Prompt and delayed gamma rays induced by epithermal and fast neutrons with indium

The emission of prompt and delayed gamma rays from (n,γ) and (n,n´γ) reactions induced by irradiation of indium with epithermal and fast neutrons was investigated with the instrument FaNGaS operated at Heinz-Maier-Leibnitz Zentrum (MLZ) in Garching. The average neutron energy of the neutron spectrum was 2.30 MeV. The measurement was done at an angle of 90° between neutron beam and detector. A total of 136 prompt gamma lines from which 42 are related to the capture of epithermal and fast neutrons and 94 to the inelastic scattering of fast neutrons were detected together with the delayed gamma lines of the activation products 113mIn, 114m2In, 115mIn, 116m2In and 116mIn. Intensities and neutron spectrum averaged isotopic partial cross section of the gamma lines are presented. Additionally the neutron spectrum averaged cross sections of the reactions, 113In(n,n´)113mIn, 113In(n,γ)114m2In, 115In(n,n´)15mIn, 115In(n, γ)116m2In and 115In(n, γ)116mIn were determined from the corresponding delayed gamma rays of the formed isotopes as 143 ± 22, 288 ± 13 194 ± 18, 201 ± 10 and 508 ± 24 mb respectively. The various results obtained were found consistent with the literature data. However, our measurement indicate the need to reevaluate the cross section of the 115In(n,γ)116m2In reaction for thermal neutrons.

Measurements of the cross-section of 111Cd(n,n′)111mCd reaction for 241Am/Be neutrons

The cross-section of the [Formula: see text] reaction was measured with [Formula: see text] neutrons using a natural cadmium target [Formula: see text]. The neutron fluence and mean neutron energy of the source were determined using the ISO 8529-1 neutron spectrum and the known cross-sections of the monitor reaction [Formula: see text]. In order to measure the poor [Formula: see text]-ray activity of the reaction products, an HPGe detector with 70% detection efficiency surrounded by an adequate graded shield was applied. The efficiency calculations for the detector were performed using standard point calibration sources and the EFFTRAN efficiency code. Using the measured values of the neutron flux and the induced [Formula: see text]-ray activity of [Formula: see text], the cross-section of the [Formula: see text] reaction at the average neutron energy of 4.05 MeV was found to be [Formula: see text] mb. An estimation of the contribution to the total cross-section by the accompanied reactions [Formula: see text] and [Formula: see text] was achieved and the related cross-sections were found to be 0.16 mb and 8.99 mb, respectively.

NEUTRONIC MODELING OF A FAST CORE WITH MODERATING MATERIALS USING APOLLO3R CODE

In order to improve passive safety of Sodium-cooled Fast Reactors The French Alternative Energies and Atomic Energy Commission (CEA) has proposed a new core design called CADOR - an SFR core with enhanced Doppler reactivity feedback. One of its most important design features is the introduction of solid moderating materials inside each fuel assemblies to slightly decrease the average neutron energy. The article focuses on development and validation of a neutronics calculation scheme able to produce accurate results in case of CADOR and other fast cores with moderating materials. The study uses two different fuel assembly models moderated by metallic beryllium and zirconium hydride (ZrH2) respectively The study includes discussion of neutron scattering treatment and different ways of spatial homogenization and energy condensations. The results indicate that the accurate scattering treatment leads to much better estimation of Doppler constant, especially in case of ZrH2 moderated core. By using combined deterministic-Monte Carlo calculation scheme we are able to quantify the biases on global reactivity, reactivity feedbacks and control rod worth. We demonstrate that spatial homogenization plays a more important role in case of moderated CADOR assemblies and thus preserving certain level of heterogeneity within fuel assemblies can lower the calculation bias significantly.

Reactor with metallic fuel and lead-208 coolant

The paper considers the concept of a fast lead cooled 25MW reactor for a variety of applications, including incineration of minor actinides, production of medical radioisotopes, testing of radiation-damaged nuclear technology materials, etc. A specific feature of the proposed reactor is rather a high neutron flux of 2.6·1015 n/(cm2·s) at the core center, high average neutron energy of 0.95 MeV at the core center, and a large fraction (40%) of hard neutrons (En > 0.8 MeV). The extremely high estimated reactor parameters are achieved thanks to the small core dimensions (DxH ≈ 0.50×0.42 m2), innovative metallic fuel of the Pu-Am-Np-Zr alloy, and the 208Pb enriched lead coolant. A relatively high probability of 241Am fission (about 50%) is achieved in the reactor core’s hard spectrum, this making it possible to incinerate up to 4 kg of 241Am during one reactor campaign of 1000 effective days.

Impact of average neutron energy on the fast neutron fluency measurement by Np237 fission to capture ratio and reverse dark current of planar silicon detector methods

This work is a subsequent step to study the feasibility of fast neutron fluency measurements using two different complementary methods. Np-237 samples and planar silicon detectors were mounted very close to each other on different sections of a subcritical assembly irradiated with the proton beam of 0,66 GeV (the Quinta assembly at the Joint Institute for Nuclear Research, Dubna, Russia) to provide both samples with the same neutron fluency. We have processed the experimental data of irradiated Np-237 actinide samples and silicon detectors directly placed on two sections of the QUINTA setup without a lead shield-reflector. Applying the try and error method we have found found that the neutron energy for which the ratio of the fission cross section to the capture cross section of the actinide Np-237 from the nuclear data base is equal to the measured ratio of the fissioned and captured actinide isotopes. The retrieved distinct fission and capture cross sections for the distinct neutron energy from the nuclear data base describe the average values. The considered above experimental and earlier obtained data have been shown that the higher is the average neutron energy the smaller is the difference of the neutron fluency measurement between the two methods. This effect has been expected since the silicon detector method efficiently measures the fast neutrons of the energy higher than about 170 keV while the actinide method covers a wider energy range.

A Parametric Study of Fuel Lattice Design for HTR-10

In this study, the multiplication factor and neutron spectrum behaviors were investigated against the moderator-to-fuel ratio, the fuel loading height, and the detector location in high-temperature gas-cooled reactor (HTR)-10. The MCNP5 computer code (version 1.51) was employed to perform all the simulation computations. The results revealed that the multiplication factor varies significantly depending on the moderator-to-fuel ratio and the fuel loading height due to the competition among the neutron moderation and absorption abilities of the moderator as well as the neutron production ability of the fuel. Due to its inherent stability, HTR-10 is deliberately designed such that the multiplication factor decreases and the neutron spectrum softens as the moderator-to-fuel ratio increases. The average neutron energy level in the HTR-10 fuel balls is approximately 240 keV and ranges from smallest to largest at the middle, bottom, and top of the reactor core, respectively.

A Parametric Study of Fuel Lattice Design for HTR-10

In this study, the multiplication factor and neutron spectrum behaviors were investigated against the moderator-to-fuel ratio, the fuel loading height, and the detector location in HTR-10. The MCNP5 computer code (version 1.51) was employed to perform all the simulation computations. The results revealed that the multiplication factor varies significantly depending on the moderator-to-fuel ratio and the fuel loading height due to the competition among the neutron moderation and absorption abilities of the moderator as well as the neutron production ability of the fuel. Due to its inherent stability, HTR-10 is deliberately designed such that the multiplication factor decreases and the neutron spectrum softens as the moderator-to-fuel ratio increases. The average neutron energy level in the HTR-10 fuel balls is approximately 200 keV and ranges from smallest to largest at the middle, bottom, and top of the reactor core, respectively.