Nuclear Safety Institute Research Articles

Results of SOCRAT code development, validation and applications for NPP safety assessment under severe accidents

Abstract After Fukushima Daiichi accident the importance of software tools that enable the NPP safety assessment for the overall sequence of severe accident processes has revived. In accordance with this issue Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) in cooperation with the nuclear industry organizations is developing an integrated software package SOCRAT that is intended for numerical analysis of the behaviour of LWR at various stages of design extension conditions, including reactor core degradation, RPV failure and long-term melt retention in a core catcher. This paper presents a review of current state of SOCRAT models development, validation and practical applications, with emphasis on best-estimate mechanistic description of relevant physical processes.

Coupled Experimental and Computational Approach for CABRI Power Transients Analysis

CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute and operated by Commissariat a l’Energie Atomique et aux Energies Alternatives at the Cadarache Research Center. It is designed to study fuel behavior under reactivity-initiated accident conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in the so-called transient rods inside the reactor core. The CABRI reactivity injection system allows us to generate structured transients based on specific sequences of depressurization. For such transients, the time difference between the openings of two valves of the reactivity injection system has an important impact on the power pulse shape. A kinetic point code, SPARTE, was developed in order to replace the older DULCINEE code dedicated to the modeling and prediction of CABRI power transients. The SPARTE code includes new models of 3He depressurization based on CFD calculations, variable Doppler coefficient based on Monte Carlo calculations, and variable axial neutron flux profile. The density and Doppler models have a large impact on power transient prediction. For low initial pressure transients, the major uncertainty comes from the reactivity injected by the 3He depressurization. For high initial pressure transients, the 3He heating during the power pulse (“TOP effect”) is responsible of an additional injection of reactivity that needs to be modeled precisely.

Modeling of the 3He Density Evolution Inside the CABRI Transient Rods During Power Transients

CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under reactivity-initiated accident conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly model gas density evolution in transient rods during a power transient. The 3He depressurization was studied by computational fluid dynamics (CFD) calculations validated by measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Two methods were identified to evaluate the gas density evolution: CFD calculations and reverse point kinetics (PKs). The first one consists in adding a heat source in transient rods based on the experimental power conversion. The second one consists in using the measured power by boron ionization chambers to evaluate the net reactivity by a reverse PKs method and to subtract the reactivity feedbacks calculated with the DULCINEE multiphysics code.

The consumption of locally-produced foodstuff: an important parameter of the radiological exposure for population living in the vicinity of French nuclear facilities

Either in normal or post-accidental situations, the ingestion of locally-produced foodstuffs, potentially contaminated, is one of the main exposures to the population living in the vicinity of nuclear facilities. For this reason, the French Institute for Radioprotection and Nuclear Safety has developed a specific methodology and has carried out surveys on dietary habits of some populations concerned. This study presents the methodology used for local food surveys, the amount consumed and how much food is produced locally. The results from the 6 surveys already done show that food surveys even if conducted with a specific aim as impact assessments provide further in-depth understanding of the food habits of local residents. There are still local specific population groups that have a high level of “home-consumption” for products more sensitive to contamination. For leafy vegetables and milk (products highly sensitive to contamination), home-consumption rates vary widely from one region to another from 76 to more than 90% and from 1 to 17% respectively. Specifically for the Andra survey carried out in a rural area in the North-East of France, self-sufficiency had reached 100% for a few products (game meat for example).

実用発電用原子炉運転技術におけるリーダーシップ能力の向上を目指して(安全における人材育成・教育指導)

実用発電用原子炉運転技術におけるリーダーシップ能力の向上を目指して(安全における人材育成・教育指導)

Study of the linearity of CABRI experimental ionization chambers during RIA transients

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center and funded by the French Nuclear Safety and Radioprotection Institute (IRSN). For the purpose of the CABRI International Program (CIP), operated and managed by IRSN under an OECD/NEA framework it has been refurbished since 2003 to be able to provide experiments in prototypical PWR conditions (155 bar, 300 °C) in order to study the fuel behavior under Reactivity Initiated Accident (RIA) conditions. This paper first reminds the objectives of the power commissioning tests performed on the CABRI facility. The design and location of the neutron detectors monitoring the core power are also presented. Then it focuses on the different methodologies used to calibrate the detectors and check the consistency and co-linearity of the measurements. Finally, it presents the methods used to check the linearity of the neutron detectors up to the high power levels (~20 GW) reached during power transients. Some results obtained during the power tests campaign are also presented.

Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN) and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

Development, Automation, and Validation of a Numerical Methodology to Assess the TOP Onset for the RIA CABRI Experiments

The CABRI facility is an experimental pulse nuclear reactor funded by the French Nuclear Safety and Radioprotection Institute and operated by the French Atomic Energy Commission. It is designed to study the behavior of fuel rods at high burnup under reactivity-initiated accident (RIA) conditions, such as a control rod ejection. The distinctive feature of this reactor is its reactivity injection system. The fast depressurization into a discharge tank of 3He (strong neutron absorber) previously introduced inside 96 tubes (so-called transient rods) located among the fuel rods allows us to create a power burst from 100 kW to 20 GW with a full-width at half-maximum of 10 to 80 ms. The total energy deposit in the tested rod is adjusted by dropping the control and safety rods after the power transient. The neutron flux is measured online by compensated boron chambers located outside the reactor and operated in the current mode. These neutron detectors are calibrated during a commissioning phase thanks to standards given by a conventional heat balance. To assess the energy released into the test rod, we had to integrate the driver core power signal measured online. Thus, the beginning of the transient, called transient overpower (TOP) onset, has to be estimated. The TOP onset of a transient test is defined as the instant of the beginning of the test. It is determined by experimentalists during the processing phase. It corresponds to the beginning of the increase of the neutron detector signal, measured by the compensated boron chamber devices sufficiently sensitive at low current levels. So far, the choice of this instant has been realized by a visual choice zooming in the zone of interest, which may induce some shift according to experimentalists. In an attempt to overcome this issue, we put forward in this paper a theoretical method of determination to calculate the TOP onset. The main asset of the method is to formalize the TOP onset determination. Furthermore, it provides the possibility of associating an uncertainty, which is impossible by the manual process. The methodology relies on the fact that at the beginning of the RIA transient, the neutron flux at any point of the reactor core undergoes an exponential evolution as a function of the time. Then, a logarithmic transform allows us to show that the search for the TOP onset is equivalent to solving a nonlinear regression. The methodology has been validated in the last 14 experiments. Moreover, the reactor restarted in October 2015 and now gives us the opportunity to apply this methodology on signals recently acquired and pertaining to the power commissioning phase with a view to preparing the experiment foreseen next year.

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.

Transposing an active fault database into a seismic hazard fault model for nuclear facilities – Part 1: Building a database of potentially active faults (BDFA) for metropolitan France

Abstract. The French Institute for Radiation Protection and Nuclear Safety (IRSN), with the support of the Ministry of Environment, compiled a database (BDFA) to define and characterize known potentially active faults of metropolitan France. The general structure of BDFA is presented in this paper. BDFA reports to date 136 faults and represents a first step toward the implementation of seismic source models that would be used for both deterministic and probabilistic seismic hazard calculations. A robustness index was introduced, highlighting that less than 15 % of the database is controlled by reasonably complete data sets. An example of transposing BDFA into a fault source model for PSHA (probabilistic seismic hazard analysis) calculation is presented for the Upper Rhine Graben (eastern France) and exploited in the companion paper (Chartier et al., 2017, hereafter Part 2) in order to illustrate ongoing challenges for probabilistic fault-based seismic hazard calculations.

An investigation of tritium transfer in reactor loops

The work is devoted to the important task of the numerical simulation and analysis of the tritium behaviour in the reactor loops. The simulation was carried out by HYDRA-IBRAE/LM code, which is being developed in Nuclear safety institute of the Russian Academy of Sciences. The code is intended for modeling of the liquid metal flow (sodium, lead and lead-bismuth) on the base of non-homogeneous and non-equilibrium two-fluid model. In order to simulate tritium transfer in the code, the special module has been developed. Module includes the models describing the main phenomena of tritium behaviour in reactor loops: transfer, permeation, leakage, etc. Because of shortage of the experimental data, a lot of analytical tests and comparative calculations were considered. Some of them are presented in this work. The comparison of estimation results and experimental and analytical data demonstrate not only qualitative but also good quantitative agreement. It is possible to confirm that HYDRA-IBRAE/LM code allows modeling tritium transfer in reactor loops.

Simulation of hydromechanical behaviour of bentonite seals for containment of radioactive wastes

Geological disposal of radioactive wastes relies on a multiple barrier system to provide long-term containment and isolation of the wastes. The excavation of the repository creates openings and disturbed zones in the host rock formations that need to be properly sealed. Bentonite-based materials are being considered worldwide as a preferred type of sealing material, since they possess desirable characteristics such as low permeability, high sorption capability, and swelling potential allowing them to close internal cracks and gaps at interfaces with other materials. The French Institute for Radiation Protection and Nuclear Safety (IRSN) has led an experimental program consisting of a series of laboratory and large in situ experiments to assess the hydromechanical behaviour of bentonite seals. The experiments consisted of the forced re-saturation of pre-fabricated blocks of bentonite–sand mixture, with technological voids between bentonite seals and the walls of the steel cell (in the laboratory tests) and between bentonite seals and the host rock (in the in situ experiment). Relative humidity and total stress were monitored during both tests. The Canadian Nuclear Safety Commission (CNSC) collaborated with Geofirma Engineering, IRSN, and Commissariat à l’énergie atomique (CEA) to develop a mathematical model to simulate the experiments. The model was developed within the framework of poromechanics, with the inclusion of partial saturation characteristics and swelling potential to simulate the behaviour of the bentonite-based material. The model results were in good agreement with the experimental measurements for relative humidity and swelling stresses. The model also predicted the closure of technological voids and gaps due to swelling. Although swelling into the technological voids leads to an increase in permeability, that permeability remains low and insignificant from a safety perspective.

Development of a viscoelastoplastic model for a bedded argillaceous rock from laboratory triaxial tests

Argillaceous rocks are candidate host and (or) cap formations for the geological disposal of nuclear wastes in many countries, including Canada, France, and Switzerland. The understanding of the long-term mechanical behaviour of such rocks is an essential requirement for the assessment of their performance as a barrier against radionuclide migration. The French Institute for Radiological Protection and Nuclear Safety (IRSN) operates an Underground Research Laboratory (URL) in Tournemire, France, in a rock formation known as the Tournemire shale. Many types of experiments are conducted at the Tournemire URL to better understand the physical and chemical behaviour of this shale and its interaction with seal materials intended to be used in the geological disposal of radioactive wastes. The Canadian Nuclear Safety Commission (CNSC) collaborates with the IRSN and CanmetMINING laboratories to perform experimental and theoretical research on the mechanical behaviour of the Tournemire shale. Using data from creep tests, and monotonic and cyclic triaxial tests performed at CanmetMINING Laboratories, we developed constitutive relationships for the mechanical behaviour of the Tournemire shale. The model is based on the theory of plasticity, and takes into consideration the inherent anisotropy due to the existence of bedding planes, hardening behaviour before the peak strength, and viscosity.

Совершенствование инфраструктуры и материально-технической базы ФГУП «Атомфлот» для обеспечения экологической безопасности на северо-западе России

The article reviews FSUA Atomflot assets and facilities used to manage spent nuclear fuel (SNF) and radioactive wastes (RW), their modernization and development, as well as sources of the aforementioned processes financing, which include both Russian Federation and International funding. FSUE Atomflot participation in the environmental rehabilitation of the Northwest of Russia is analyzed, part of which is the status of retired atomic icebreakers and special maintenance vessels decommissioning. The key direction of FSUE Atomflot policy is safe handling of spent nuclear fuel and radioactive wastes accumulated by the Company itself and other organizations in Murmansk region. Modernization of FSUE Atomflot infrastructure and material and technical resources is meant to solve this task. FSUE Atomflot operates state atomic icebreaking fleet, which includes commercial operation, maintenance, modernization, repair and decommissioning of nuclear-powered icebreakers and special maintenance vessels. Handling of spent nuclear fuel and radioactive wastes that are accumulated during the process of operation through the assets life cycle is an important part of the Company’s activity. The important role in the provision of environmental safety belongs to the Strategic Master Plan for Decommissioning of Retired Russian Nuclear Fleet and Environmental Rehabilitation of Its Supporting Infrastructure in the Northwest of Russia, which was ordered by Rosatom and developed under scientific guidance by Nuclear Safety Institute of Russian Academy of Science. The article further describes the process of SNF and RW handling by FSUE Atomflot, brings up the statistical data of the materials management and transportation. Second part of the article describes the decommissioning of atomic icebreakers and maintenance vessels as part of the environmental safety provision process. The current status of the process, results and development plans are resumed.

Comparison of the results of 85Kr transport modeling with the ACURATE field experiment data

Currently the Nuclear Safety Institute of the Russian Academy of Sciences (NSI RAS) jointly with the Hydrometcenter of Russia is developing the system for forecasting the transfer of radio-active substances in the atmosphere in case of radiation accidents at Russian nuclear power plants. The operation of the system is based on the numerical hydrodynamic model which allows forecasting meteorological parameters and is coupled with the mesoscale dispersion model of the transfer ofradioactive substances in the atmosphere. The results are presented of 85Kr transport modeling under the conditions of the ACURATE experiment with three transport models: FLEXPART, HYSPLIT, and the model from the NOSTRADAMUS software package. It is demonstrated that all three Lagrangian models can give a qualitative description of concentration fields from the ACURATE experiment with the best value of the RANK metric (2.5) based on three statistics.

Evaluation of RayXpert®for shielding design of medical facilities

In a context of growing demands for expert evaluation concerning medical, industrial and research facilities, the French Institute for radiation protection and nuclear safety (IRSN) considered necessary to acquire new software for efficient dimensioning calculations. The selected software is RayXpert® . Before using this software in routine, exposure and transmission calculations for some basic configurations were validated. The validation was performed by the calculation of gamma dose constants and tenth value layers (TVL) for usual shielding materials and for radioisotopes most used in therapy (Ir-192, Co-60 and I-131). Calculated values were compared with results obtained using MCNPX as a reference code and with published values. The impact of different calculation parameters, such as the source emission rays considered for calculation and the use of biasing techniques, was evaluated.

Ionization Quenching Factor measurement of 1 keV to 25 keV protons in Isobutane gas mixture

The French Institute for Radiation protection and Nuclear Safety (IRSN) is providing reference neutron fluence energy distribution at its standard monoenergetic neutron fields, produced at the AMANDE facility. The neutron energy is assessed by measuring the recoil nuclei energy in a μTPC detector, the LNE-IRSN/MIMAC detector. The knowledge of the ionization quenching factor (IQF) is fundamental to determine the kinetic energy of the recoil nuclei. For some various gases and pressures, discrepancies of about 15% were observed between IQF calculations using the SRIM software and experimental measurements. No data are available for the iC4H10 + 50% CHF3 gas mixture which are used for measurements from a few keV up to 565 keV neutron energies in the μTPC detector. The experimental determination of the IQF is of primary importance to provide reference neutron fluence energy distribution. After a short description of the experimental set-up, this paper presents the first results of the IQF measurements in a iC4H10 + 50% CHF3 gas mixture in the energy range 1 keV – 25 keV.

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

In-depth characterisation of a mixture composed of powder/pellets MX80 bentonite

Mixtures made up of bentonite powder and pellets are a possible candidate for making sealing plugs used in deep radioactive waste disposal due to their low permeability, high swelling capacity, favourable properties with respect to radionuclide retention and operational advantages in terms of placement in situ, which is much easier than that of pre-compacted bricks of bentonite/sand mixture. It is therefore essential to better understand their hydro-mechanical behaviour to optimize the design of the repository. In this context, the French Institute for Radiation Protection and Nuclear Safety (IRSN) has launched the SEALEX project (SEALing performance EXperiments) in which this work has been conducted.Once the initially heterogeneous unsaturated powder/pellet (80/20) MX80 bentonite mixture is put in place, these sealing materials will be subject to coupled hydro-mechanical loadings: hydration due to the infiltration of pore water from the natural barrier and mechanical confinement resulting from the engineered barriers. The present work focuses on the different scales of the material: at the macroscopic scale, it is characterised by a heterogeneous distribution of pellets and powder of bentonite; at the microscopic scale, it is studied by several techniques (MIP, μ-CT observations and SEM). From MIP results, a typical bimodal distribution was found for both pellet and powder. From μ-CT and SEM observations, a heterogeneity was revealed in the internal structure of the pellet: heterogeneous density distribution of the clay minerals and presence of several high density elements.

Estimation and mapping of uranium content of geological units in France

In France, natural radiation accounts for most of the population exposure to ionizing radiation. The Institute for Radiological Protection and Nuclear Safety (IRSN) carries out studies to evaluate the variability of natural radioactivity over the French territory. In this framework, the present study consisted in the evaluation of uranium concentrations in bedrocks. The objective was to provide estimate of uranium content of each geological unit defined in the geological map of France (1:1,000,000). The methodology was based on the interpretation of existing geochemical data (results of whole rock sample analysis) and the knowledge of petrology and lithology of the geological units, which allowed obtaining a first estimate of the uranium content of rocks. Then, this first estimate was improved thanks to some additional information. For example, some particular or regional sedimentary rocks which could present uranium contents higher than those generally observed for these lithologies, were identified. Moreover, databases on mining provided information on the location of uranium and coal/lignite mines and thus indicated the location of particular uranium-rich rocks. The geological units, defined from their boundaries extracted from the geological map of France (1:1,000,000), were finally classified into 5 categories based on their mean uranium content. The map obtained provided useful data for establishing the geogenic radon map of France, but also for mapping countrywide exposure to terrestrial radiation and for the evaluation of background levels of natural radioactivity used for impact assessment of anthropogenic activities.