Nuclear Safety Institute Research Articles

Monte Carlo calculations and validation of a gold foil-based Bonner sphere system

The Grup de Física de les Radiacions (GFR) of the Universitat Autònoma de Barcelona (UAB), in collaboration with the Institute for Radiological Protection and Nuclear Safety (IRSN), has developed a passive Bonner sphere system (UAB-BSS), with gold foils as thermal neutron detectors, for application in pulsed neutron fields or in mixed neutron-photon fields with high photon intensities. In such fields, active devices suffer from saturation and dead-time effects. The MCNPX Monte-Carlo code has been used to determine the response to neutrons of different energies of each polyethylene sphere belonging to the BSS. The passive UAB-BSS system was characterised with the ISO (252)Cf reference source at the IRSN facilities. The energy distribution of the reference source neutron fluence was folded with the response functions for comparison with the experimental data. A good agreement between the experimental and calculated count rates was found.

Design of a new IRSN thermal neutron field facility using Monte-Carlo simulations

The Institute for Radiological Protection and Nuclear Safety owns a graphite-moderated AmBe neutron field facility, SIGMA, that has to be reconstructed. Monte-Carlo simulations were performed to study the design of a new thermal facility based on IRSN existing facilities. Studies related to an update version of SIGMA concerned the enhancement of the thermal neutrons contribution to the dose equivalent. Calculations were mainly performed for a (252)Cf neutron source distribution located at the centre of a graphite moderator block. A quasi-pure thermal neutron field was obtained with a 2.4 x 2.4 x 2.4-m(3) block of graphite. A second acceptable neutron field was obtained with 3.3-MeV mono-energetic neutrons created by a 400-kV accelerator coupled to a graphite assembly of 1.5 x 1.5 x 1.5 m(3). The characteristics of the studied thermal fields with the requirement for a reference calibration field are compared, and the advantages and drawbacks of the different producing methods are discussed.

Feasibility study of extremity dosemeter based on polyallyldiglycolcarbonate (CR-39) for neutron exposure

In nuclear facilities, some activities such as reprocessing, recycling and production of bare fuel rods expose the workers to mixed neutron-photon fields. For several workplaces, particularly in glove boxes, some workers expose their hands to mixed fields. The mastery of the photon extremity dosimetry is relatively good, whereas the neutron dosimetry still raises difficulties. In this context, the Institute for Radiological Protection and Nuclear Safety (IRSN) has proposed a study on a passive neutron extremity dosemeter based on chemically etched CR-39 (PADC: polyallyldiglycolcarbonate), named PN-3, already used in routine practice for whole body dosimetry. This dosemeter is a chip of plastic sensitive to recoil protons. The chemical etching process amplifies the size of the impact. The reading system for tracks counting is composed of a microscope, a video camera and an image analyser. This system is combined with the dose evaluation algorithm. The performance of the dosemeter PN-3 has been largely studied and proved by several laboratories in terms of passive individual neutron dosemeter which is used in routine production by different companies. This study focuses on the sensitivity of the extremity dosemeter, as well as its performance in the function of the level of the neutron energy. The dosemeter was exposed to monoenergetic neutron fields in laboratory conditions and to mixed fields in glove boxes at workplaces.

Fire PSA of the French PWR 900 MWe series

Abstract The Institute for Radioprotection and Nuclear Safety (IRSN) has developed a Fire Probabilistic Safety Assessment of the operating French 900 MWe PWR in order to organize into a hierarchy the items of equipments and the compartments, impacting the core melt frequency following a fire, and to reinforce deterministic safety analysis with functional analysis and realistic fire scenarios. The IRSN Fire PSA is composed of two stages: the first one concerns the selection of critical zones and the second one concerns the detailed study of the critical compartments, corresponding to the development of the fire scenarios and of the fire-induced core damage sequences. The first version of the IRSN Fire PSA has been updated by integrating the specificities of the Fire Action Plan, implemented by the utility on the French 900 MWe PWR. This study shows that four compartments are preponderant. The integration of the Fire Action Plan allowed decreasing the number of preponderant compartments. The integration of the symptom-oriented procedures and the removal of conservatisms should lead to a lower core damage frequency for these preponderant compartments. The results show that fire is not a negligible risk but the implementation of the Fire Action Plan has a significant positive impact on safety.

ICONE15-10471 A TONUS Code Application : Hydrogen Mitigation Analysis on CNP300

The TONUS code incorporates both Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) formulations, is jointly developed by the French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) for hydrogen risk safety analysis. During the cooperation between CEA and Chinese Atomic Energy Authority (CAEA) for severe accident management associated laboratory, the incorporation of a Chinese recombiner model into the TONUS code was carried out and an analysis was performed with this new recombiner model for the Chinese 300MWe Nuclear Power Plant (CNP300). The analysis of hydrogen issue for the CNP300 was performed with two recombiner models, the standard TONUS and Chinese one using several hypothetical accident scenarios. The obtained results show that the Chinese recombiner model results in a lower hydrogen concentration in CNP300 than the standard TONUS one due to its geometrical structure and high efficiency.

Characterization of neutron field in a NPP workplace

At the Krsko Nuclear Power Plant (NPP), albedo dosimeters are used for personal neutron dosimetry. Spectrometric measurements allow determination of reference dosimetric values of realistic neutron fields to be used for calibration of albedo dosimeters. The Laboratory for Neutron Metrology and Dosimetry from the Institute for Radiological Protection and Nuclear Safety (IRSN) was in charge of characterising neutron fields in the plant at two representative points with high neutron and gamma dose rate. Calibration of the dosimeters in the workplace used to be performed only by a spherical survey meter. Based on the reference dosimetric values, the Plant Dosimetry Laboratory has verified the response of albedo dosimeters.

Monte Carlo simulation of the operational quantities at the realistic mixed neutron-photon radiation fields CANEL and SIGMA

The Institute for Radiological Protection and Nuclear Safety owns two facilities producing realistic mixed neutron-photon radiation fields, CANEL, an accelerator driven moderator modular device, and SIGMA, a graphite moderated americium-beryllium assembly. These fields are representative of some of those encountered at nuclear workplaces, and the corresponding facilities are designed and used for calibration of various instruments, such as survey meters, personal dosimeters or spectrometric devices. In the framework of the European project EVIDOS, irradiations of personal dosimeters were performed at CANEL and SIGMA. Monte Carlo calculations were performed to estimate the reference values of the personal dose equivalent at both facilities. The Hp(10) values were calculated for three different angular positions, 0 degrees, 45 degrees and 75 degrees, of an ICRU phantom located at the position of irradiation.

Bonner sphere neutron spectrometry at nuclear workplaces in the framework of the EVIDOS project

The Institute for Radiological protection and Nuclear Safety was engaged in the EC funded EVIDOS project to provide reference spectrometry data using its Bonner sphere system. The data were processed by means of two unfolding codes, NUBAY and GRAVEL, both provided by the Physikalisch-Technische Bundesanstalt. The NUBAY program, based on Bayesian parameter estimation methods, assumes a parameterised spectrum and provides posterior probability distributions for the parameters. The code GRAVEL, an iterative algorithm based on SAND-II, was used with various default spectra, among them the NUBAY solution. The BS measurements were used to establish the neutron fluence energy distributions and reference values for the neutron ambient dose equivalent. As this quantity depends strongly on the high energy neutrons, a sensitivity analysis was done by unfolding the BS data with GRAVEL using the NUBAY solution spectrum as default with various changes in the parameters of the high energy peak. This new method of analysing Bonner sphere data allowed the determination of reliable neutron spectra, as well as a very good estimate of the corresponding integral quantities with small associated uncertainties.

Evaluation of dose equivalent by the electronic personal dosemeter for neutron 'Saphydose-N' at different workplaces of nuclear facilities

This paper presents the results of measurements made with the electronic personal neutron Saphydose-N during the four campaigns of the European contract EVIDOS (EValuation of Individual DOSimetry in mixed neutron and photon radiation fields). These measurements were performed at Institute for Radiological Protection and Nuclear Safety (IRSN) in France (C0), at the Krümmel Nuclear Power Plant in Germany (C1), at the VENUS Research Reactor and the Belgonucléaire fuel processing plant in Belgium (C2) and at the Ringhals Nuclear Power Plant in Sweden (C3). The results for Saphydose-N are compared with reference values for dose equivalent.

Childhood Leukaemia Incidence Around French Nuclear Installations Using Geographic Zoning Based on Gaseous Discharge Dose Estimates

P-431 Introduction: The epidemiological studies of leukaemia incidence around nuclear sites analyzed incidence with respect to the distance from the plants, without considering any information on the levels or geographic distribution of the radiation dose due to discharges from the plants. However, the dispersion of radionuclides is generally not isotropic and does not decrease linearly with the distance, while the dose levels are not equal for all sites. Methods: The present study investigated for the first time the incidence of childhood leukaemia around French nuclear installations using a geographic zoning based on estimated doses to the red bone marrow due to gaseous radioactive discharges. The cases were all children under 15 years of age diagnosed with leukaemia between 1990 and 2001 and living around nuclear sites at the time of diagnosis. They were provided by the French National Registry of Childhood Leukemia and Lymphoma, which registered 5,330 cases of acute leukaemia from 1990 to 2001 for the mainland France. The doses were estimated by the Institute for Radiation Protection and Nuclear Safety using radionuclide discharge data, local climate data and mathematical models of nuclide transfers in the environment. A spatial zoning based on dose-level order of magnitude in 40 km squares centered on 23 nuclear installations was derived. The studied area was cut up in five dose-based categories, with equilibrated numbers of expected cases. Results: The observed number of cases of acute leukaemia (O=750) within the study area was lower than expected (E=795.01), but the difference was not statistically significant (SIR=0.94, 95% confidence interval CI=[0.88–1.01]). In none of the five zones defined on the basis of the estimated doses was the SIR significantly greater than 1. No evidence of heterogeneity between the five dose-based categories, and no trend toward an increase in SIR with increasing estimated dose were found for all the children or for any of the three age-groups studied, for any of the periods (1990–1995, 1996–2001), or for any of the leukaemia types (acute lymphoblastic leukaemia, acute myeloid leukaemia). Discussion and Conclusions: This approach has the advantage of considering the non-isotropic distribution of gaseous discharges in the environment. It also illustrates that the radiation dose due to routine discharges from the studied nuclear installations is much smaller than that due to natural sources of exposure to ionizing radiation. This study confirmed that there was no evidence of an increased incidence of childhood leukaemia around the 23 French nuclear sites.

Lung cancer attributable to indoor radon exposure in france: impact of the risk models and uncertainty analysis.

ObjectiveThe inhalation of radon, a well-established human carcinogen, is the principal—and omnipresent—source of radioactivity exposure for the general population of most countries. Scientists have thus sought to assess the lung cancer risk associated with indoor radon. Our aim here is to assess this risk in France, using all available epidemiologic results and performing an uncertainty analysis.MethodsWe examined the exposure–response relations derived from cohorts of miners and from joint analyses of residential case-control studies and considered the interaction between radon and tobacco. The exposure data come from measurement campaigns conducted since the beginning of the 1980s by the Institute for Radiation Protection and Nuclear Safety and the Directorate-General of Health in France. We quantified the uncertainties associated with risk coefficients and exposures and calculated their impact on risk estimates.ResultsThe estimated number of lung cancer deaths attributable to indoor radon exposure ranges from 543 [90% uncertainty interval (UI), 75–1,097] to 3,108 (90% UI, 2,996–3,221), depending on the model considered. This calculation suggests that from 2.2% (90% UI, 0.3–4.4) to 12.4% (90% UI, 11.9–12.8) of these deaths in France may be attributable to indoor radon.DiscussionIn this original work we used different exposure–response relations from several epidemiologic studies and found that regardless of the relation chosen, the number of lung cancer deaths attributable to indoor radon appears relatively stable. Smokers can reduce their risk not only by reducing their indoor radon concentration but also by giving up smoking.

Development of a global rainbow refractometry technique to measure the temperature of spray droplets in a large containment vessel

In order to study the heat and mass transfers between a spray of droplets and the atmosphere in thermal-hydraulics conditions representative of a severe accident in a Pressurized Water Nuclear Reactor, the French Institute for Radioprotection and Nuclear Safety (IRSN) developed the TOSQAN facility. The present paper presents the development and the quantification of an optical diagnostic, global rainbow refractometry, in order to measure falling droplet temperature.

Mineralogical, chemical and isotopic perturbations induced in shale by fluid circulation in a fault at the Tournemire experimental site (Aveyron, France)

Nuclear waste confinement systems in shales depend, among other factors, on their low permeability, alkaline pH and low Eh. These properties may be altered close to fractures, in the presence of fluids in chemical disequilibrium with the host formation. Since shales are well known for their self-sealing response to fracturing, the question of possible perturbations close to faults is not often treated. In order to understand how the shale may be affected in such cases, a mineralogical, chemical and isotopic study of a fractured zone was conducted at the French Institute for Nuclear Safety and Radioprotection (IRSN) experimental site. This site is located in a 250 m thick, fractured, Domerian and Toarcian shale layer, located between two aquifers, near Tournemire, on the western margin of the Causses basin (France). Oxygen isotopic analyses of calcite veins from the last tectonic phase suggest that cements crystallised from both pore water and meteoric fluids. Variable Fe content in calcite and presence of pyrite indicate that fault cements crystallized in different reduced environments. In the shale, δ 13C values are similar in the cement and in calcite veins. In the studied fault, however, δ 13C are closer to values measured in calcite from veins in the overlying limestone. This suggests that the fault allowed fluids from the upper aquifer to penetrate the shale. In the host rock, variations in Ca and Sr concentrations and δ 18O suggest that carbonate is altered and partly recrystallised up to 1 m from the main fault zone (MFZ). Ba and Zn are slightly concentrated near the MFZ as authigenic baryte and pyrite. Evidence of partial REE mobilisation was found. Variations of U and Th concentrations are quite low and suggest that if they are mobilised, they do not reach the MFZ.

Study of the tools available in biological dosimetry to estimate the dose in cases of accidental complex overexposure to ionizing radiation: The Lilo accident

Purpose: To compare the efficiency of different cytogenetic tools in estimating the doses received by four people involved in the Lilo accident and to monitor the dose estimate over 4.5 years.Materials and methods: Several young Georgian frontier guards handled at least one of the 12 Caesium sources found in a former Russian military camp. Overexposure lasted from July 1996 to May 1997. The Institute for Radiological Protection and Nuclear Safety (IRSN) obtained blood samples taken at several intervals post-exposure from the four most highly-exposed people. Dose estimation was performed using dicentric and translocation scoring.Results: The first dose estimations performed by dicentric scoring gave whole-body doses ranging from 0.4 to 1.3 Gy. Overexposure was complex and several mathematical models were used to take this complexity into account. This could provide information concerning the circumstances of overexposure. Concerning follow-up, the yield of dicentrics decreased by about 50% in the first 4 months following the end of overexposure whereas translocations were stable over the period of analysis.Conclusion: It has been useful to compare cytogenetic results with clinical results. The results presented here reveal good stability of translocations. However the first dose estimation was not attempted until 6 months after the last exposure.

Indoor Radon and Lung Cancer

The authors respond: The design of the French case-control study on lung cancer risk associated with indoor radon exposure1 was conducted within the framework of a common European program focusing on lung cancer risk linked to radon in dwellings. The same protocol and questionnaire were applied in other countries (and the same approach to data collection was used in Germany and Great Britain). Intercomparison of measurement techniques has been performed, and major collaboration also was realized in the field of statistical analysis (same logistic regression approach in Germany and France, with the same definition of time-weighted variables and discussion of log-linear and linear excess relative risk models). Our study was consistent with international recommendations and satisfied specific criteria to be part of a pooled analysis regrouping major European studies; this joint analysis was recently published by Darby et al.2 Considering the specific criticisms of Lachet,3 we agree with the fact that, in Table 3, the last column does not correspond to the category of exposure >400 Bq/m3 but presents the results of the model considering radon exposure as a continuous variable. A clear separation between the last 2 columns would have made this more evident. Other criticisms derive from a misunderstanding of the nature of the data and of the statistical approach. As stated in the Methods section of our article,1 the analysis was based on logistic regression. This approach is classically used in epidemiology for the analysis of case-control design studies,4 and was used in previous analyses of indoor radon case-control studies.5,6 This regression method relies on a log-linear model and, therefore, it is not surprising that the estimated exposure-risk relationship presented in Figure 1 is not a linear one. As also stated in the Methods section, 2 approaches were performed to analyze the relationship between radon and lung cancer risk: the first one considers radon as a categorical variable with the <50 Bq/m3 category being the reference (RR is fixed to 1.0 in this category) and the second one considers radon as a continuous variable on the whole range of exposure. The second approach does not depend on a reference category. It is therefore not surprising that the fitted model represented in Figure 1 starts from the origin. Finally, the reanalysis presented by Lachet in the last paragraph of his letter does not take into account the distribution of cases and controls in the different exposure categories. His approach reflects a misinterpretation of the nature of the data and is not scientifically justified. When Lachet is criticizing the results of our study, he seems to ignore the fact that other groups of researchers with independent data have come to the same conclusion as we have. The European study, with very large numbers of cases and controls, has been recently published in the British Medical Journal.2 There is no heterogeneity between the studies having contributed to this joint analysis. This analysis showed strong evidence of an association between residential radon and lung cancer; the risk of lung cancer increased by 8.4% per 100 Bq/m3 increase of radon exposure. The classic logistic regression approach we used1 gave results similar to those obtained with the French data by Darby et al.2 when they applied a linear excess relative risk model. Collectively, all these data allowed direct assessment of the risk of lung cancer from indoor radon exposure. The appropriate debate should move now from the question of the existence of a risk to the reduction or elimination of this risk. WHO is presently evaluating the risk linked to indoor radon, and our hope is that this international organization will help to implement the appropriate risk management and communication to the general public for this well-documented public health problem. Hélène Baysson Margot Tirmarche Klervi Leuraud Dominique Laurier Institute for Radiological Protection and Nuclear Safety (IRSN) Fontenay aux Roses, France

Current developments at IRSN on computational tools dedicated to assessing doses for both internal and external exposure

The paper presents the OEDIPE (French acronym that stands for tool for personalised internal dose assessment) and SESAME (for simulation of external source accident with medical images) computational tools, dedicated to internal and external dose assessment, respectively, and currently being developed at the Institute for Radiological Protection and Nuclear Safety. The originality of OEDIPE and SESAME, by using voxel phantoms in association with Monte Carlo codes, lies in their ability to construct personalised voxel phantoms from medical images and automatically generate the Monte Carlo input file and visualise the expected results. OEDIPE simulates in vivo measurements to improve their calibration, and calculates the dose distribution taking both internal contamination and internal radiotherapy cases into account. SESAME enables radiological overexposure doses to be reconstructed, as also victim, source and accident environment modelling. The paper presents the principles on which these tools function and an overview of specificities and results linked to their fields of application.

Radioprotection

The growing number of human activities based on ionising radiation, along with the demonstration of its effects on the human body, has led to the creation of a radiation protection system. This system is based on the recommendations of the International Commission on Radiological Protection (ICRP). Its purpose is to provide a framework for the use of ionising radiation, exclude deterministic effects and restrict stochastic ones. The French regulations follow these recommendations quite closely, via the EURATOM European Directives. The French system constitutes a coherent whole, relevant to the general population, patients and workers, in both normal situations and those involving exceptional exposure or radiological emergencies. It lays down a strict framework for such practices, which makes the implementation of an effective risk control strategy easier. Within this radiation protection organisation, IRSN (the French institute for radiological protection and nuclear safety) supplies expertise, while DGNSR (the French general department for nuclear safety and radiation protection) is in charge of technical and regulatory radiation protection monitoring.

Calibration factor for estimating personal dose equivalent with imaging plates

A personal imaging plate (IP) dosimeter is in the process of being developed for neutron fields using the BaFBr: Eu 2 + phosphor. A configuration incorporating a polyethylene radiator placed before the IP detector is used to produce protons via (n,p) elastic scattering. For a dosimeter sensitive to thermal neutrons, a Nylon plate ( 100 μ m thick) is placed between the polyethylene (1.2 mm thick) radiator and the IP ( 130 μ m thick sensitive layer) detector to produce protons via the 14 N ( n , p ) 14 C reaction. Dosimeters having these configurations have been exposed to neutrons from 241 Am –Be and 252 Cf sources at the Institute for Radioprotection and Nuclear Safety of Cadarache at angles of 0 ∘ (normal incidence), 30 ∘ and 60 ∘ and several dose equivalents. The personal dose equivalent response in terms of H p ( 10 ) is evaluated from the net measured photostimulated luminescence densities ( PSL mm - 2 ). The calibration factor obtained for estimating the personal dose equivalent with this dosimeter is 9.20 × 10 - 5 mSv PSL - 1 mm 2 for 241 Am –Be and 13.07 × 10 - 5 mSv PSL - 1 mm 2 for 252 Cf .

Management of the post-accidental situation applied to Nogent-sur-Seine nuclear power plant environment. First results of the “decontamination of soil and food chain" working group

From the beginning of 2002, Troyes prefecture has initiated a study about the management of a nuclear crisis caused by an accident at the Nogent-sur-Seine nuclear power plant. Several working groups were created, notably the `Decontamination of soil and food chain' group. The first results of this group, which involves the District Agricultural and Forestry Department, the Farming Chamber, the Institute for Radiological Protection and Nuclear Safety and the Veterinary Division are presented here. The scenario that had been developed for the accident considers the release of 3 radionuclides (131 I, 134 Cs and 137 Cs) in the environment. The scale of the crisis didn't require the evacuation and the sheltering of the population during the emergency phase. The consequences on the local agricultural products (cereal, sugar beet, vine, milk, cow meat ...) have been assessed up to 50 km and different strategies of agricultural countermeasures have been studied regarding to the local constraints (soil types, surfaces and quantities concerned) and to the consequences of their implementation (waste types and quantities, costs). Then, decision-making diagrams summed up the technical results and allowed to deepen the global thought.

Dose evaluation in criticality accident conditions using transient critical facilities fueled with a fissile solution

Neutron dose measurement and evaluation techniques in criticality accident conditions using a thermo luminescence dosemeter (TLD) was studied at the Transient Experiment Critical Facility (TRACY) of Japan Atomic Energy Research Institute (JAERI). In the present approach, the absorbed dose is derived from the ambient dose equivalent measured with a TLD, using the appropriate conversion factor given by computation. Using this technique, the neutron dose around the SILENE reactor of the Institute for Radioprotection and Nuclear Safety (IRSN) of France was measured in the Accident Dosimetry Intercomparison Exercise (June 10-21, 2002) organized by OECD/NEA and IRSN. In this exercise, the gamma dose was also measured with a TLD. In this report, measurements and evaluation results at TRACY and SILENE are presented.