Philippe Fougeras

Application of the Modified Source Multiplication (MSM) technique to subcritical reactivity worth measurements in thermal and fast reactor systems

The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEAs EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. The ASM methodology uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. While this method works quite well for small reactivity variations, the raw results need to be corrected to take into account the flux perturbation at the fission chamber location. This is performed by applying to the measurement a correction factor called MSM. This paper describes in detail both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw ASM data.

Monte Carlo Modelling of Increasing Void Fraction in 100% MOX ABWR: Lessons Drawn from the FUBILA Program

The French Atomic Energy Commission CEA and the Japanese Incorporated Administrative Agency JNES (Japan Nuclear Energy Safety Organisation) have undertaken first-of-a-kind full MOX core physics experiments, FUBILA, in the EOLE critical facility of the CEA Cadarache Centre. The experiments have been designed to obtain core physics data under high-burn-up 9 × 9 and 10 × 10 BWR MOX assemblies operating conditions. The experimental program, consisting of eight different core configurations, started in January 2005 and ended on September 1, 2006. The analysis of the void increase part of the experimental data between 0 and 70% void has been carried out using the French TRIPOLI-4.5 continuous-energy Monte Carlo calculation code with the newly released JEFF3.1.1 nuclear data library. The average C/E discrepancies obtained on critical masses, reactivity worth, and pin-by-pin power distributions enable us to estimate all the integral and local parameters with uncertainties largely within the target uncertainties, demonstrating the capability of the code to treat complex geometries with a high degree of accuracy. Additional keff calculations performed with the latest ENDF/B-VII evaluation exhibit a clear tendency to overestimate the keff by about 500 to 650 pcm and the void worth by more than 4%, showing that the JEFF3.1.1 library is more precise for MOX lattices.

A neutron noise-based experiment in a TRIGA reactor

A neutron noise-based experiment performed in the TRIGA reactor of the ENEA-Casaccia centre in Italy is presented. This experiment is part of the first phase of the international TRADE programme that seeks a global demonstration of an accelerator-driven system at power. The main objective of that first experimental phase is the measurement at zero-power of fundamental neutronic parameters. The present experiment performed in the reference critical core configuration has first of all aimed at the inference of the prompt neutron decay constant at critical, which is equal to the ratio of the delayed neutron fraction to the neutron generation time. The neutron-noise technique that has been applied is the power spectral density technique, based on the use of fission chambers operated in fluctuation mode. Two fission chambers, each with a sensitivity of 10-1 cps/nv, have been placed within the core region close to the reflector. The prompt neutron decay constant at delayed critical has been estimated at 132 ± 2 s-1 with 68% confidence level.

Fission Products Particular Peak Measurement for UO2-Gd2O3-MOX γ-Scanning Renormalization in 100% MOX ABWR Mock-Up Cores

This paper presents the principles of the peak check measurements by gamma spectrometry. One details the main equations used for the analysis of the raw data as the calculation of the different sources of uncertainties and their propagation on the result. The method is illustrated with actual examples from the French-Japanese BASALA ABWR 100% MOX experimental program. • For each individual fission rate, the systematical uncertainty due to the radioactive decay data is smaller than the statistical uncertainty due to the counting process. • The main part of the final uncertainty on the scaling factor is brought by the systematical uncertainty on the average fission yields. This study enables to propose some recommendations for fission products and nuclear data evaluation used. • Only the analysis of the 140 La peak at 1596 keV leads to acceptable uncertainties on the fission rate maps renormalization, with a good consistency with the integral γ-scanning results. • The JEF-2.2 evaluation on the fission yields and their associated uncertainties seems more realistic than the ENDF/B-6 and is recommended for the scaling factor analysis.Copyright

The OSMOSE Experimental Program in MINERVE for the Qualification of the Integral Cross Sections of Actinides

Conjointly to the long term or sub-surface storage studies, an important effort is made on plutonium management and waste incineration in existing or innovative systems. In support to these studies, specific neutron integral experiments have been defined and in particular in the critical MINERVE facility of the CEA Cadarache. The OSMOSE program will be performed between 2003 and 2009. It deals with qualifying the integral cross sections of actinides appearing in reactor and fuel cycle physics, especially for multi-recycling of plutonium. Reactivity worth of several samples containing separated actinides from 232Th to 245Cm will be measured by an oscillation technique related to the perturbation theory with an accuracy better than 3%. Experiments will cover a wide range of neutron spectra, from thermal spectrum to fast spectrum.

Preliminary Analysis of the BASALA-H Experimental Programme

This paper is focused on the preliminary analysis of results obtained on the first cores of the first phase of the BASALA (Boiling water reactor Advanced core physics Study Aimed at mox fuel LAttice) programme, aimed at studying the neutronic parameters in ABWR core in hot conditions, currently under investigation in the French EOLE critical facility, within the framework of a cooperation between NUPEC, CEA and COGEMA. The first “on-line” analysis of the results has been made, using a new preliminary design and safety scheme based on the French APOLLO-2 code in its 2.4 qualified version and associated CEA-93 V4 (JEF-2.2) Library, that will enable the Experimental Physics Division (SPEx) to perform future core designs. It describes the scheme adopted and the results obtained in various cases, going to the critical size determination to the reactivity worth of the perturbed configurations (voided, overmoderated, and poisoned with Gd2 O3 -UO2 pins). A preliminary study on the experimental results on the MISTRAL-4 is resumed, and the comparison of APOLLO-2 versus MCNP-4C calculations on these cores is made. The results obtained show very good agreements between the two codes, and versus the experiment. This work opens the way to the future full analysis of the experimental results of the qualifying teams with completely validated schemes, based on the new 2.5 version of the APOLLO-2 code.Copyright