SILENE Reactor Research Articles

Simulation and Analysis of List Mode Measurements on SILENE Reactor

ABSTRACTNeutron time correlation measurements are used to characterize fissile systems. For this purpose, detailed information are obtained using list mode measurements where the neutron detection times are recorded in a file called the time list file. In this article, we present the simulation and analysis of list mode measurements performed on SILENE reactor. Analogue Monte Carlo simulations are performed using MCNPX code and the synthetic time list file is obtained by processing the MCNPX-PTRAC coincidence capture file. Both measured and computed time list files are processed using the same postprocessor that we have developed. This software is used to calculate counting statistics data like counting number probability distribution, Rossi alpha and Feynman functions for time gates of increasing duration. The parameter uncertainties are evaluated using a block bootstrap method.

Representing Nuclear Criticality Excursion Experiment Data by an Artificial Neural Network

A feedforward artificial neural network (ANN) is constructed using select nuclear criticality excursion experiment data sets from the French Consequences Radiologiques d’un Accident de Criticité (CRAC) and SILENE reactor campaigns. The ability to represent initial spike characteristics by an ANN provides a new method that is aligned to excursion data more directly and to a wider variable data set than traditional analytic approaches. The ANN is configured, trained, validated, and tested to 85 unique highly enriched uranium (HEU) excursion experiments, considering six input variables and two output variables (specific power and energy). The fidelity of the ANN is enhanced by normalizing the input and output data. The trained ANN is then used to determine output values for 19 select Kinetic Energy Water Boiler experiments and 14 additional CRAC excursions not used in the ANN construction. Furthermore, the same trained ANN is also used for an extensive comparison (80 cases) for a combination of uranium concentrations, ramp feed reactivity insertion rates, system volumes, and vertical container sizes. The specific spike energy and power ranges determined are bracketed by published experiment results and are more realistically represented than results derived from well-known analytical methods. The ability to predict initial peak fissions by an ANN does not require determining, a priori, a volume-dependent energy quench parameter (“b”) specific to HEU solutions. The results derived from the ANN can aid in designing realistic emergency planning constructs or criticality accident alarm system hardware placements without undue penalty for fission source term uncertainties. Neither excursion characteristics after the initial spike nor explicit time dependencies are modeled by an ANN at this time. The extension of the methods presented is left for further work.

Recalibration of indium foil for personnel screening in criticality accidents

At the Nuclear Fuel Cycle Engineering Laboratories of the Japan Atomic Energy Agency (JAEA), small pieces of indium foil incorporated into personal dosemeters have been used for personnel screening in criticality accidents. Irradiation tests of the badges were performed using the SILENE reactor to verify the calibration of the indium activation that had been made in the 1980s and to recalibrate them for simulated criticalities that would be the most likely to occur in the solution process line. In addition, Monte Carlo calculations of the indium activation using the badge model were also made to complement the spectral dependence. The results lead to a screening level of 15 kcpm being determined that corresponds to a total dose of 0.25 Gy, which is also applicable in posterior-anterior exposure. The recalibration based on the latest study will provide a sounder basis for the screening procedure in the event of a criticality accident.

Hair Dosimetry Following Neutron Irradiation

Use of hair as a biological dosimeter of neutron exposure was proposed a few years ago. To date, the (32)S(n,p)(32)P reaction in hair with a threshold of 2.5 MeV is the best choice to determine the fast neutron dose using body activation. This information is essential with regards to the heterogeneity of the neutron transfer to the organism. This is a very important parameter for individual dose reconstruction from the surface to the deeper tissues. This evaluation is essential to the adapted management of irradiated victims by specialized medical staff. Comparison exercises between clinical biochemistry laboratories from French sites (the CEA and COGEMA) and from the IRSN were carried out to validate the measurement of (32)P activity in hair and to improve the techniques used to perform this examination. Hair was placed on a phantom and was irradiated at different doses in the SILENE reactor (Valduc, France). Different parameters were tested: variation of hair type, minimum weight of hair sample, hair wash before measurement, delivery period of results, and different irradiation configurations. The results obtained in these comparison exercises by the different laboratories showed an excellent correlation. This allowed the assessment of a dose-activity relationship and confirmed the feasibility and the interest of (32)P measurement in hair following fast neutron irradiation.

TL detectors for gamma ray dose measurements in criticality accidents

Determination of gamma ray dose in mixed neutron+gamma ray fields is still a demanding task. Dosemeters used for gamma ray dosimetry are usually in some extent sensitive to neutrons and their response variations depend on neutron energy i.e., on neutron spectra. Besides, it is necessary to take into account the energy dependence of dosemeter responses to gamma rays. In this work, several types of thermoluminescent detectors (TLD) placed in different holders used for gamma ray dose determination in the mixed fields were examined. Dosemeters were from three different institutions: Ruder Bosković Institute (RBI), Croatia, JoZef Stefan Institute (JSI), Slovenia and Autoridad Regulatoria Nuclear (ARN), Argentina. All dosemeters were irradiated during the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002. Three accidental scenarios were reproduced and in each irradiation the dosemeters were exposed placed on the front of phantom and 'free in air'. Following types of TLDs were used: 7LiF (TLD-700), CaF2:Mn and Al2O3:Mg,Y-all from RBI; CaF2:Mn from JSI and 7LiF (TLD-700) from ARN. Reported doses were compared with the reference values as well as with the values obtained from the results of all participants. The results show satisfactory agreement with other dosimetry systems used in the Intercomparison. The influence of different types of holders and applied corrections of dosemeters' readings are discussed.

エボナイトを用いた臨界事故時における中性子被ばく線量の評価

The authors studied the feasibility of utilizing ebonite as a personal neutron dosemeter in criticality accidents. A disc-shaped ebonite, a hard rubber containing 30wt% sulfur, can be used as a highly effective criticality neutron dosemeter because of a simplicity of measurements of beta activity arising from 32S(n, p)32P reactions. The counting efficiency of beta particles with an end-window GM counter for an ebonite disc in 50mm diameter and 3mm thick was determined by 252Cf neutron irradiation. The neutron spectrum dependency of 32P activity per neutron dose was computed using Monte-Carlo calculations of various neutron spectra that could be encountered in criticality accidents, and the results were tabulated as a set of spectrum correction factors. Performance tests using the SILENE reactor indicated that neutron doses could be evaluated within ±15% with the application of suitable correction factors.

Study and presentation of a fast neutron and photon dosemeter for area and criticality monitoring using radiophotoluminescent glass

This paper describes the results of a study performed on a mixed field neutron/gamma (n/gamma) area dosemeter incorporating radiophotoluminescent (RPL) glass detectors. RPL glass is known to be virtually insensitive to neutrons. The aim of the study was therefore to determine the neutron response of a dosemeter designed to combine n/gamma conversion with RPL detection capability. Monte Carlo calculations as well as measurements using monoenergetic beams and isotopic neutron sources showed this response to be constant, to within 30% in terms of H*(10), and independent of neutron energy from 250 keV to 10 MeV. For area monitoring, tests carried out in nuclear facilities (around PuO2 glove box and shipping casks containing PWR, MOX spent fuels or vitrified fission product) demonstrated that dosemeter response was accurate to within 15%, where the gamma component of the mixed n,gamma field remained below 1 MeV. When exposed in the Silene reactor simulating a criticality accident (10(17) fissions-liquid 235U--e.g. 1 Gy neutron and 1 Gy photon), the dosemeter exhibited good correlation with reference values and other measurement technologies (again to within 30%), for both neutron and gamma absorbed dose.

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.

Criticality Accident Dosimetry: An International Intercomparison at the Silene Reactor

criticality accident dosimetry, specialised techniques, which differ markedly from those used in routine radiological protection, are used to measure separately the gamma ray and neutron components of dose. In order to improve these techniques and to evaluate the performance of their dosimetry systems, fifteen laboratories participated in an international intercomparison at the SILENE reactor, Valduc, France. The physical techniques used in criticality dosimetry are first described, for personnel and installed dosemeters, and for the initial sorting of those exposed. The SILENE reactor is briefly described together with a review of the dosimetry and spectrometry techniques used to characterise the radiation field at a reference position and to study the uniformity of the field on an arc around the source. The experimental arrangements of the participants' dosemeters for the intercomparison irradiations (in free-air and on phantoms) are given, followed by a summary of the doses measured by the participants. From an overall analysis of this intercomparison, it is concluded that the participants' results are in good agreement with the reference measurements for both the neutron and gamma ray components of dose.

Chromosome Aberrations Scoring for Biological Dosimetry in a Criticality Accident

During an international intercomparison of dosimetry systems for the simulation of a criticality accident which took place at the SILENE Reactor (Valduc, France), whole blood was exposed to reference pulses of mixed gamma and neutron irradiation. This was done with specimens supported free-in-air and attached to a phantom. Each blood sample was shared between the NRPB and IPSN laboratories, cultured and scored for chromosomal aberrations. Each laboratory made estimates of the neutron and gamma components of dose by reference to their own pre-existing in vitro calibration curves. It was necessary to have an estimate of the ratio of the neutron to gamma dose for this purpose. Good agreement was obtained between the results from both laboratories and with the reference doses. The conclusion is that biological dosimetry using peripheral lymphocytes gives credible estimates of personal doses in criticality accidents.

Silene, A Tool for Neutron Dosimetry

The SILENE reactor today is a reference source able to meet the expectation of a great number of researchers in the fields of physics, neutronics, biology and dosimetry. The latest technical developments and its multiple operating possibilities enable this experimental reactor of the French Nuclear Protection and Safety Institute (IPSN) to provide a large range of calibrated neutron and gamma radiation fields which can be reproduced very easily.

Criticality accident dosimetry with ESR spectroscopy

The suitability of the ESR alanine and sugar detectors for criticality accident dosimetry was experimentally investigated during an intercomparison of dosimetry techniques. Tests were performed irradiating detectors both free-in-air and on-phantom during controlled criticality excursions at the SILENE reactor in Valduc, France. Several grays of absorbed dose were imparted in neutron gamma-ray fields of various relative intensities and spectral distributions. Analysed results confirmed the potential of these systems which can immediately provide an acute dose assessment with an average underestimate of 30% in the various fields. This performance allows for the screening of severely exposed individuals and meets the IAEA recommendations on the early estimate of accident absorbed doses.

<sup>114</sup>In<sup>m</sup> as a Long Half-Life Exposure Indicator for Criticality Accident Dosimetry

Indium is activated, principally by epithermal neutrons, to give 114Inm with a half-life of about 50 d. Indium foils exposed in the UK criticality dosemeter, at an International Intercomparison Experiment at the SILENE reactor, were measured three months later both by gamma ray spectrometry and beta ray counting. The specific activity measured was in good agreement with that estimated independently from gold foil measurements of the thermal and intermediate energy neutron fluences. Beta ray counting with a low background end window Geiger-Muller counter was the more sensitive method of measurement, giving a detection efficiency of about 0.14 count.disintegration-1. This gave a practical lower limit of measurement for 114Inm of 0.5 Bq.g-1, which corresponded to total neutron doses of about 16mGy and 21mGy for the lead shielded and bare SILENE neutron spectra respectively. Such measurements demonstrate the value of indium as a long-term exposure indicator, in addition to its usually accepted uses as a short-term exposure indicator (116Inm) and a fast neutron detector (115Inm).

The EDAC System and New Developments under Consideration at the Commissariat à I’Energie Atomique for Criticality Accident Detection

As a result of the Commissariat à I’Energie Atomique CRAC experimental program, which studied the phenomenology and the radiological consequences of a criticality excursion in fissile solution, the EDAC system has been developed. This system detects a criticality accident and warns personnel as early as possible by triggering the necessary audiovisual alarm. The main features of this equipment are its ability to cover all types of accidental kinetics and to use sensors giving a total dose response in neutron and gamma radiation. According to new results acquired with the SILENE reactor in the field of criticality accidents, an evolution is taking place in France. An improved EDAC system is being designed not only to trigger a criticality alarm but also to provide information on the accident, to assist in accident diagnosis, and to contribute to being better equipped to cope with an accident situation, for example, if intervention is needed or if reoccupation of evacuated areas is desired.