Neutron spectra measurement and calculations using data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 in iron benchmark assemblies

The leakage neutron spectra measurements have been done on benchmark spherical assemblies - iron spheres with diameter of 20, 30, 50 and 100 cm. The Cf-252 neutron source was placed into the centre of iron sphere. The proton recoil method was used for neutron spectra measurement using spherical hydrogen proportional counters with diameter of 4 cm and with pressure of 400 and 1000 kPa. The neutron energy range of spectrometer is from 0.1 to 1.3 MeV. This energy interval represents about 85 % of all leakage neutrons from Fe sphere of diameter 50 cm and about of 74% for Fe sphere of diameter 100 cm. The adequate MCNP neutron spectra calculations based on data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 were done. Two calculations were done with CIELO library. The first one used data for all Fe-isotopes from CIELO and the second one (CIELO-56) used only Fe-56 data from CIELO and data for other Fe isotopes were from ENDF/B-VII.1. The energy structure used for calculations and measurements was 40 gpd (groups per decade) and 200 gpd. Structure 200 gpd represents lethargy step about of 1%. This relatively fine energy structure enables to analyze the Fe resonance neutron energy structure. The evaluated cross section data of Fe were validated on comparisons between the calculated and experimental spectra

Analysis of cross sections on iron and oxygen using Cf-252 neutron source

The leakage neutron spectra measurements have been done on benchmark spherical assemblies with Cf-252 source in center of 1) heavy water sphere with diameter of 30 cm (with Cd cover) and of 2) iron spheres with diameter of 100 cm and 50 cm. It has been stated for years that transport calculations by iron overestimate measured spectra in energy region around 300 keV by about 20-40 % (calculation to measurement ratio C/E = 1.2-1.4). The influence of an artificial changes in cross-section XS-Fe-56 (n,elastic)designed by IAEA, Nuclear Data Section, has been studied on the iron spheres. Influence of those XS-corrections to calculated neutron spectrum is presented

Developing Remote Residential Construction Methods.

A systematic study of existing configurations and design methods for developing remote residential construction systems

DESIGN RESEARCH: OPTIMISING ROW-HOUSE ORIENTATION

ABSTRACT A collaborative research study of modular terrace houses in Auckland, New Zealand has sought to find optimal building orientation, façade design and layout to inform the master plan for a housing development. Two terrace house modules of different size within a large multi-unit development were analysed. EnergyPlus models were based on the preliminary design concepts with sufficient zones that the thermal performance of each functional space within the terrace house could be studied. The smaller terrace house had the greatest difference in performance between best and worst orientations. Further investigation determined that the major determinant of this difference was neither size, nor plan layout of the buildings but was the larger imbalance in window to wall ratios of the two exposed facades

Use of nickel sphere and copper cube with

The leakage neutron spectrum measurements have been done on benchmark spherical assembly-nickel sphere with a diameter of 50 cm and a copper cube (block) with dimensions of 49.5 x 49.5 x 48 cm3 in Research Centre Rez (RC Rez). The 252Cf neutron source was placed into the centre of nickel sphere and copper cube. The proton recoil method was used for the neutron spectrum measurement using spherical hydrogen proportional detectors (HPD) with pressure of 400 and 1000 kPa (diameter of detectors is 4 cm) and scintillation stilbene (ST) detector (diam. of 1 x 1 cm). The neutron energy range of spectrometer is from 0.04 MeV to 1.3 MeV for HPD and from 1 MeV to 12 MeV for ST. The adequate MCNP neutron spectrum calculations based on data libraries ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, BROND-3.1 were done and compared with the experiment, i.e., calculation to experiment ratio C/E was determined

Ratio of spectral averaged cross sections measured in standard 252Cf(sf) and 235U(nth,f) neutron fields

The results of systematic evaluations of spectrum averaged cross section (SACS) measurements in the fission neutron fields of 252Cf and 235U are presented. The data form a complete database of high-threshold experimental SACS measured in the same installation under the same conditions and using the same high purity germanium gamma spectrometer. This is crucial to reduce the uncertainty of the ratio and the data scattering and therefore, to minimize discrepancies compared to cross section measured under different conditions in different laboratories. This new dataset complements and extends earlier experimental evaluations. The total emission of the 252Cf neutron source during the experiments varied from 9.5E8 to 4.5E8 neutrons per second. The emission was derived in accordance to the data in the Certificate of Calibration involving absolute flux measurements in a manganese sulphate bath. Concerning 235U fission neutron field, the irradiations were carried out in a specifically designed core assembled in the zero power light water LR-0 reactor. This special core has a well described neutron field. After the irradiation, the low volume irradiated samples to be measured by gamma spectrometry were placed directly on the upper cap of a coaxial high purity germanium (HPGe) detector in a vertical configuration (ORTEC GEM35P4). High volume samples were homogenized and strewn into the Marinelli beaker. The HPGe detector is surrounded by the lead shielding box with a thin inner copper cladding and covered with rubber for suppression of background signal and bremsstrahlung. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas (NPA) measured using the semiconductor HPGe detector. The measured reaction rates are used to derive the spectrum-averaged cross sections. Furthermore, measured reaction rates are also compared with MCNP6 calculations using various nuclear data libraries, in particular IRDFF evaluations

Ratio of spectral averaged cross sections measured in standard

The results of systematic evaluations of spectrum averaged cross section (SACS) measurements in the fission neutron fields of 252Cf and 235U are presented. The data form a complete database of high-threshold experimental SACS measured in the same installation under the same conditions and using the same high purity germanium gamma spectrometer. This is crucial to reduce the uncertainty of the ratio and the data scattering and therefore, to minimize discrepancies compared to cross section measured under different conditions in different laboratories. This new dataset complements and extends earlier experimental evaluations. The total emission of the 252Cf neutron source during the experiments varied from 9.5E8 to 4.5E8 neutrons per second. The emission was derived in accordance to the data in the Certificate of Calibration involving absolute flux measurements in a manganese sulphate bath. Concerning 235U fission neutron field, the irradiations were carried out in a specifically designed core assembled in the zero power light water LR-0 reactor. This special core has a well described neutron field. After the irradiation, the low volume irradiated samples to be measured by gamma spectrometry were placed directly on the upper cap of a coaxial high purity germanium (HPGe) detector in a vertical configuration (ORTEC GEM35P4). High volume samples were homogenized and strewn into the Marinelli beaker. The HPGe detector is surrounded by the lead shielding box with a thin inner copper cladding and covered with rubber for suppression of background signal and bremsstrahlung. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas (NPA) measured using the semiconductor HPGe detector. The measured reaction rates are used to derive the spectrum-averaged cross sections. Furthermore, measured reaction rates are also compared with MCNP6 calculations using various nuclear data libraries, in particular IRDFF evaluations

Spectral averaged cross sections as a probe to a high energy tail of

The systematic evaluations of spectrum averaged cross sections of dosimetric reactions over a broad range of energies were performed in 252Cf (spontaneous fission) and 235U(nth,f) neutron fields. The neutron sources used in this study were LR-0, VR-1 zero power research light water reactors, LVR-15 10 MW research light water reactor, and 252Cf neutron source with emission specified precisely by the manganese sulphate bath. All spectral averaged cross sections were inferred from measured reaction rates which were derived from gamma spectrometry. These gamma spectrometry measurements were performed using a single detector in all cases. The ratios of 252Cf and 235U spectral averaged cross sections can be used to specify the high energy tail of the 235U prompt fission neutron spectrum as the 252Cf spontaneous fission spectrum is considered as a standard. Furthermore, ratios are independent of cross section uncertainties since uncertainties in the cross sections are eliminated. Theoretical models of fission can be tested based on our measurements. The calculations were performed in MCNP6.2 transport code using different prompt fission neutron spectra and IRDFF-II cross sections for threshold reactions. The ratios are in good agreement using only ENDF/B-VIII.0 235U prompt fission neutron spectrum suggesting it to be harder than in other evaluations

Validation of selected (n,2n) dosimetry reactions in IRDFF-1.05 library

Spectrum-averaged cross sections (SACS) have been measured in the reference 252Cf(sf) neutron field for the following high-threshold (n,2n) neutron dosimetry reactions since they are especially important due to the high threshold which allows validation of upper parts of prompt fission neutron spectrum. This work includes 59Co(n,2n)58Co, 197Au(n,2n)196Au, 169Tm(n,2n)168Tm, 55Mn(n,2n)54Mn, 93Nb(n,2n)92 mNb and 89Y(n,2n)88Y and for the 59Co(n,p)59Fe threshold reactions. SACS were inferred from experimentally determined reaction rates by gamma spectrometry using a semiconductor high-purity germanium detector to measure irradiated samples. Measured reaction rates agree within quoted uncertainties with those calculated from the IRDFF-1.05 library, except for the reaction 55Mn(n,2n)54Mn, for which the measured value is underestimated by 16%. For this reaction the ENDF-B/VII.1 evaluation agrees with measured reaction rate within uncertainties.This work has been realized within the SUSEN Project (established in the framework of the European Regional Development Fund (ERDF) in project CZ.1.05/2.1.00/03.0108 and of the European Structural Funds and Investment Funds (ESIF) in the project CZ.02.1.01/0.0/0.0/15_008/ 0000293), which is financially supported by the Ministry of Education, Youth and Sports - project LM2015093 Infrastructure SUSEN