[en] Reactivity initiated accidents (RIAs) are nuclear reactor accidents that involve an unwanted increase in fission rate and reactor power. Reactivity initiated accidents in power reactors may occur as a result of reactor control system failures, control element ejections or events caused by rapid changes in temperature or pressure of the coolant/moderator. our current understanding of reactivity initiated accidents and their consequences is based largely on three sources of information: 1) best-estimate computer analyses of the reactor response to postulated accident scenarios, 2) pulse-irradiation tests on instrumented fuel rodlets, carried out in research reactors, 3) out-of-pile separate effect tests, targeted to explore key phenomena under RIA conditions. In recent years, we have reviewed, compiled and analysed these 3 categories of data. The results is a state-of-the-art report on fuel behaviour under RIA conditions, which is currently being published by the OECD Nuclear Energy Agency. The purpose of this paper is to give a brief summary of this report

[en] The software Fire Dynamics Simulator aids in the simulation of different phenomena related to fire, such as smoke generation and distribution, flame propagation and surface temperature evaluation, among others. The room geometries of nuclear installations are not trivial, normally containing large amounts of equipment, cables and other obstacles. IDOM has devised a methodology by means of using open source software for the optimization of all the steps related to the simulation of any hypothetical fire in a nuclear installation with complex geometries. It has meant a reduction in time and errors for the model generation, its calculation by means of FDS and the analysis of the result. With a methodical erection of the geometry to be simulated using 3D modeling software, and through a series of scripts developed in Python language, it has been possible to automate part of the pre-processing of the FDS cases. As well, the post-process has been automatized. The methodology developed has been applied to generate a case with a complex geometry based on an experimental nuclear facility. It has been possible to generate the geometry requiring the mínimum of time, allowing the analyst to focus his time on more valuable tasks. This technique allows the slight modifications of the geometry required to create a quality mesh with the FDS software to be much less time-consuming, thus improving the results of the simulations. Currently, IDOM is working in the addition of extra physical models and the coupling FDS with other codes in order to be able to transport radionuclides in case of a fire in a contaminated area.

[en] In this interview the director of the French institute for radiation protection and nuclear safety (IRSN) details the research priorities to prevent future nuclear accidents. The first of all is to continue studying what happens when a reactor core melts and assesses whether we can act to regain full control of the nuclear fuel despite the lack of cooling. The second priority is to get a better understanding of what happens on a more or less contaminated area that is partially habitable in order to optimize its management. The second priority implies a real cooperation between experts in radioprotection and sociologists in order to avoid situations that generate a lot of anxiety in the local population when an accident happens as we have seen in the Fukushima accident. The ASTEC software simulates nuclear reactor functioning and has enabled IRSN to be the first organization in the world to propose the source term of the Fukushima accident. The Cabri research reactor has resumed activities in Cadarache in 2016, it will allow a real-time radiography of a spent fuel rod submitted to a reactivity surge. This experiment will allow a better understanding of the interaction of fuel pellets with cladding. The Mircom facility will provide a particle micro-beam that will be used to study the effect of a single electron impinging on a cell nucleus. (A.C.)

[en] Significant differences have been identified in loss of offsite power (LOSP or LOOP) event description, category, duration, and applicability between the LOSP events used in NUREG/CR-6890 and ENS'LOSP packages, which were based on EPRI LOSP reports with plant-specific applicability analysis. Thus it is appropriate to reconcile the LOSP data listed in the subject NUREG and EPRI reports. A cross comparison showed that 62 LOSP events in NUREG/CR-6890 were not included in the EPRI reports while 4 events in EPRI reports were missing in the NUREG. Among the 62 events missing in EPRI reports, the majority were applicable to shutdown conditions, which could be classified as category IV events in EPRI reports if included. Detailed reviews of LERs concluded that some events did not result in total loss of offsite power. Some LOSP events were caused by subsequent component failures after a turbine/plant trip, which have been modeled specifically in most ENS plant PRA models. Moreover, ENS has modeled (or is going to model) the partial loss of offsite power events with partial LOSP initiating events. While the direct use of NUREG/CR-6890 results in SPAR models may be appropriate, its direct use in ENS' plant PRA models may not be appropriate because of modeling details in ENS' plant-specific PRA models. Therefore, this paper lists all the differences between the data in NUREG/CR-6890 and EPRI reports and evaluates the applicability of the LOSP events to ENS plant-specific PRA models. The refined LOSP data will characterize the LOSP risk in a more realistic fashion. (authors)

[en] KAERI is developing a localized severe accident code, MIDAS, based on MELCOR. MELCOR uses pointer variables for a fixed-size storage management to save the data. It passes data through two depths, its meaning is not understandable by variable itself. So it is needed to understand the methods for data passing. This method deteriorates the readability, maintainability and portability of the code. As a most important process for a localized severe accident analysis code, it is needed convenient method for data handling. So, it has been used the new features in FORTRAN90 such as a dynamic allocation for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring for each package was developed and tested. And then integration of each restructured package was being processed one by one. In this paper, the integrating scope includes the BUR, CF, CVH, DCH, EDF, ESF, MP, SPR, TF and TP packages. As most of them use data within each package and a few packages share data with other packages. The verification was done through comparing the results before and after the restructuring

[en] A study was performed using a CATHENA based generic model of a CANDU6 reactor to demonstrate the sensitivity of system response to single and two-stage break opening characteristics following a Reactor Inlet Header Large Break LOCA. Figures of merit for the analysis were the power pulse amplitude and the 5 second integrated normalized power. The analysis demonstrates in a qualitative manner a threshold time for 'instantaneous' break opening, and a threshold initial break size above which shutdown system trip times remain unchanged. The analysis also attempts to establish the relative impacts of initial and second stage opening times on figures of merit. (author)

[en] The trip delay time that is determined by using RFSP-IST is applied to the Loss of Regulation (LOR) simulations of Wolsong-1, which is scheduled to undergo a major refurbishment project. The safety analyses required for licensing after refurbishment are currently being carried out using the update-to-date Canadian Industry Standard Toolset code suite. Since the POINTSIM program used in the past to calculate trip delay time is not included in IST, the RFSP-IST code is being mandatorily used in place of the POINTSIM functions. The CATHENA LOR simulation results obtained based upon the RFSP-IST determined trip delay times clearly justify the use of procedures that are developed to run *CERBERUS, *INTREP and *TRIP_-TIME modules in the context of determining trip delay time. (author)

[en] Mechanical consequences which might be caused by core disruptive accident (CDA) are one of the major concerns in safety of fast reactors (FRs). Once a severe re-criticality occurs, the core materials are vaporized creating CDA bubbles which consists of fuel vapor, steel vapor, sodium vapor and fission gases. The high-pressured CDA bubbles expand rapidly and drive a sodium slug in the upper plenum. The upward-accelerated sodium slug compresses the cover-gas region with increase of the pressure. This pressure increase might cause to threaten the reactor vessel integrity. To evaluate the mechanical energy affecting the boundary integrity, the energy conversion from thermal energy to mechanical energy plays an important role. This paper describes model validation study of ASTERIA-FBR application to the thermal-to-mechanical energy conversion process, focusing on calculation models through the THINA test simulation. As a result, it was confirmed that the energy conversion process and its ratio were in good agreements with the experimental results. The mechanism of CDA bubble expansion and the uncertainty brought by the calculation models were also discussed. (author)

[en] REDOU (REleases Due to Oxidation of UO_2) 2.0 is an AECL® computer code for modelling the iodine release during the air oxidation of UO_2 fuel. REDOU 2.0 calculates the fractional release of grain-matrix iodine from within a UO_2 fragment that is exposed to air due to fuel damage during a postulated end-fitting failure accident. REDOU 2.0 has been extensively verified, validated and fully documented in accordance with the Canadian software quality assurance requirements. A brief description of the code and a summary of the validation results are given in this paper. (author)

[en] Many evolving nuclear energy technologies use advanced predictive multi-scale, multi-physics modeling and simulation (MS) capabilities to reduce the cost and schedule of design and licensing. A new methodology is needed for the validation of these predictive tools. The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification-steps similar to the components of the traditional US Nuclear Regulatory Commission (NRC) licensing approach, with the exception of the calibration step. An enhanced calibration concept is introduced here, and is accomplished through data assimilation. The goal of this methodology is to enable best-estimate prediction of system behaviors in both normal and safety-related environments. This goal requires the additional steps of estimating the domain of validation, and quantification of uncertainties, allowing for the extension of results to areas of the validation domain that are not directly tested with experiments. These might include the extension of the MS capabilities for application to full-scale systems. The new methodology suggests a formalism to quantify an adequate level of validation (predictive maturity) with respect to existing for data, so that required new testing can be minimized, saving cost by demonstrating that further testing will not enhance the quality of the predictive tools. The proposed methodology is at a conceptual level. The document is an extended abstract