF) systems in nuclear reactors. To analyze the performance of ATFs with SiC-based duplex claddings, a comprehensive computational analysis framework is presented that captures the essential properties and behaviors of the UO2-SiC fuel system. Utilizing a previously developed continuum damage model, the pseudo-ductile behavior of SiC/SiC fiber composites is accurately modelled, connecting damage evolution parameters to instantaneous stiffness matrix degradation. This framework is used to investigate the performance of UO2-SiC fuel rods under normal operating conditions and a typical Loss of Coolant Accident (LOCA) scenario. We assess the effects of the thickness ratio of the monolithic SiC and SiC-based composite layers, as well as pellet-clad cold gap thickness on the failure and leakage probabilities of the cladding. These claddings, with a thickness ratio ranging from 0.25 to 0.75, demonstrated minimal failure and leakage probabilities for both the original and reduced pellet-clad gap thickness (82.5/70 µm). When the gap thickness was further reduced to 57.5 µm, pellet-cladding mechanical interaction was observed and this greatly elevated the failure probability of the MSiC layer, thus giving rise to a loss of hermeticity. This research underscores the significant role of varying individual layer thicknesses in shaping fuel rod safety and offers potential for optimization across diverse operational conditions.

PCMI Uncertainty and Sensitivity Analysis of Es-Salam Reactor Fuel Rods During Steady-State Operations

Nuclear safety relies heavily on the quality of the results of numerical simulation codes. Among the various components of the simulation of the installation are the pellet-cladding mechanical interaction (PCMI), and the peak cladding temperature (PCT). Although the correlations describing the physical, mechanical, chemical, and thermal phenomena that occur in nuclear installations have reached a high level of quality, there remain uncertainties on the final results due to uncertainties in the input parameters which cannot be eliminated. A realistic estimate of these uncertainties is necessary to evaluate the reliability of the simulation results. When the best-estimate approach plus uncertainty (BEPU) is employed in the design of a nuclear installation, design-basis accidents are studied more realistically. This method must be used even in the design of research reactors because they are at the origin of any development of nuclear technology. We propose through this study an uncertainty and sensitivity analysis of PCMI and PCT of a heavy water nuclear research reactor fuel rod. To determine the input parameters that influence PCMI and PCT, we utilize the FEMAXI-6 code. The thermodynamic table of the FEMAXI-6 code is adapted to the case of heavy water. Two system codes are used for uncertainty and sensitivity analysis: RELAP5 and PARET. The study confirmed that in the event of a shortage of heavy water, light water can be injected in its place to remove decay heat from the core and shut down the reactor safely. The safety margin between the PCT and the saturation temperature is reduced from about 10°C in the conservative approach to less than 1°C by the BEPU approach.

High temperature nanoindentation of (U,Ce)O2 compounds

Continuing to refine our knowledge of the evolving mechanical properties of nuclear fuel over the entire fuel service cycle is necessary to understand the pellet-clad mechanical interaction that occurs in the fuel rods during the operation. A challenge with measuring the mechanical properties of irradiated fuels is their high levels of radioactivity that usually require the use of hot cells making testing time consuming and expensive. Nanoindentation based techniques can be employed on minute volumes of material to measure mechanical properties, including Young’s modulus, hardness, and creep stress exponents. Increasing the mixed oxide fuels mechanical properties database through a variety of testing techniques should enhance modelers’ abilities to predict failure mechanisms in the fuel/clad interface. A current challenge to testing mixed oxide fuels is the plutonium component in the fuel. Mixed fluorite type oxides with ceria (CeO2) can be used as a surrogate for mixed oxide fuels. In this study, (U,Ce)O2 solid solutions samples are used to develop elevated temperature nanoindentation and nanoindentation creep testing methods for use on mixed oxide fuels. Nanoindentation testing was performed on 3 separate (Ux-1,Cex)O2 compounds ranging from x equals 0.1 to 0.3 in equal steps at temperatures up to 800 °C: their Young’s modulus, hardness, and creep stress exponents were evaluated. The Young’s modulus decreases in the expected linear manner while the hardness decreases in the expected exponential manner. The nanoindentation creep experiments at 800 °C give stress exponent values, n = 4.7–6.9, that suggests dislocation motion as the deformation mechanism.

Temperature-Dependent Mechanical Anisotropy in Textured Zircaloy Cladding

This work evaluates deformation activity in textured zircaloy-4 cladding with the aim of elucidating the mechanisms that govern mechanical behavior across a temperature range anticipated in reactor service. In particular, the dependence of mechanical anisotropy on temperature and stress state (ratio of applied stresses) is examined via uniaxial and biaxial tensile experiments at room temperature and 400 °C. The mechanical behavior is interpreted based on micro-texture and dominant slip system activity. Thermal activation of basal systems is found as a key mechanism affecting mechanical behavior across the temperature range. The biaxial stress states produce resolved shear stresses contrasting to uniaxial stress states. Mechanical testing and implications for reactor pellet-cladding mechanical interaction stress conditions are discussed.

CRUD deposition induced multi-physics effects analysis in PWR

In PWR, the deposition of corrosion products from the steam generator leads to the formation of Chalk River Unidentified Deposits (CRUD) on the surface of fuel rods. CRUD deposition can cause issues such as CRUD induced localized corrosion, axial power anomaly and deterioration of cladding heat transfer. A new multi-physics program named CIFF (CRUD Induced eFFects analysis) is developed to analyze the effects caused by CRUD deposits. CIFF is a 1.5D model and can be used to simulate the variations in surface temperature, cladding corrosion and fuel deformation caused by CRUD deposition. A three-node CRUD deposition model is developed based on mass transfer theory and chemical equilibrium. The calculations of heat transfer, mechanical deformation, oxide layer growth, CRUD deposition and fission gas behavior are fully coupled based on finite element analysis and numerical solution technology. Axisymmetric element is used to solve the deformation of fuel rod under thermal expansion, gas pressure, creep and radiation swelling. The performance of CIFF is validated through IFA-513, ANO-2, Oconee and PCCL cases. The DBN-1 case is used to explore the influence of CRUD. The results indicate that CRUD deposition and CRUD induced local corrosion deteriorate the heat transfer of cladding, resulting in a three-stage increase of cladding temperature and accelerated growth of the oxide layer. CRUD deposition also leads to a relative increase in the gap pressure exceeding the coolant pressure by 8.4 %, a relative increase in the fuel rod strain by 10 % and a relative decrease in the gap size by 11 %, which indirectly leads a premature pellet-cladding mechanical interaction by 90 days.

A study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions

ABSTRACT In a part of the historical reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the condition of such fracture pattern change. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of ℇz/ℇθ = 0.5–1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio ℇz/ℇθ >0.75 and a concurrent hoop strain < 1.7% at the failure instant.

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

This paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO2. The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB2 Integral Fuel Burnable Absorber (IFBA) and B4C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd2O3-doped UO2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

A reactivity-initiated accident (RIA)-simulated test CN-1 on a high-burnup 64 GWd/t mixed-oxide fuel rod sheathed with M5™ cladding was conducted at the Nuclear Safety Research Reactor, resulting in fuel failure. A small opening with slight ballooning deformation characterized the post-test visual appearance of the test fuel rod. Simulation using fuel performance codes FEMAXI-8/RANNS predicted rod survival under early phase loading induced by pellet-cladding mechanical interaction and subsequent boiling transition, and the cladding-surface temperature measured online confirmed the occurrence of boiling transition. The experimental observation and simulation indicate that the failure was caused by a high-temperature rupture following increased rod-internal pressure. The RANNS sensitivity analysis revealed that a mechanical state parameter dedicated to predicting plastic instability might be an effective index for evaluating the risk of rupture failure during RIAs.

A plasma-based pressure pulse generator for simulating pellet-cladding mechanical interaction during reactivity-initiated accident.

This paper describes the characteristics of a plasma-based pressure pulse generator and its potential use as a simulator for studying pellet-cladding mechanical interaction (PCMI) during reactivity-initiated accidents (RIAs). In this device, a transient pressure pulse is generated by rapid heating and expansion of hot, dense plasma inside a nuclear fuel cladding. Thus, the parameters of a pressure pulse, such as peak pressure and pressure rise-rate, can be controlled by modifying the electrical parameters of a pulse discharge circuit. The pulse discharge circuit utilizes a capacitor bank comprising several energy storage capacitors connected in parallel and a high-power solid-state switch. A pressure loading system is attached as a load to the pulse discharge circuit. The power and energy delivered to the load are calculated by measuring the voltage and current waveforms at one end of the loading system. A piezoelectric sensor is connected at the other end of the loading system to simultaneously measure the pressure pulse inside the cladding tube. Preliminary experiments are carried out with a stainless-steel tube to characterize the performance of the device as well as with a pre-hydrided ZIRLO™ cladding tube to demonstrate the potential of the device as a simulator for studying the failure characteristics of the cladding as a result of an RIA. The high pressurization rate of the device is expected to offer unique advantages for studying the PCMI mechanism.

Analysis of thermal fuel behaviour under steady-state irradiation using selected cases of the IAEA CRP FUMEX

Results of the project “Comprehensive Verification of the FALCON Code for Calculation of Nuclear Fuel Temperature”, relating to thermal fuel behaviour under steady-state (SST) irradiation and slow power ramps are presented. Two groups of irradiation experiments carried out in the Halden Boiling Water Reactor (HBWR) are analyzed. Improvement of the prediction capability with respect to fuel temperature under SST irradiation after the earlier calibration of the selected models is confirmed. The data from the on-line measurement of fuel temperature are interpreted in consideration of the effects of Fuel Thermal Conductivity (FTC) degradation with burnup, Fission Gas Release (FGR), reduction of FTC due to development of gaseous fuel porosity, and Pellet-Cladding Mechanical Interaction (PCMI). Relevance of the analyzed data to current new concepts of Advanced Technology Fuel (ATF) is noted.

Computer code analysis of irradiation performance of an annular mixed oxide fuel element

ABSTRACT Computer code analyses were conducted for irradiation behaviors of a fuel element with annular mixed oxide fuel pellets attaining a pellet peak burnup of 60 GWd/t in a fast reactor. Post-irradiation examinations of annular and solid fuel elements revealed that the diameter strain of the cladding of the annular fuel element was smaller than that of the solid fuel element and that the cesium concentration increased locally near the interface between the fuel pellet column and the lower blanket pellet in the annular fuel element, while cesium concentration was rather uniform over the fuel pellet column in the solid fuel element. The code analysis clarified the mechanism of these experimental results. The smaller cladding diameter strain of the annular fuel element was due to a delay in the onset of fuel pellet-cladding mechanical interaction, which increased the outer surface temperature of the fuel pellet, thereby enhancing the axial migration of cesium toward the fuel–blanket interface.

Fuel performance uncertainty quantification and sensitivity analysis in the presence of epistemic and aleatoric sources of uncertainties

Fuel performance modeling and simulation includes many uncertain parameters from models to boundary conditions, manufacturing parameters and material properties. These parameters exhibit large uncertainties and can have an epistemic or aleatoric nature, something that renders fuel performance code-to-code and code-to-measurements comparisons for complex phenomena such as the pellet cladding mechanical interaction (PCMI) very challenging. Additionally, PCMI and other complex phenomena found in fuel performance modeling and simulation induce strong discontinuities and non-linearities that can render difficult to extract meaningful conclusions form uncertainty quantification (UQ) and sensitivity analysis (SA) studies. In this work, we develop and apply a consistent treatment of epistemic and aleatoric uncertainties for both UQ and SA in fuel performance calculations and use historical benchmark-quality measurement data to demonstrate it. More specifically, the developed methodology is applied to the OECD/NEA Multi-physics Pellet Cladding Mechanical Interaction Validation benchmark. A cold ramp test leading to PCMI is modeled. Two measured quantities of interest are considered: the cladding axial elongation during the irradiations and the cladding outer diameter after the cold ramp. The fuel performance code used to perform the simulation is FAST. The developed methodology involves various steps including a Morris screening to decrease the number of uncertain inputs, a nested loop approach for propagating the epistemic and aleatoric sources of uncertainties, and a global SA using Sobol indices. The obtained results indicate that the fuel and cladding thermal conductivities as well as the cladding outer diameter uncertainties are the three inputs having the largest impact on the measured quantities. More importantly, it was found that the epistemic uncertainties can have a significant impact on the measured quantities and can affect the outcome of the global sensitivity analysis.

Towards coupling conventional with high-fidelity fuel behavior analysis tools

Multi-dimensional fuel performance codes have the potential to enhance the understanding of complex phenomena affecting fuel integrity, but their application is often limited by the high computational cost. Recognizing that many multi-dimensional aspects such as the ballooning of the cladding during a loss of coolant accident or the pellet-cladding mechanical interaction (PCMI) during power ramps are of interest mainly for operational and accidental transients, we have developed a coupling methodology between conventional and high-fidelity fuel behavior analysis. This consists in using a well-validated 1.5-D code to rapidly simulate the base irradiation and set the initial and boundary conditions for the 2-D or 3-D transient model, leading to an optimized use of computational resources. In this paper we present the coupling methodology, and we apply it to link the TRANSURANUS code and the multi-dimensional code OFFBEAT. To demonstrate the capabilities of the coupling tool, we study the PCMI during the end-of-life power ramp of the AN3 segment from the Risø-3 experiment and we compare the predicted cladding deformation and bamboo structure against the experimental profilometry provided by the Post Irradiation Examination (PIE). In the conclusions, we also outline further potential applications of the coupling methodology.

Evaluation of anisotropic elastic and plastic parameters of zircaloy-4 fuel cladding from biaxial stress test data and their application to a fracture mechanics analysis

ABSTRACT The mechanical properties of fuel cladding near the elastic limit are essential in considering its failure limit during a pellet-cladding mechanical interaction phase under reactivity-initiated accident (RIA) conditions. The mechanical properties of a Zircaloy-4 cladding tube, such as orthotropic elasticity and anisotropic constants for Hill’s plasticity law, were evaluated based on the biaxial stress test data, focusing on the equivalent plastic strain up to ~2.5%. Samples with various fabrication conditions, such as cold-worked, recrystallized, and stress relieved after cold-work with Q-factors of 2, 3, and 4 were investigated. The evaluation revealed noticeable difference in the elastic and plastic properties between cold-worked, recrystallized and stress relieved samples, including their anisotropic feature, while the impact of Q-factors was minor. The derived mechanical properties were applied to calculation of the fracture mechanics parameter, J-integral, which was found sensitive to the mechanical property differences. Critical J-integral values were then estimated by comparing the calculation and failure limit data from biaxial-expansion-due-to-compression tests on precracked tubes. This produced systematically lower J-integral values of the stress relieved tube than those previously evaluated based on the failure limit data from in-pile RIA-simulated tests.

Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

Abstract The scope of the current research in the field of fuel performance is primarily aimed at an improvement of the operating reliability, safety, and cost-effectiveness of the reactors in operation. The current requirement of nuclear industry is to have fuel suitable for load follow operation. Fission gas release, pellet-cladding mechanical interaction, and stress corrosion cracking are the main phenomena that limit the variability of reactor operation from a safety perspective. To reasonably predict the fuel performance limits, it is necessary to benchmark the computational tools against high-quality experimental data. This work is devoted to the calculation of fuel performance using the code FEMAXI-6 based on the longest irradiation experiment in the Halden reactor. The fuel burn-up was approaching 90 MWd/kgUO2 in three selected rods, which were equipped with the pressure sensors and were subjected to extensive postirradiation examination. During the experiment, the rods were exposed to several periods of power cycling. The rods were manufactured with different fuel grain size and fuel-to-clad gap size.

BISON validation of FeCrAl cladding mechanical failure during simulated reactivity-initiated accident conditions

A reactivity-initiated accident (RIA) is a postulated design basis accident in light water reactors (LWRs) in which a rapid reactivity insertion induces a fission rate increase and a fuel pellet temperature rise. During RIA, the fuel pellet thermally expands and may cause pellet-cladding mechanical interaction (PCMI). Separate effects PCMI tests were performed on C26M FeCrAl cladding tube samples, introducing biaxial stress via a well understood modified burst test (MBT) system. A high-speed camera in the MBT system captured the projections, covering a 360° view of the cladding deformation and enabling a digital image correlation (DIC) method to quantify the surface strains with high fidelity. Representative hot zero-power RIA mechanical loading conditions were applied to the sample, and the test duration ranged from 20 to 500 ms at an average temperature of 573 K. BISON finite element–based fuel performance code modeling was performed against the high-fidelity DIC data produced from the MBTs. Validation calculations were conducted with 2D models and systematically compared with test data of the burst time, burst pressure, burst hoop strains, and hoop strain rates. Based on the behaviors from the separate effects test, BISON calculations satisfactorily predicted the cladding deformation behaviors. Sensitivity analysis was conducted to identify highly influential mechanical properties responsible for the cladding failure behavior during the MBT experiments. The results highlight the significance of the cladding's mechanical strength in governing cladding strain, followed by the significance of the mechanical interactions between the pellet and the cladding.

Analysis of pellet-to-cladding gap closure for a metallic micro-cell pellet under normal operating conditions

To reduce fission gas release under both accident conditions and normal operation, Korea Atomic Energy Research Institute has developed a metallic micro-cell pellet following the 2011 Fukushima accident. One concern, however, about the metallic micro-cell pellet is that due to the presence of metallic wall that reduces both temperature and gas release, its capability of higher gas retention could result in increased gaseous swelling, leading to earlier pellet-to-cladding gap closure than the conventional UO 2 pellet. To analyze and compare the gap closure in two fuel rods in which two kinds of pellets, UO 2 and metallic micro-cell, are loaded into each of the two identical Zr claddings, we developed a model to estimate gaseous swelling in the two pellets. Then considering mainly different gaseous swelling and thermal expansion by different temperature profiles in the two pellets, we compared the pellet-to-cladding gaps as a function of burnup for the case where they would be irradiated with the same constant linear power of 30 kW/m up to 60 MWd/kgU. Due to both reduced gaseous swelling and lower thermal expansion, gap closure in the fuel rod loaded with the metallic micro-cell pellet is expected to be delayed by 6–16 MWd/kgU compared with the one loaded with the UO 2 pellet, implying pellet-cladding mechanical interaction occurrence would also be postponed by the same degree. We note that the result of this approach is preliminary until confirmed by measured data from metallic micro-cell pellets irradiated up to high burnup. • Higher gas-retention capability in the metallic micro-cell pellet could lead to earlier pellet-to-cladding gap closure than the UO 2 pellet. • Despite the higher gas-retention capability, gaseous swelling in the metallic micro-cell pellet would be reduced by its lower temperature. • Pellet-to-cladding gap closure in the metallic micro-cell pellet could be delayed by 6–16 MWd/kgU compared with the UO 2 pellet.

The effect of strain biaxiality on the fracture of zirconium alloy fuel cladding

• Experimental device for tubular specimen under biaxial load. • Biaxiality effect on the fracture behavior of zirconium alloy cladding. • Fracture mechnisims of zirconium alloy cladding. During a Reactivity Initiated Accident (RIA), nuclear fuel cladding experiences a multiaxial loading state in which the Pellet-Cladding Mechanical Interaction (PCMI) produces a strain biaxiality ratio (ε zz /ε θθ ) of between 0 and 1. This study examines the effect of the strain biaxiality ratio on the fracture of Zircaloy-4 fuel cladding. A new mechanical test has been developed in order to apply several levels of strain biaxiality. Digital Image Correlation (DIC) is used to measure the strain field, to identify the onset of cladding failure, and to obtain the corresponding critical loading state. Results show that the strain biaxiality has a significant effect on the hoop strain at failure, and the smallest fracture strain is obtained for nearly plane strain conditions. A model is proposed for the accumulation of damage due to plastic deformation, based on the Lode parameter of the plastic strain rate tensor. The model is used to predict the failure hoop strain as a function of the strain biaxiality ratio (ε zz /ε θθ ), and good agreement is found between the experimentally measured and predicted failure strains.

Pellet-cladding mechanical interaction analysis of heavy water fuel rods under power ramps

A numerical simulation is used to analyze the pellet-cladding mechanical interaction (PCMI) of a heavy water research reactor (RR) fuel rods under power ramps. A mechanical simulation that best reproduces the thermomechanical behavior of the pellet-cladding system is a major industrial challenge for engineers because it involves defining operating rules that are not too restrictive but that guarantee safety requirements. The majority of currently existing codes for processing thermomechanical behavior, under irradiation, of the PCMI evolution are designed for light water reactors. This is justified by the fact that most of the world nuclear power reactor park is composed of light water reactors. However, many nuclear reactors use heavy water as coolant, e.g. type CANDU reactors. In addition to the coolant thermodynamics table which is different for the two types of codes, the conditions of use in terms of pressure and temperature in the case of RRs are less severe than those in the case of nuclear power reactors (NPRs). In this study, we used several computer codes namely MCNP5, RODBURN, FEMAXI-6, and an analytical application, developed locally, to simulate the thermomechanical behavior of the Es-Salam RR fuel rods. The fission product cross-section library that was used with the RODBURN code was produced by using the ORIGEN code fed with the neutron flux spectrum of the Es-Salam research reactor. A new approach has been developed for the PCMI analysis of a heavy water reactor by using a code originally developed for light water reactors. Operating limits have been determined when the two tables of thermodynamics of coolants are different.