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.

Core parametric study for enhancing the radioisotope production in the IEA-R1 research reactor

In this work a parametric study was carried to increase the production of radioisotopes in the IEA-R1 research reactor. The changes proposed to implement in the IEA-R1 reactor core were the substitution of graphite reflectors by beryllium reflectors, the removal of 4 fuel elements to reduce the core size and make available 4 additional locations to be occupied by radioisotope irradiation devices. The key variable analyzed is the thermal neutron flux in the irradiation devices. The proposed configuration with 20 fuel elements in an approximately cylindrical geometry provided higher average neutron flux (average increment of 12.9 %) allowing higher radioisotope production capability. In addition, it provided 4 more positions to install irradiation devices which allow a larger number of simultaneous irradiations practically doubling the capacity of radioisotope production in the IEA-R1 reactor. The insertion of Be reflector elements in the core has to be studied carefully since it tends to promote strong neutron flux redistribution in the core. A verification of design and safety parameters of the proposed core was carried out. The annual fuel consumption will increase about 17 % and more storage space for spent fuel will be required.

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.

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.

Parameter Optimization and Sensitivity Studies of Spontaneous Fission with FREYA

For many years, the state of the art for simulating fission in transport codes amounted to sampling from average distributions. However, such "average" fission models have limited capabilities. Energy is not explicitly conserved and no correlations are available because all particles are emitted independently. However, in a true fission event, the emitted particles are correlated. Recently, Monte Carlo codes generating complete fission events have been developed, thus allowing the use of event-by-event analysis techniques. Such techniques are particularly useful because the complete kinematic information is available for the fission products and the emitted neutrons and photons. It is therefore possible to extract any desired observables, including correlations. The fast event-by-event fission code FREYA (Fission Reaction Event Yield Algorithm) generates large samples of complete fission events, employing only a few physics-based parameters. A recent optimization of these parameters for the isotopes in FREYA that undergo spontaneous fission is described and results are presented. The sensitivity of neutron observables in FREYA to the input yield functions is also discussed and the correlation between the average neutron multiplicity and fragment total kinetic energy is quantified.

Neutron Multiplicity Correlations with Fission Fragment Mass and Energy from 239Pu(n,f)

There exists experimental evidence for strong fluctuations of the average neutron multiplicity from resonance to resonance in 239Pu(n,f). These fluctuations have been shown to impact nuclear reactor benchmarks by reducing the criticality. The fluctuating neutron multiplicity can be explained as a consequence of the competition between direct fission and the (n,γf) process. However, there is also evidence for fluctuations of the fission fragment mass yields from resonance to resonance. The mass yield fluctuations may also contribute to fluctuations of the neutron multiplicity averaged over all fission fragment masses. In order to model the contribution to the neutron multiplicity fluctuations by the fission fragment mass yield fluctuations new data on the correlations between fission fragment properties and neutron multiplicities are in need. We present experiments carried out to determine prompt neutron multiplicity correlations with fission fragment masses and total kinetic energies in the reaction 239Pu(n,f). The experiment has been performed at the GELINA facility at JRC-Geel. A twin position-sensitive Frisch-grid ionization chamber is used for fission fragment identification via the double kinetic energy technique. An array of scintillation detectors is employed for neutron counting. Correlations between average neutron multiplicities and fission fragment properties have been measured with improved resolution in both mass and TKE, compared to data from the literature.

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.