LWR Reactor Research Articles

ATF Claddings after High-Temperature Steam Oxidation: WDS and Nanoindentation Studies to Characterize in-depth Material Changes

After the Fukushima accident in 2011 year, ATF nuclear fuel cladding concept was accelerated to achieve the reactor operation with the new accident tolerant structural materials. However, several designed solutions do not fulfil the accident tolerant concept but particularly increase the corrosion resistance of Zr-cladding tubes at normal operating conditions, so-called “Advaced Technology Fuel, EATF”. Cr-coated zirconium claddings following the first concept, have been the widely tested and the first full Cr-coated fuel rods have been planned to operate in LWR reactor conditions around the 2022 year. Our contribution describes the Cr-coated Zr-%1Nb cladding tube microstructure after high-temperature steam oxidation at 1200°C by means of Scanning Electron Microscopy and nanoindenta-tion methods. The article is focused on WDS line-profile studies of oxygen and chromium diffusion into the Zr-matrix. The increased Cr-diffusion with oxygen is evident causing a change in local me-chanical properties which is well-described by measurements of nanohardness and Young's modulus. In addition, the developed methodology of the WDS & nanoindentation line-analyses was also opti-mized to apply in hot-cell conditions to measure the effect of neutron-irradiation on the different coat-ings and coating/matrix interface.

Thermal-hydraulic validation of two-phase models in THUNDER code against benchmark results and CFD codes

The study of two-phase flow behavior is a very important requirement in nuclear reactor systems, due to it can appear during normal operating conditions in a BWR or fast and slow transients for both PWR and BWR reactors. This work presents the evaluation at steady-state of two-phase flow prediction performed by THUNDER code (Thermal-Hydraulic Unit of Numerical Development for Elements with Rod-bundles), a new computational code that has been developed to perform a thermal-hydraulic analysis of LWR reactors with rod-type fuel assemblies. The physic and mathematical models used to develop the steady-state analysis of a nuclear reactor core are presented. The conservation equations of mass, momentum, and energy were established based on operating and geometric conditions, for single and two-phase flow using the subchannel and control volume methods, it applied to square arrays to the four sub-channel types (center-typical, center-thimble, edge, and corner) found in a PWR fuel assembly. THUNDER code performs a tridimensional analysis for the whole core solving simultaneously the conservation equations for each control volume. For two-phase flow treatment, the code uses the drift flux model. It provides detailed information of operational conditions e.g. coolant, cladding and center fuel temperatures, core pressure drop, coolant velocity, void fraction, coolant quality, onset of nuclear boiling (ONB) and departure from nucleate boiling ratio (DNBR). To validate the code, we consider experimental data from OECD/NRC benchmark based NUPEC PWR subchannel and bundle tests (PSBT) and CFD calculations from ANSYS-CFX®. It was evaluated the void fraction, average density and pressure drop for individual subchannel as well as for a rod bundle 5 × 5 fuel assembly. THUNDER code reproduced well the benchmark results presenting a mean absolute discrepancy of void fraction equal to 0.039 and density relative discrepancy equal to 0.11. This indicates that THUNDER code performs a two-phase flow evaluation within the acceptance criterion for reactors with rod-type fuel assemblies under steady-state conditions.

Assessment of representativity of the PETALE experiments for validation of Swiss LWRs ex-core dosimetry calculations

The international experimental program PETALE will be carried out at the CROCUS research reactor of EPFL. The program aims at measuring neutron penetration in slabs made of materials composing typical LWR reactor pressure vessel. The measurements will be used for code and nuclear data validation and for the evaluation of the reflecting properties of these materials. In this paper the representativity of the PETALE experiments is assessed with respect to operational LWR reactors dosimetry and activation evaluations using the Paul Scherrer Institute (PSI) in-house tool NUSS. The NUSS tool allows the stochastic sampling of nuclear data using covariance matrices available in modern nuclear data libraries and the subsequent running of a Monte Carlo code with the modified data files. The representativity can then be assessed based on the Pearson correlation coefficients. The ultimate goal of the work is first of all to assess if the planned PETALE measurements could be applicable beyond their primary purpose and serve for extending the PSI validation database for LWR reactor dosimetry evaluations. Secondly, provided that the PETALE measurements are found useful for the task above, the information on the correlations between the PETALE neutron detectors’ responses and the reactor dosimetry quantities of interest, shall be presented and discussed.

Influence of accident tolerant fuel cladding material on the European pressurised reactor core neutronic characteristics

Research on accident tolerant fuel became an important research area to improve safety of the nuclear reactors, especially following the Fukushima disaster. Numerous studies have demonstrated that silicon carbide is a promising candidate material that could potentially replace the zirconium cladding in LWR reactors, which is currently under active research by multiple research and industry organizations. Many of material testing of silicon carbide were already performed and the results are available. Also molybdenum is considered as fuel cladding. This paper is focused on the potential of using silicon carbide and molybdenum as a fuel cladding in the European Pressurized Reactor (EPR) core from the neutronic point of view.

Modeling porosity migration in LWR and fast reactor MOX fuel using the finite element method

An engineering-scale finite element simulation of pore migration in oxide fuel is presented. The porosity field is governed by an advection-diffusion equation which is coupled to the fuel temperature and stress fields through the thermal conductivity and volumetric heat source term. The engineering-scale porosity equation models the microscopic process of vapor transport of fuel across pores, taking into account thermal and vapor pressure gradients within the fuel. In the simulations, the porosity is initialized to a constant value at every point in the domain, and as the temperature gradient is increased by application of a heat source, the pores move up the thermal gradient and accumulate at the center of the fuel in a time frame that is consistent with experimental observations. Results from representative simulations are provided to demonstrate the new capability, and we show that a sufficiently high power ramp rate limits restructuring and leads to a corresponding increase in fuel temperature. We also discuss the finite element mesh density required to compute pore migration and present multidimensional results.

Discrimination of plutonium from thermal reactors in the frame of nuclear forensics

The potential discrimination of 1st generation Pu, averaged over different assemblies of spent fuels from thermal reactors, has been assessed through a simulated fingerprinting procedure. Plutonium, retrieved from spent fuels of burnup in the range 5–50 GWd/MTU and different 235U enrichments at charge, has been considered. The study was based on the Pu isotopic composition, as the required characteristic signatures, retrieved from the reprocessing of spent fuels from PWR, VVER, BWR, RBMK, AGR, MAGNOX and CANDU reactors, charged with different assemblies. The procedure has resolved the Pu of the spent fuels from the different reactors and burnup range considered. The Pu from the different PWR, VVER and BWR spent fuels were clustered together in accordance with the reactor type and the initial enrichment of the fuels. Regarding the AGR, MAGNOX, RBMK and CANDU reactors, the Pu from the spent fuels are discriminated between them and from the LWR reactors considered.

PCI analysis of Zircaloy coated clad under LWR steady state and reactor startup operations using BISON fuel performance code

Research following the earthquake and subsequent tsunami that devastated the Fukushima Daiichi nuclear power plant, causing severe damage to the reactor cores, has led to the development of accident tolerant fuel designs. Accident tolerant fuels are described as a fuel form which exhibits improved material response, when compared to traditional uranium dioxide fuel clad by a zirconium alloy, during accident (e.g., loss of coolant accident or reactivity insertion accident) while maintaining or exceeding normal reactor operational expectations. One of the primary goals of accident tolerant fuel concepts is to reduce cladding corrosion by either replacing traditional Zircaloy cladding with a new material or by applying a coating to the outer surface of the cladding with enhanced resistance to corrosion at high temperatures. However, while coating the clad may provide an increase in high temperature corrosion resistance, it also changes the mechanical state of the cladding and potentially creates alternate failure mechanisms during reactor operations or accident conditions. This research has primarily focused on the pellet-cladding interactionmechanical response of FeCrAl (Iron-Chromium-Aluminum alloy) coated cladding during light water reactor steady state and reactor startup operations. Using the MOOSE-based, finite-element fuel performance code BISON and commercial pressurized water reactor data obtained from Diablo Canyon Unit 2, a preliminary analysis was performed to evaluate the performance of FeCrAl coated cladding under steady state operation and its mechanical stability under pellet-cladding interaction during a reactor startup. This work summarizes BISON 2-D radial-axially symmetric (R-Z) and radial-circumferentially symmetric (R-θ) simulation results by comparing 20, 50, and 100-µm FeCrAl coating thicknesses as well as FeCrAl cladding to traditional Zr-4 cladding under pellet-cladding interaction condition. The results of this study will provide a feasibility analysis for steady state and transient operational response of FeCrAl coated, Zircaloy cladding. Furthermore, this work will provide a fundamental basis for the assessment of the ability of coated cladding, as an ATF candidate, to maintain or exceed current steady state reactor operating expectations.

Nuclear data uncertainty propagation analysis for depletion calculation in PWR and FR pin-cells

In order to assess the nuclear data uncertainty propagation in the depletion calculation, a computational code named SUNDEW has been developed based on the home-developed lattice code NECP-CACTI. In the SUNDEW, Generalized Perturbation Theory (GPT) is applied to calculate sensitivity coefficients of response function with respect to the nuclear cross sections. Method of Characteristics (MOC) is employed to solve the transport equation, adjoint and generalized adjoint transport equations. Chebyshev Rational Approximation Method (CRAM) is implemented to solve the depletion equation and adjoint depletion equation. The sensitivity coefficients of Keff and nuclide density with respect to the nuclear cross sections are verified by comparing with the results of direct perturbation calculation. The uncertainties on Keff and the nuclide density at different depletions, which are induced by the nuclear cross sections uncertainties, are analyzed based on ENDF/B-VII.1 covariance data for LWR and fast reactor pin-cells. The numerical results show that there are significant differences between LWR and fast reactor pin-cells. The differences are mainly caused by the differences of the sensitivity coefficients between the LWR and fast reactor pin-cells. In addition, to identify the cross section improvement priority for nuclides, reactions and energy ranges, the dominant contributors to Keff and nuclide density uncertainties are analyzed at different depletions.

Recent advances in accurate neutronic calculation of GEN-3 LWR reactors

New French GEN-III reactors are characterized by an improved safety (core catcher for corium, stainless-steel neutron reflector to minimize vessel embrittlement, permanent in-core instrumentation thanks to SPND, …). Moreover, the assembly central guide-tube can be replaced by a fuel pin and the First core is controlled through numerous 8% Gd burnable poison fuel pins. These new features harden the neutron spectrum and increase the neutronic calculation challenge. Therefore, a new calculation package APOLLO2.8 was developed, based on an efficient Method of Characteristics (MOC) and the new SHEM 281-group energy mesh that enables accurate resonant absorption prediction. The JEFF3.1.1 recent nuclear data library, which involves the feedback from Critical Experiments and LWR Post Irradiation Experiments, was used. The experimental validation of the APOLLO2.8/JEFF3.1.1 package allowed the Uncertainty Quantification associated with PWR design parameters. In order to meet target-accuracy required in specific GEN-III parameters, targeted integral experiments where achieved. To validate heavy neutron reflector the mock-up experiment PERLE was carried out in EOLE critical facility. PERLE allowed the calculation check of SS reflector-saving δrefl and the improvement of Fe cross-sections. Thanks to the large French experimental database, we determined reliable covariance matrices associated to JEFF3.1.1 evaluations, in order to perform the propagation of nuclear data uncertainties. Sensitivity/uncertainty analysis has shown that the predominant component on radial power map comes from the 238U inelastic scattering cross-section. To reduce this 3% uncertainty component, we performed a 238U(n,n′) re-estimation, using relevant selected experiment benchmarks.This theoretical and experimental extensive work performed since 2003 allows a reliable prediction of the local flux and Keff (±200pcm in 1σ) in large GEN-III reactors. The challenging pin-by-pin power prediction is today overcome, and 3D-stochastic TRIPOLI4 or 2D-deterministic APOLLO2.8 calculation uncertainty is decreased from prior ±8% to current ±2.6% value.

Sustainability of the Chinese nuclear expansion: The role of ADS to close the nuclear fuel cycle

The civil nuclear energy deployment in China is important for future “Nuclear Renaissance” of China and worldwide. Compared to the other nations that developed their nuclear power energy system in last century, China can take advantage of the research and mistakes made by those states with regards to the back-end of the nuclear fuel cycle (NFC). The spent fuel accumulated by decades of operations of civil nuclear power is today a big burden for the nuclear industry. China must carefully plan the NFC for a sustainable development of the nuclear energy with special consideration to closing the fuel cycle. The present paper addresses the NFC option and implication of a LWR reactors scenario development and of a fast reactor park developed after 2035 and 2050, and covers the historical development of nuclear energy in China (i.e. from the first criticality of the first reactor) to the year 2100. The paper studies the partition and transmutation strategy with the use of accelerator driven system (ADS) to burn the MA to understand the ADS impact on the NFC and to estimate the number and the necessary deploying schedule of the ADS reactors to limit the minor actinides stock build up. The code INFCIS developed by the International Atomic Energy Agency (IAEA) is used in the present study.

Advanced multi-physics simulation for reactor safety in the framework of the NURESAFE project

Since some years, there is a worldwide trend to move towards “higher-fidelity” simulation techniques in reactor analysis. One of the main objectives of the research in this area is to enhance the prediction capability of the computations used for safety demonstration of the current LWR nuclear power plants through the dynamic 3D coupling of the codes simulating the different physics of the problem into a common multi-physics simulation scheme.In this context, the NURESAFE European project aims at delivering to the European stakeholders an advanced and reliable software capacity usable for safety analysis needs of present and future LWR reactors and developing a high level of expertise in Europe in the proper use of the most recent simulation tools including uncertainty assessment to quantify the margins toward feared phenomena occurring during an accident. This software capacity is based on the NURESIM European simulation platform created during FP6 NURESIM project which includes advanced core physics, two-phase thermal–hydraulics, fuel modeling and multi-scale and multi-physics features together with sensitivity and uncertainty tools. These physics are fully integrated into the platform in order to provide a standardized state-of-the-art code system to support safety analysis of current and evolving LWRs.

Implementation of the exponential directional weighted SN differencing scheme in RAPTOR-M3G for LWR radiation transport and dosimetry applications

A high-accuracy spatial differencing scheme is implemented in RAPTOR-M3G, a Westinghouse-developed three-dimensional parallel discrete ordinates (SN) radiation transport code. The exponential directional weighted (EDW) scheme offers enhanced accuracy and superior convergence properties compared to conventional differencing schemes at a relatively modest additional computing cost. This method enables complete convergence to typical industrial convergence thresholds on large commercial reactor problems where other differencing schemes fail. This paper explores the theoretical foundations of the EDW differencing scheme and discusses its implementation in LWR reactor problems to calculate exposure quantities of interest to the commercial reactor integrity analysis community. Comparisons to existing differencing schemes are included for several important characteristics. The computing requirements of EDW are quantified relative to the requirements of more conventional theta-weighted (TW) formulations. The convergence behavior of both methods is investigated with consideration given to spatial areas of non-convergence. Finally, the practical accuracy of both methods is assessed through comparisons of reactor dosimetry measurements to calculations from the operating fleet of commercial reactors. The EDW method and RAPTOR-M3G are demonstrated to be particularly powerful for solving the challenging three-dimensional radiation transport problems faced by today’s radiation analysis community, specifically in the extended beltline region. The EDW method enables accurate characterization of the total radiation environment in large LWR models on timescales conducive to commercial applications.

Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

Neutronic characteristics of linear-assembly breed-and-burn reactors

Linear-assembly breed-and-burn (B&B) reactors are B&B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B&B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B&B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B&B reactors is extended to apply to linear-assembly B&B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B&B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B&B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which is then used in a transition model that shows how the desired equilibrium cycle can be established. The doubling time of this fuel cycle is calculated as an example of what is achievable using B&B reactors.

Degradation of structural steel properties under the prolonged influence of operating temperatures

The generalized data of the effects of operating temperatures of 250–350°C on mechanical properties of the materials used for manufacturing of the main equipment and pipelines of Russian NPPs with LWR (VVER and RBMK) reactors for a prolonged operating time (100000 h) are presented. Degradation of mechanical properties of structural steels from this factor during their designed service life is evaluated, and also the possible use of these materials in prolonging their service life is predicted.

The simulation of a model by SIMULINK of MATLAB for determining the best ranges for velocity and delay time of control rod movement in LWR reactors

According to the complexity of nuclear reactor technology, applying a highly developed simulation is necessary for controlling the nuclear reactor control rods, so in this paper the processes of a controlling model for nuclear reactors have been developed and simulated by the SIMULINK tool kit of MATLAB software and all responses, including oscillation and transient responses, have been analyzed. In this work an arbitrary value of K eff as a comparable value is purposed and attributed to input block ( H) of diagram and then this value with the received feedback value from block diagram is compared. Since the stability of the cited simulation depends on either velocity or delay time values, therefore according to this simulation the best response and operation which a reactor can have from stability aspect, have been derived. Meantime by viewing the results, the best ranges of velocity and delay time of control rod movement (in unit per second and millisecond respectively) for stability a nuclear reactor has been deduced.

Study on the LLFPs transmutation in a super-critical water-cooled fast reactor

The performance of the super-critical water-cooled fast reactor (Super FR) for the transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with the soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super FR. First region is in the blanket assembly due to the ZrH 1.7 layer which was utilized to slow down the fast neutrons to achieve a negative void reactivity. Second region is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected in the transmutation analysis include 99Tc, 129I and 135Cs discharged from LWR or fast reactor. Their isotopes, such as 127I, 133Cs, 134Cs and 137Cs were also considered to avoid the separation. By loading the isotopes ( 99Tc or 127I and 129I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe year and 2.79%/GWe year can be obtained for 99Tc and 129I, respectively. The transmuted amounts of 99Tc and 129I are equal to the yields from 11.8 and 6.2 1000 MWe-class PWRs. Because of the very low capture cross section of 135Cs and the effect of other cesium isotopes, 135Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000 MWe-class PWRs. Based on these results, 99Tc and 129I can be transmuted conveniently and higher transmutation performance can be obtained in the Super FR. However, the transmutation of 135Cs is very difficult and the transmuted amount is less than that produced by the Super FR. It turns out that the transmutation of 135Cs is a challenge not only for the Super FR but also for other commercial fast reactors.

Analysis of temperature distribution, burn-up and breeding parameters in nuclear fuel rod in fusion-fission reactor system fueled with mixed ThO2-UO2 fuel

Temperature distribution in nuclear fuel rod, burn-up (BU), possibility of nuclear fuel rejuvenation and breeding parameters are investigated for different coolants under various first wall loads (Pw=2, 5 and 7 MW m−2) in a deuterium–tritium driven fusion–fission reactor system, fueled with mixed UO2-ThO2 fuel. The fuel mixture is considered to be mixed with various mixture fractions. A (D-T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. In the tritium breeding zone of the blanket Li2O is used and blanket is reflected by graphite for neutron economy. The fissile fuel zone is considered to be cooled with three different coolants [gas (He, CO2), Flibe (Li2BeF4) and natural lithium (Li)] with the volumetric ratio of coolant-to-fuel [e=(Vcoolant/Vfuel)=1:1 (45.515% coolant, 45.515% fuel and 8.971% clad)] for nuclear heat transfer. The behavior of the blankets fueled with mixed fuel mentioned above are observed for 48 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber. The fissile fuel breeding occurs through the neutron capture reaction in the 232Th (ThO2), in the 238U (UO2) isotopes. As a result of the calculations, fuel mixtures having the best performance parameters have been obtained as follows for different coolants and under the 7 MW m−2 first wall load during operation period without reaching the melting point: • The blanket fueled with 70 wt% UO2−30 wt% ThO2 for Flibe (Li2BeF4) coolant • The blanket fueled with 80 wt% UO2−20 wt% ThO2 for Gas (He, CO2) coolant • The blanket fueled with 90 wt% UO2−10 wt% ThO2 for natural lithium (Li) coolant Clad surface temperature Tc controlled by coolant velocity is taken between certain operation temperatures (Tc=200, 300, 400, 500 and 600°C). The best Cumulative Fissile Fuel Enrichment (CFFE) grade of the nuclear fuel calculated as the sum of the isotopic ratios of all fissile materials (233U, 235U, 239Pu, and 241Pu) is obtained in Flibe (Li2BeF4) coolant blankets, followed by Gas (He, CO2), whereas natural lithium (Li) coolant shows a poor rejuvenation performance in all fuels. CFFE reaches the maximum value (CFFE=10.6200%) in blanket cooled with Li2BeF4 coolant, followed by gas coolant with 9.0319% and natural lithium coolant with 7.9863% without reaching the melting point (Tmax=Tm>2830°C) of the fuel materials. At this point, the maximum temperature in centerline of the fuel rod (Tm) reach to 2727.57°C in blanket cooled with Li2BeF4 coolant. However, in the blanket fueled with pure UO2 fuel and cooled with Flibe, CFFE and Centerline Temperature of the fuel rods (Tm) are reached to 9.7277% and 2811.13°C respectively at the end of 39 months. In the blanket fueled with pure UO2 and cooled with gas, CFFE and Centerline Temperature of the fuel rods (Tm) are reached to 7.8115% and 2814.57°C at the end of 38 months and 7.2777% and 2724.34°C at the end of the 43 months in the blanket cooled with natural Li. In the blanket fueled with pure ThO2, CFFE reaches the maximum value with 9.2843% in the blanket cooled with Flibe. The enrichment is sufficient for LWR reactor. The temporal change of the fuel composition during the hybrid reactor plant operation and the most pertinent nuclear reactions been considered in the course of BU. The BU reaches the maximum value (72 749.1 MWd t−1 (tone)) in the gas coolant blanket, followed by Flibe coolant blanket with 67 641.0 MWd t−1 natural lithium coolant blanket with 63 897.5 MWd t−1 during the operation period without reaching the melting point. Copyright © 2010 John Wiley & Sons, Ltd.

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.

Mechanistic modelling of urania fuel evolution and fission product migration during irradiation and heating

The models of the mechanistic code MFPR (Module for Fission Product Release) developed by IBRAE in collaboration with IRSN are described briefly in the first part of the paper. The influence of microscopic defects in the UO2 crystal structure on fission-gas transport out of grains and release from fuel pellets is described. These defects include point defects such as vacancies, interstitials and fission atoms, and extended defects such as bubbles, pores and dislocations. The mechanistic description of chemically active elements behaviour (fission-induced) is based on complex association of diffusion–vaporisation mechanism involving multi-phase and multi-component thermo-chemical equilibrium at the grain boundary with accurate calculation of fuel oxidation. In the second part, results of the code applications are given to different situations: normal LWR reactor operation, high temperature annealing, loss of coolant accident (LOCA) and severe accidents conditions.