[en] As part of the Tractebel contribution to the IAEA FUMAC project, Tractebel has used the updated FRAPTRAN-TE-1.5 code for simulation of selected Halden LOCA tests IFA-650.9 and IFA-650.10, together with the improvement in the thermal hydraulic modelling by using the imposed thermal hydraulic boundary conditions from SOCRAT calculations and in the thermal boundary conditions (axial power profile, plenum temperature). In particular, the impacts of the model improvements such as the Quantum Technologies’ axial relocation model and errors corrections in the adapted FRAPTRAN-TE-1.5 version on the calculation results were identified and discussed. In addition, the statistical uncertainty and sensitivity analysis has been performed on the FRAPTRAN-TE-1.5 modelling of the selected Halden LOCA test IFA-650.10, which helped the identification of significant input parameters for LOCA fuel Behaviour modelling. The final objective is to apply the qualified fuel rod transient analysis codes FRAPCON/FRAPTRAN to develop an efficient methodology for assessing the performance and quantifying the margins for advanced technology fuel (ATF) designs under design basis accident conditions (in particular LOCAs). (author)

[en] The safe, reliable and economic operation of nuclear power reactor fleet has always been a top priority for worldwide nuclear industry. Typically, fuel rod Behaviour in accident conditions is one of the main concerns. After the severe Fukushima accidents, enhancing the accident tolerance of light water reactor (LWRs) became a hot issue in the world. Particularly, the fuel materials are expected to improve accident tolerance. Fuel performance is influenced by design parameters, physical properties and thermal hydraulic condition. In this report, the main result on FUMAC which is in order to better understand fuel rod Behaviour under LOCA condition and assess the predictive ability of the code, was present. Furthermore, take case IFA-650.10 for example, sensitivity and uncertainty study is performed to evaluate the ATF fuel system’s (UO2-BeO + ODS FeCrAl) performance under LOCA condition, and find out an optimal design by using the statistical tool DAKOTA coupling with fuel performance analysis code. (author)

[en] Following the experience that came from the Fukushima Daiichi accident, one possible way of reducing risk in a nuclear power plant operation would be the replacement of the existing fuel rod cladding material (based on zirconium alloys) by other materials which could fulfill the requirements of the accident tolerant fuel (ATF) concept. In this sense, ATF should be able to keep the current fuel system performance under normal operation conditions; moreover, it should present superior performance than the existing conventional fuel system (zirconium based alloys and uranium dioxide) under accident conditions. The most challenging and bounding accident scenarios for nuclear fuel systems in Pressurized Water Reactors (PWR) are Loss of Coolant Accident (LOCA) and Reactivity Initiated Accident (RIA), which are postulated accidents. This work addresses the performance of ATF using iron based alloys as cladding material under RIA conditions. The evaluation is carried out using modified versions of the coupled system FRAPCON/FRAPTRAN. These codes were modified to include the material properties (thermal, mechanical, and physics) of an iron based alloy, specifically FeCrAl alloy. The analysis is performed using data available in the open literature related to experiments using conventional PWR fuel system (zirconium based alloys and uranium dioxide). The results obtained using the modified code versions are compared to those of the actual existing fuel system based on Zircaloy-4 cladding using the original versions of the fuel performance codes (FRAPCON/FRAPTRAN). (author)

[en] This publication is the result of an IAEA technical meeting and reports on Member States’ capabilities in modelling, predicting and improving their understanding of the behaviour of nuclear fuel under accident conditions. The main results and outcomes of a coordinated research project (CRP) on this topic are also presented.

[en] Several models were integrated to the DIONISIO code within the framework of the IAEA Research Project “Fuel Modeling in Accident Conditions (FUMAC)”, to take account of accidental conditions, in particular the loss of coolant accidents (LOCA). A specially designed thermal-hydraulic subroutine provides a simplified description of the rod environment in normal or accidental conditions. The heat transfer coefficients that account for the different coolant regimes, in single or double phases, are activated as the corresponding conditions occur. The simulation of a considerable number of experiments has shown that, despite its simplicity this subroutine gives adequate predictions of the conditions in a vertical cooling channel, quite similar to those given by the thermal-hydraulic codes. The description of the fuel rod atmosphere is improved with the incorporation of this subroutine since it provides fairly realistic boundary conditions for the simulation of the fuel rod behavior, without requiring the intervention of external specific codes. Models of high temperature oxide growth (ZrO2) and hydrogen capture and release by the cladding in steam were also included. Moreover, the model of cladding creep predicts the conditions for ballooning and eventually, those for catastrophic failure (burst) and its localization. The calculation scheme makes a partition of the rod length into a number of segments defined by the user. In each segment the local conditions are considered to calculate, with the synchronous work of all the subroutines, the physical and chemical parameters in one representative pellet. Then, a description of the whole rod is obtained by coupling all the segments. This strategy has yielded accurate simulations of a wide variety of cases, either in normal or LOCA type conditions. Exhaustive comparisons were carried out with several thermal-hydraulic codes (COBRA-IV, RELAP5-Mod3.1, SOCRAT, ATHLET-Mod 1.1) and with a number of experiments like those of the IFA–650 series (-1,-2,-9,-10,-11), PUZRY, QUENCH-L0/L1 (for which a new working scheme was specially developed in DIONISIO), CORA-15, IAEA-SPE-4, among others. (author)

[en] Fuel heat-up, fuel degradation in an accident and the resultant fuel failure with release of the fission product (FP) into the primary system during Design Basis Accident (DBA) and Design Extension Conditions (DEC) are the key aspects to demonstrate the safety of the NPPs. Post Fukushima more emphasis is also laid on the development of the Accident Tolerant Fuel Designs (ATFD) to avoid fuel degradation and hydrogen generation. Terms like practically eliminated and ATF, need to be substantiated with physical and analytical evidence. Along with ATF development efforts should also be dedicated in prevention of loss of heat removal and/or quick restoration and lining up of the emergency coolant inventories with in the capabilities/survivability of the ATF. The aspects related to DBA and DEC fuel/core modelling are evolving specially the later one. The expectations, development, is also discussed here along with the comparison of ATF with the existing fuels. The development of ATF may be an iterative process shuffling from nuclear requirements to materials and safety performance while testing out of pile and in-pile. ATF aspect w.r.t reactivity loads is also elaborated for PHWRs. The developments on RIA and high burn-up fuel are also summarised. With improved ATF fuel performance improved MHT design and configuration are also envisaged. It is possible to configure MHT, ECCS and passive safety system in such a manner that the possibility of prolonged loss of cooling is reduced. (author)

[en] This paper presents the results of TRANSURANUS code testing for modelling the Behaviour of WWER nuclear fuel in LOCA accident conditions. The code was tested using data of the IAEA Coordinated Research Project Fuel Modelling in Accident Conditions (FUMAC). Fuel pins of Westinghouse and TVEL design are currently relevant for Ukrainian NPPs. Fuel from both vendors is presented in the FUMAC project. A part of the experimental data (MTA-EK data, IFA 650.10, IFA 650.11 and Studsvik 192&198) was simulated at SSTC NRS. Some of these data sets and KIT QUENCH-L1 set were calculated by other teams using the TRANSURANUS code (INRNE, Bulgaria; JRC, Germany). TRANSURANUS code demonstrated good capabilities for predicting the Behaviour of nuclear fuel rod cladding. The predicted cladding geometry and time of burst for both general types of cladding (Westinghouse and TVEL) show good correlations with the experimental data for such regimes. In addition, experimental data of the FUMAC project contained the results of post irradiation measurements after operation in a commercial reactor. These data were used to test TRANSURANUS code capabilities for modelling the fuel rod Behaviour in the core under burnup. (author)

[en] CEA is developing the ALCYONE fuel performance code for PWR fuel in the PLEIADES software environment. It is dedicated to normal, off-normal and accident conditions such as RIA and LOCA. CEA’s participation to the IAEA FUMAC CRP led to an improvement of the fuel modelling in LOCA conditions. Specific developments of the fuel performance code for the LOCA conditions have been done regarding cladding behaviour modelling, fission gas release and stress evaluation in the pellet before and during the tests. The improved code has been used to simulate some of the experiments of interest of the FUMAC project (IFA650.10 and Studsvik 192 LOCA test), the paper summarizes the results. For IFA650.10, the cladding outer temperature profile calculated with the SOCRAT code and provided to the participants of the FUMAC CRP has been used. The results obtained with ALCYONE are in a good agreement with the experimental data. In terms of uncertainty quantification, it seems that the uncertainty on the determination of the boundary conditions like cladding outer temperature results in a large uncertainty on the cladding deformation and the burst time. Recent developments have also been done in ALCYONE to improve the modelling of fuel behaviour in RIA conditions, in particular about fuel mechanical behaviour and the consequences on fission gas release. (author)

[en] The FINIX fuel behaviour module has been under development at VTT for six years for modelling traditional western light water reactor and Russian WWER fuel. The module is designed to be implemented in various other reactor safety analysis tools, such as neutronics and reactor dynamics codes. At VTT, it has been internally coupled with Serpent 2 to provide fuel temperatures for neutronics, and to thermal hydraulics and reactor dynamics codes TRAB1D, TRAB3D and HEXTRAN. FINIX will also be a part of the new VTT reactor analysis framework Kraken, the development of which is currently underway. In such couplings, the fuel temperature distribution is one of the most important parameters passed on to the host code. From the start, FINIX has been able to model the fuel temperature distribution in reactivity insertion accidents (RIA) comparably to established codes such as FRAPTRAN. The irradiated state was typically taken into account with a restart file from FRAPCON. Over the years, additional models have been implemented in the code, and long irradiation periods can now also be modelled with FINIX. The average error in temperature predictions across several Halden irradiations is 6.6%, and FINIX predicts typically 50 to 100 K higher temperatures at the highest temperatures compared to FRAPTRAN. Recently, the ability of FINIX to model loss-of-coolant accidents (LOCA) has been investigated. The main limitation of the current FINIX version is in mechanical modelling, as finite strain deformation cannot be accurately modelled, and the currently implemented failure models are rudimentary. Temperature predictions of FINIX were compared to FRAPTRAN and found to be close to FRAPTRAN until the blowdown phase, at which FINIX overpredicts the temperatures. FINIX LOCA predictions have also been compared to results calculated by FRAPTRAN for the Halden IFA-650.5 LOCA test. (author)

[en] Three dimensional characterized unit was used for TRISO particle with UN as the kernel to analyze the influence of PyC layers and internal pressure on the thermal mechanical performance of TRISO coated fuel particles by using the multi physics coupling software COMSOL. The influence of operation conditions including steady state and reactivity introduced accident (RIA) conditions on the thermal mechanical performance of TRISO particle was conducted. The results indicate that the structure integrity of SiC layer was maintained but the IPyC layer was failed in steady state condition. Thermal expansion is a dominant factor resulting in the loss of structure integrity of the TRISO coated fuel particle under RIA condition. (author)