Analysis of fast neutron transport in chloride salts using Monte Carlo method

The aim of this paper is to present results of fast neutron behavior analysis within the chloride salts environment using simulations based on Monte Carlo method (MCNP 6.2). Three non-fueled salts (NaCl, KCl, MgCl2) and two salts containing fissile material (UCl3, ThCl4) were studied. Results of this theoretical study will be used for design of an experimental assembly, which will serve to achieve goals of the international research project ADAR (Accelerator Driven Advanced Reactor). One of the project objectives is to investigate chloride salts as potential coolant and a dissolved fuel carrier of advanced nuclear reactor cooled by molten salts and driven by an accelerator

Measurement of low-mass e + e − pair production in 1 and 2 A GeV C–C collision with HADES

HADES is a secondary generation experiment operated at GSI Darmstadt with the main goal to study dielectron production in proton, pion and heavy ion induced reactions. The first part of the HADES mission is to reinvestigate the puzzling pair excess measured by the DLS collaboration in C+C and Ca+Ca collisions at 1A GeV. For this purpose dedicated measurements with the C+C system at 1 and 2A GeV were performed. The pair excess above a cocktail of free hadronic decays has been extracted and compared to the one measured by DLS. Furthermore, the excess is confronted with predictions of various model calculations. © 2009 Springer-Verlag / Società Italiana di Fisica. 62 1 81 84 Cited By :

The Lead-Based VENUS-F Facility: Status of the FREYA Project

The GUINEVERE project in the 6th European Framework Program (FP6) [1] aimed to check the methods for sub-criticality monitoring. To execute the project, the water-moderated thermal VENUS facility was modified into the lead fast VENUS-F facility in the period 2007–2010. To prove the reliability of the reactivity monitoring methods, first of all a critical reference configuration was assembled and characterized by measurements of criticality, power distribution, and spectral indexes. These experiments were communicated for benchmarking at ISRD-14 [2]. The Monte Carlo MCNP 5-1.60 code with the JEFF 3.1.2 data library is used to perform simulations of the VENUS-F core, in particular to obtain Calculated-to-Experimental ratios (C/E) for fission rates and spectral indices. A sensitivity study is performed focusing on the impact of global and local parameters on C/E. In most cases C/E is close to unity within the uncertainties. Only a few exceptions were found, e.g. for the F28/F25 spectral index [3]. In order to investigate the discrepancies, a new measurement campaign with the same critical configuration was included in the currently ongoing FREYA project in FP7 [4]. The facility status, experimental plans, and the sensitivity study are presented in this paper

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented

The Lead-Based VENUS-F Facility: Status of the FREYA Project

The GUINEVERE project in the 6th European Framework Program (FP6) [1] aimed to check the methods for sub-criticality monitoring. To execute the project, the water-moderated thermal VENUS facility was modified into the lead fast VENUS-F facility in the period 2007–2010. To prove the reliability of the reactivity monitoring methods, first of all a critical reference configuration was assembled and characterized by measurements of criticality, power distribution, and spectral indexes. These experiments were communicated for benchmarking at ISRD-14 [2]. The Monte Carlo MCNP 5-1.60 code with the JEFF 3.1.2 data library is used to perform simulations of the VENUS-F core, in particular to obtain Calculated-to-Experimental ratios (C/E) for fission rates and spectral indices. A sensitivity study is performed focusing on the impact of global and local parameters on C/E. In most cases C/E is close to unity within the uncertainties. Only a few exceptions were found, e.g. for the F28/F25 spectral index [3]. In order to investigate the discrepancies, a new measurement campaign with the same critical configuration was included in the currently ongoing FREYA project in FP7 [4]. The facility status, experimental plans, and the sensitivity study are presented in this paper

Impact of delayed neutron constants on reactivity effects measured in a fast reactor

Delayed neutron parameters of fast VENUS-F reactor core configurations are determined with Monte Carlo calculations using various nuclear data libraries. Differences in the calculated effective delayed neutron fraction and the impact of the delayed neutron data (6- or 8-group precursors) that are applied in the experimental data analysis on the measured reactivity effects are studied. Considerable differences are found due to application of 235U and 238U delayed neutron data from JEFF, JENDL and ENDF evaluations

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented

Impact of delayed neutron constants on reactivity effects measured in a fast reactor

Delayed neutron parameters of fast VENUS-F reactor core configurations are determined with Monte Carlo calculations using various nuclear data libraries. Differences in the calculated effective delayed neutron fraction and the impact of the delayed neutron data (6- or 8-group precursors) that are applied in the experimental data analysis on the measured reactivity effects are studied. Considerable differences are found due to application of 235U and 238U delayed neutron data from JEFF, JENDL and ENDF evaluations

Calibration of CFUL01 fission chambers in the standard neutron fields of the BR1 reactor at SCK CEN

Recent subcritical VENUS-F experiments showed that fission chambers with a threshold deposit like U-238 can essentially improve the on-line sub-criticality measurments with the beam interruption method, which is currently supposed to be the main method for the ADS MYRRHA. To suppress the uncertainty caused by fissions in the U-235 impurities, the fraction of U-235 in the U deposit should be accurately known. Three PHOTONIS CFUL01 type fission chambers with U-238 deposit were purchased for sub-critical experiments in the VENUS-F reactor. To verify the purity of their deposits, the effective U-235 masses were measured in the empty cavity of the BR1 reactor with a well-known thermal neutron spectrum. It turned out that the measured effective U-235 mass in two fission chambers is lower than the declared mass (as it should be), but this is not the case for the third fission chamber. Then, the effective U-238 mass in these FCs was measured in the well-known fast spectrum of the MARK-III convertor in the BR1 reactor. Finally, the isotopic composition was obtained and it was found that the purity of two CFUL01 FCs is in agreement with the values declared in the certificates but it is not the case for the third fission chamber. As the length of the deposit is bigger than the length of the MARK-III convertor, necessary corrections were calculated with MCNP. The developed procedure using the BR1 standard irradiation fields can be applied for calibration and impurity determination of large fission chambers