Institut De Radioprotection Et De Surete Nucleaire Research Articles

Characterization of closed-doors electrical cabinet fires in compartments

An important cause of fire departure in industrial facilities is due to electrical origin and particularly to electrical cabinets. The investigation of such fires has been scarce up to now and has been investigated exclusively in the nuclear industry. The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) conducted a large number of experiments involving electrical cabinets burning either under a calorimetric hood or inside a mechanically ventilated compartment to investigate this topic. Calorimetric hood experiments demonstrated that the most important parameter is the size of the vents of the cabinet and that the time to flashover depends on many factors and seems somewhat random with regard to the observable parameters. The influence of the compartment on the fire behavior depends on the temperature of the surrounding atmosphere of the cabinet and on the oxygen content in the compartment at the level of the inlet vent of the cabinet. The compartment strongly impacts the pyrolysis of the combustible, affecting the fire duration, but has a weak effect on the Heat Release Rate (HRR). Experiments were usually remarkably reproducible, opening the way to a phenomenological description of this type of fire. A semi-empirical model based on the coupled solution of ventilation limit and excess pyrolysate could then be developed. This model was introduced in a zone code, and an ad-hoc modeling of the fire extinction based on a critical surfacic mass loss rate is proposed. The major features of the compartment fire experiments such as characteristic HRR and fire duration could then be reproduced with acceptable error. The development of such a semi-empirical model is a common practice in fire safety engineering concerned with complex combustibles.

Mesoscale dispersion of85Kr in the vicinity of the AREVA La Hague reprocessing plant

The Institut de Radioprotection et de Surete Nucleaire (IRSN) performed a series of air sampling campaigns at mesoscale distances (10-80km) from the AREVA La Hague reprocessing plant (north west of France) between 2007 and 2009. These samples were collected in order to test and optimise a technique to measure low krypton-85 ( 85 Kr) air concentrations, and to investigate the performance of three atmospheric dispersion models (RIMPUFF, HySplit, and ADMS). This paper presents 85 Kr air concentrations measured at both land and sea locations. In addition, this paper compares the measured 85 Kr air concentrations, which varied from 2 to 8000Bqm −3 , with the predictions from the atmospheric dispersion models. During stable wind conditions, the dispersion models make reasonable estimates of the 85 Kr field measurements. In contrast, the models fail to accurately predict temporal peaks in concentration during periods of rapid and large changes in wind speed and/or wind direction.

Quantifying differences between computational results and measurements in the case of a large-scale well-confined fire scenario

Abstract The objective of this work was to quantify comparisons between several computational results and measurements performed during a pool fire scenario in a well-confined compartment. This collaborative work was initiated under the framework of the OECD fire research program and involves the most frequently used fire models in the fire community, including field and zone models. The experimental scenario was conducted at the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and deals with a full-scale liquid pool fire in a confined and mechanically ventilated compartment representative for nuclear plants. The practical use of different metric operators and their ability to report the capabilities of fire models are presented. The quantitative comparisons between measurements and numerical results obtained from “open” calculations concern six important quantities from a safety viewpoint: gas temperature, oxygen concentration, wall temperature, total heat flux, compartment pressure and ventilation flow rate during the whole fire duration. The results indicate that it is important to use more than one metric for the validation process in order to get information on the uncertainties associated with different aspects of fire safety.

Neutron Commissioning in the New CABRI Water Loop Facility

The CABRI experimental reactor is located at the Cadarache nuclear research center, southern France. It is operated by the Atomic Energy Commission (CEA) and devoted to Institut de Radioprotection et de Surete Nucleaire (IRSN) safety programs. It has been successfully operated during the last 30 years, enlightening the knowledge of FBR and LWR fuel behavior during Reactivity Insertion Accident (RIA) and Loss Of Coolant Accident (LOCA) transients in the frame of Institut de Protection et de Surete Nucleaire (IPSN) and now IRSN programs devoted to reactor safety. This operation was interrupted in 2003 to allow for a whole facility renewal program. The main goal of this reconstruction project is to meet thermal hydraulics parameters identical to LWR standard and downgraded conditions, in particular for the need of the CABRI International Programme (CIP) carried out by IRSN under the OECD umbrella. For this, the sodium cooled experimental loop is now being replaced by a pressurized water loop. The distinctive features of CABRI core allow for a specific neutron commissioning program. This paper presents the parameters to be measured with associated experimental methods and also briefly introduces the instrumentation set up.

Modeling of Iodine Radiochemistry in the ASTEC Severe Accident Code: Description and Application to FPT-2 PHEBUS Test

In the case of a hypothetical severe accident in a nuclear power plant, iodine is one of the fission products of major importance. It may be present in various gaseous forms that could be released to the environment, impacting population health. In such a case, the amount released (the so-called “source term”) has to be estimated in order to help the safety authorities protect the population from radiological consequences. This estimation is one of the main objectives of the Accident Source Term Evaluation Code (ASTEC) that is developed jointly by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and the German institute Gesellschaft für Anlagen- und Reaktorsicherheit. ASTEC is composed of various modules able to model the nuclear reactor behavior during an accident. One of these modules, named IODE, predicts iodine behavior in the reactor containment. It is able to model the kinetics of about 35 chemical reactions and mass transfer processes. IODE is validated against separate effect tests, semi-integral experiments, and integral experiments. This paper presents the experimental phenomena that would take place in reactor containment in the case of a severe accident. Then, IODE is used to model the experimental gaseous concentration of organic and inorganic iodine in the PHEBUS FPT-2 test carried out by IRSN. The comparison of experimental data and the modeling show a general good agreement for inorganic iodine even if some differences are evidenced. For organic iodides the modeling is not satisfying. These differences might be explained by the deficiencies of some models and by some assumptions that still have to be validated by dedicated experiments.

Validation Status of the ASTEC Integral Code for Severe Accident Simulation

The French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit mbH (GRS) have been jointly developing for several years a system of calculation codes (or “integral” code), ASTEC (Accident Source Term Evaluation Code), to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor from the initiating event through the possible radiological release of fission products out of the containment, the so-called “source term.” Very intensive validation work has been performed in recent years by IRSN and GRS on the V1 versions by comparison of code calculations with results of more than 160 international experiments. Complementary validation was performed by 30 partners of the SARNET European Network of Excellence in the 6th Framework Programme of the European Commission, where ASTEC is considered the European reference code. The global status of validation is good for most phenomena, as shown by several examples that are described in this paper, and even very good on fission product behavior. The main need for modeling improvement concerns reflooding of a degraded core, due to the lack in ASTEC V1 of any dedicated model, and intensive efforts will focus on this topic in the next years. Molten core concrete interaction models are at the state of the art, but new experiments under way in the international frame and a better understanding of physical mechanisms are necessary to make further progress. Version V2.0 of the new ASTEC series, released mid-2009, takes benefit of the previous very intensive validation of the ICARE2 IRSN mechanistic code since its core degradation models have now been implemented. Validation will continue in the SARNET network from 2009 to 2013.

Fuel Pin Behavior up to Cladding Failure under Pulse-Type Transient Overpower in the CABRI-FAST and CABRI-RAFT Experiments

In the CABRI-FAST and CABRI-RAFT programs within a collaboration with the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and Forschungszentrum Karlsruhe (FZK), five pulse-type transient overpower tests were performed in order to study fuel pin behavior and failure condition in the Unprotected Loss-of-Flow (ULOF) accident. In these tests, two types of low-smear-density fuels irradiated in the French Phénix reactor at different burn-up levels were used so that an experimental database extension from the former CABRI-1 and CABRI-2 programs can be obtained. Pin failure took place in three of these tests giving information on the failure threshold. In two tests, no pin failure took place and useful information related to the transient fuel behavior up to failure and failure mechanism was obtained. These test results were interpreted through detailed analysis of experimental data and PAPAS-2S code calculations. In these calculations, pretransient fuel characteristics obtained from the sibling fuels were reflected, such that the uncertainty of the boundary condition can be minimized. Through the comparison among these tests and formerly existing CABRI tests, generalized understanding on the transient fuel behavior was obtained. It was concluded that the low-smear-density fuel mitigates cavity pressurization, thereby enhancing the margin-to-failure. It was also understood that this failure-thresholdenhancing capability is dependent on the type of transient.

Reactor cooling systems thermal–hydraulic assessment of the ASTEC V1.3 code in support of the French IRSN PSA-2 on the 1300 MWe PWRs

The French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe PWRs. This PSA-2 study is relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, a wide-ranging series of comparisons with the French best-estimate thermal–hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe-accident scenarios. The present paper details 4 out of the 14 studied scenarios: a 12-in. cold leg Loss of Coolant Accident (LOCA), a 2-tube Steam Generator Tube Rupture (SGTR), a 12-in. Steam Line Break (SLB) and a total Loss of Feed Water scenario (LFW). The thermal–hydraulic behavior of the primary and secondary circuits is thoroughly investigated and compared to the CATAHRE 2 V2.5 results. The ASTEC results of the core degradation phase are also presented. Overall, the thermal–hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results.

Investigation on primary side oriented accident management measures in a hypothetical station blackout scenario for a VVER-1000 pressurized water reactor

Abstract As a consequence of a total loss of AC power supply (station blackout) at a VVER-1000 leading to unavailability of major active safety systems, the safety criteria ensuring the safe operation of the nuclear power plant would be violated and core heat-up with possible core degradation could occur. A dedicated accident management measure (primary side depressurization) can be applied to reduce the primary pressure and to activate the injection from the passive emergency core cooling systems (accumulators). The analyses presented in this paper are aiming at both a detailed investigation of the accident sequence, taking into account the depressurization of the primary circuit, and the possibilities to prevent or at least to mitigate a damage of the reactor core so as to gain additional time for taking necessary countermeasures. The analyses are performed using the codes ASTEC and ATHLET developed by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH).

Fuel Pin Behavior up to Cladding Failure under Pulse-Type Transient Overpower in the CABRI-FAST and CABRI-RAFT Experiments

In the CABRI-FAST and CABRI-RAFT programs within a collaboration with the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and Forschungszentrum Karlsruhe (FZK), five pulse-type transient overpower tests were performed in order to study fuel pin behavior and failure condition in the Unprotected Loss-of-Flow (ULOF) accident. In these tests, two types of low-smear-density fuels irradiated in the French Phénix reactor at different burn-up levels were used so that an experimental database extension from the former CABRI-1 and CABRI-2 programs can be obtained. Pin failure took place in three of these tests giving information on the failure threshold. In two tests, no pin failure took place and useful information related to the transient fuel behavior up to failure and failure mechanism was obtained. These test results were interpreted through detailed analysis of experimental data and PAPAS-2S code calculations. In these calculations, pretransient fuel characteristics obtained from the sibling fuels were reflected, such that the uncertainty of the boundary condition can be minimized. Through the comparison among these tests and formerly existing CABRI tests, generalized understanding on the transient fuel behavior was obtained. It was concluded that the low-smear-density fuel mitigates cavity pressurization, thereby enhancing the margin-to-failure. It was also understood that this failure-thresholdenhancing capability is dependent on the type of transient.

Mechanical characteristics of fuel rod claddings in transport conditions

AbstractA synthesis on the mechanical characteristics of unirradiated and irradiated fuel rod claddings was performed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) in order to have reference data for the assessment of the safety demonstrations in normal and accident conditions of transport required by the procedure of package licensing. Indeed, the transport conditions correspond to a range of cladding temperatures (200–550°C) which is only partly covered by the data acquired within the framework of the safety demonstration relative to the reactor normal operating conditions, especially beyond 400°C. This work concerned Zircaloy-4 cladding material (Zry-4) and M5TM. Data about mechanical properties (elastic and ductile properties, creep behaviour), oxidation (in reactor and under air during transport), hydrides and fracture toughness have been collected and synthesised. The laws presented in the document can be used to obtain orders of magnitude of oxide layer thickness, hydrogen...

ASTEC application to in-vessel corium retention

This paper summarizes the work done in the SARNET European Network of Excellence on Severe Accidents (6th Framework Programme of the European Commission) on the capability of the ASTEC code to simulate in-vessel corium retention (IVR). This code, jointly developed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit mbH (GRS) for simulation of severe accidents, is now considered as the European reference simulation tool. First, the DIVA module of ASTEC code is briefly introduced. This module treats the core degradation and corium thermal behaviour, when relocated in the reactor lower head. Former ASTEC V1.2 version assumed a predefined stratified molten pool configuration with a metallic layer on the top of the volumetrically heated oxide pool. In order to reflect the results of the MASCA project, improved models that enable modelling of more general corium pool configurations were implemented by the CEA (France) into the DIVA module of the ASTEC V1.3 code. In parallel, the CEA was working on ASTEC modelling of the external reactor vessel cooling (ERVC). The capability of the ASTEC CESAR circuit thermal-hydraulics to simulate the ERVC was tested. The conclusions were that the CESAR module is capable of simulating this system although some numerical and physical instabilities can occur. Developments were then made on the coupling between both DIVA and CESAR modules in close collaboration with IRSN. In specific conditions, code oscillations remain and an analysis was made to reduce the numerical part of these oscillations. A comparison of CESAR results of the SULTAN experiments (CEA) showed an agreement on the pressure differences. The ASTEC V1.2 code version was applied to IVR simulation for VVER-440/V213 reactors assuming defined corium mass, composition and decay heat. The external cooling of reactor wall was simulated by applying imposed coolant temperature and heat transfer coefficient (HTC). The obtained results (pool temperatures, heat flux distribution, reactor wall ablation) were compared with available predictions of other codes. The agreement was correct, in particular on the shape and depth of ablation, as well as the maximum heat flux in case of a thick metallic layer, while ASTEC calculated a lower maximum heat flux for a thin metallic layer.

Fission Product Experimental Program: Validation and Computational Analysis

From 1998 to 2004, a series of critical experiments referred to as the fission product (FP) experimental program was performed at the Commissariat á l’Energie Atomique Valduc research facility. The experiments were designed by Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and funded by AREVA NC and IRSN within the French program supporting development of a technical basis for burnup credit validation. The experiments were performed with the following six key fission products encountered in solution either individually or as mixtures: 103Rh, 133Cs, natNd, 149Sm, 152Sm, and 155Gd. The program aimed at compensating for the lack of information on critical experiments involving FPs and at establishing a basis for FPs credit validation. One hundred forty-five critical experiments were performed, evaluated, and analyzed with the French CRISTAL criticality safety package and the American SCALE5.1 code system employing different cross-section libraries. The aim of the paper is to show the experimental data potential to improve the ability to perform validation of full burnup credit calculation. The paper describes three phases of the experimental program; the results of preliminary evaluation, the calculation, and the sensitivity/uncertainty study of the FP experiments used to validate the APOLLO2-MORET 4 route in the CRISTAL criticality package for burnup credit applications.

R&D on support to ITER safety assessment

After performing its first ITER safety assessment in 2002 on behalf of the French “Autorité de Sûreté Nucléaire (ASN)”, the French “Institut de Radioprotection et de Sûreté Nucléaire (IRSN)” is now analysing the new ITER Fusion facility safety file. The operator delivered this file to the ASN as part of its request for a creation decree, legally necessary before building works can begin on the site. The IRSN first task in following ITER throughout its lifetime is to study the safety approach adopted by the operator and the associated issues. Such a challenging new technology calls for further in-house expertise and so in parallel a R&D program has been set up to support this safety assessment process, now and in the next years. Its main objectives are to identify the key parameters for mastering some risks (that would have been insufficiently justified by the operator) and to perform some verifications with methods and codes independent from the operator's ones. Priority has been given to four technical issues (others could be investigated in the future, like the behaviour of activated corrosion products). The first issue concerns the simulation of accident sequences with the help of the ASTEC European system code, developed by IRSN (jointly with its German counterpart, the GRS) for severe accidents in Pressurised Water Reactors. A preliminary analysis showed that most of its physical models are already applicable, e.g., for thermal-hydraulics in accidents caused by water or air ingress into the vacuum vessel (VV) or dust transport. Work has started in 2008 on some model adaptations, for instance oxidation of VV first wall materials by steam or air, and on validation on the ITER-specific ICE and LOVA experiments. Other model improvements are planned in the next years, as feedback from the work done for the other technical issues and from the code validation. The second issue concerns the risk of gas explosion due to concentrations of hydrogen and carbon monoxide in the ITER main volumes (VV but also the neighbouring volumes), produced by wall materials oxidation. A consistent program of modelling in the Computational Fluid Dynamics (CFD) TONUS IRSN code and of experiments in the TOSQAN (IRSN) and ENACEFF (CNRS/Orléans, France) facilities has been defined: inertising as means of mitigation, flammability limits, flame acceleration and transition to detonation, gas combustion. The third issue concerns the risk of dust explosion due to concentrations in the ITER volumes after remobilisation of dust deposits on walls. A consistent program of modelling in the CAST3M code (CEA), in collaboration with the Carthagene University (Spain), and of experiments in the BISE and TOSQAN IRSN facilities has been defined. The concerned phenomena are dust spatial distribution, remobilisation and entrainment, and explosion. The studies will also address the characterisation of dust and the control and mitigation processes. The fourth issue concerns the tritium behaviour. The theoretical analysis addresses its retention in the VV first walls, its chemistry in the gas phase during transport in the cooling circuits, and in the liquid phases during trapping in the neighbouring volumes or buildings. The efficiency of the diverse processes foreseen for detritiation, either in the VV or in the other rooms, in relation with the ventilation systems, is being also investigated.

Experimental and numerical approaches of aerosol removal in spray conditions for containment application

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear pressurized water reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety code validation. The present work is devoted to study a water spray injection used as a mitigation means in order to washout aerosol fission products.

Measurement of secondary emissions during laser cutting of steel equipments

In order to dismantle some equipments of an obsolete reprocessing plant in Marcoule, studies were carried out by IRSN (Institut de Radioprotection et de Sûreté Nucléaire)/DSU/SERAC in cooperation with CEA (power laser group) on the laser cutting of steel structures, on the request of AREVA NC/Marcoule (UP1 dismantling project manager) and CEA/UMODD (UP1 dismantling owner). These studies were aimed at: • quantifying and characterizing the secondary emissions produced by Nd-YAG laser cutting of Uranus 65 steel pieces and examining the influence of different parameters, • qualifying a prefiltration technique and particularly an electrostatic precipitator, • comparing the Nd-YAG laser used with other cutting tools previously studied especially on aerosol production and aerosol size distribution.

Heat, Mass and Aerosol Transfers in Spray Conditions for Containment Application

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Surete Nucleaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulating typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

Evaluation of safety of French type B package designs in severe accident environments other than regulatory

Abstract Packages used to transport radioactive materials in France must be designed to meet the safety performance requirements when subject to the test conditions set forth in the International Atomic Energy Agency (IAEA) Regulations. During actual use, the packages may be subject to quite different accident conditions. The Institut de Radioprotection et de Surete Nucleaire (IRSN) has evaluated the behaviour of typical packages designed to transport spent fuel, high activity waste, fresh mixed oxide (MOX) fuel and plutonium oxide powder under realistic conditions of mechanical impact and fire. The studied designs remain safe after impact onto targets present in the real environment of transport. The energy absorption by the package ancillary equipment (transport frame) compensates for the kinetic energy increase by comparison to the energy expended during the regulatory tests. New software was developed to correctly simulate the thermal behaviour of the neutron shielding materials. The studied package d...

Two-phase flows in porous media: Prediction of pressure drops using a diffuse interface mathematical description

In the framework of its research programme on severe nuclear reactor accidents, the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) investigates the water flooding of an overheated porous bed, where complex two-phase flows are likely to exist. A better understanding of the flow at the pore level appears to be necessary in order to justify and improve closure laws of macroscopic models. To investigate the local features of a two-phase flow in complex geometries, two alternative ways are available: experiments with high accuracy local measurements or direct numerical simulation (DNS) of the flows in small-scale volumes. Within the framework of the second gradient theory, diffuse interface models can be derived for both single-component and binary fluids. The latter model is often referred to as the Cahn–Hilliard model. In this paper, the Cahn–Hilliard model is used to simulate immiscible two-phase flows in a representative geometry of a particle bed. The results are used to characterize the configuration of the phases and possible flow regimes as a function of the saturation. Correlations for the relative permeabilities of each phase as a function of saturation are also presented. The results are discussed and compared with classical correlations.

Experimental investigation in the TOSQAN facility of heat and mass transfers in a spray for containment application

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear-pressurized water reactor (PWR) containment. The TOSQAN facility which is highly instrumented with non-intrusive optical diagnostics is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we perform detailed characterization of the two-phase flow.