Neutron Self-Shielding Factor of Scattering-Predominant Resonance Foils

The neutron self-shielding factor of ⁵⁹Co resonance foil as an example of foils whose scattering cross section predominate over their absorption cross sections was obtained by both Monte Carlo method (analog) and the collision probability method for various thicknesses of the foil. Also, the transmission and reflection probabilities of neutrons which have various energies near the resonance energy were obtained, and the effects of multiple scattering of neutrons on the neutron self-shielding factor are discussed.

The neutron self-shielding factors obtained by the Monte Carlo method and by the collision probability method agreed well with each other in the cases Σ t ～ 4.0, in which the Monte Carlo method requires considerably longer machine time. Although for the cases of large Σ t (～4.0) the agreement is not always good because of the flat flux approximation in the collision probability method, the calculation time by Monte Carlo is conveniently short. A combination of both methods is useful in obtaining the neutron self-shielding factor of resonance foils.     

Neutron yield in inelastic collisions of 14-Mev neutrons with nuclei and cross section for the (n, 2n) reaction

A measurement has been made of , the number of neutrons produced in one inelastic scattering event between a neutron and a number of elements of natural isotopic composition: Fe, Cu, Mo, Cd, Sn, Sb, Hg, Pb, Bi and U. The measurements were performed by determining the relative change in the total neutron flux and the attenuation of the primary neutrons after passage through samples of the materials being investigated. It was also possible to obtain data on the cross section _in for inelastic collisions of neutrons with the above-mentioned nuclei. The values of and _in, in conjunction with the known cross sections for neutron capture, were then used to compute the cross section for the (n, 2n) reaction (averaged over the isotopic composition) in nonfissile nuclei.This work was completed in 1952.The authors wish to thank A.A. Malinkin for comparing the neutron yields from sources used to measure the dependence of counter sensitivity on neutron energy.     

Measurements of Average Cross Section for 232Th(n, 2n)231Th Reaction to Neutrons with Fission-Type Reactor Spectrum and of Gamma-Ray Intensities of 231Th

The average cross section for the ²³²Th(n, 2n)²³¹Th reaction to neutrons with the energy spectrum close to that of fission neutrons was obtained in the core of the Kyoto University Reactor, KUR. The value obtained was 12.5±0.84 mb. This value agrees satisfactorily with Phillips' and with the calculated value obtained with the cross section in the U-K library and the Maxwellian fission neutron spectrum given by Leachman. A somewhat poorer agreement is seen with the calculated value obtained from Butler & Santry's cross section and Leachman's spectrum. The discrepancy amounts to 24 and to 39% respectively, for the average cross sections calculated with these two excitation functions and the fission neutron spectrum given by McElroy.

By making use of a Ge(Li) counter whose photopeak efficiency had been carefully calibrated, the absolute intensities were determined for eleven photopeaks observed on the γ-ray spectrum emitted by ²³¹Th.     

Measurements and model calculations of activation cross sections for 232Th(n,2n)231Th reaction between 13.57 and 14.83 MeV neutrons

In this study, activation cross sections were measured for the reaction of ²³²Th(n,2n)²³¹Th (T_1/2 = 25.5 h) by using neutron activation technique at six different neutron energies from 13.57 and 14.83 MeV. Neutrons were produced via the ³H(²H,n)⁴He reaction using SAMES T-400 neutron generator. Irradiated and activated high purity Thorium foils were measured by a high-resolution γ-ray spectrometer with a high-purity Germanium (HpGe) detector. In cross section measurements, the corrections were made for the effects of γ-ray self-absorption in the foils, dead-time, coincidence summing, fluctuation of neutron flux, low energy neutrons. For this reaction, statistical model calculation, which the pre-equilibrium emission effects were taken into consideration, were also performed between 13.57 and 14.83 MeV energy range. The cross sections were compared with previous works in literature, with model calculation results, and with evaluation data bases (ENDF/B-VII, ENDF/B-VI, JEFF-3.1, JENDL-4.0, JENDL-3.3, and ROSFOND-2010).     

Calculated neutron cross sections for Cu and Nb up to 32 MeV for neutron damage analysis

Cross sections for neutron interaction with Cu and Nb, with emphasis on spectra of light particles from binary reactions, are calculated for neutron energies from 4 to 32 MeV for estimating recoil probability densities for the analysis of damage experiments with a Be (d, n) neutron source. Nuclear model parameters were adjusted to reproduce the available cross-section data around 14 MeV. Helium production cross sections were also calculated for ⁶³Cu for neutrons below 20 MeV, as an illustration of the Hauser-Feshbach method for calculating tertiary reaction cross sections.     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

Measurements of the Thermal Neutron Macroscopic Absorption Cross Section for Neutron Absorbing Layers

Objective of this study is measuring the macroscopic cross section of a neutron absorbing layer for thermal neutrons. For this purpose a neutron source and BF ₃ detector have been applied. For measuring macroscopic cross section of thermal neutrons by the \( I = BI_{0} e^{{ - \sum\nolimits_{tot} t}} \) Formula, it is necessary to provide suitable geometric conditions in order to assume the production and build-up coefficient to be the unit value (=1). To fulfill required conditions for this assumption, surface of the detector is covered with a 2 mm thick layer of cadmium. Radiation window of the detector has a 3 cm diameter, situated directly in front of the source. By placing the cadmium cover over the detector, variation of \( Ln{\frac{{I_{0}^{{}} }}{I}} \) values verses thickness of absorbent layer, renders linear function behavior, making it possible to measure the macroscopic cross section. The next stage is applying the MCNP code by simulating F1 tally and cosine-cards for calculating Total Macroscopic Cross-Section. Validation of this study is achieved through comparison of simulation by the MCNP code and results rendered by experiment measurements.     

Emission Probabilities of Gamma Rays from the Decay of 233Pa and 238Np,and the Thermal Neutron Capture Cross Section of 237Np

In order to determine the thermal neutron capture cross section of ²³⁷Np, the relevant γ emission probabilities of the 312-keV γ-ray from the decay of ²³³Pa and the 984-keV γ-ray from the decay of ²³⁸Np are deduced from the ratio of the emission rate to the activity. The emission rate and activity are measured with a Ge detector and a Si detector, respectively. The measured emission probability for 312-keV γ-ray is 41.6±0.9% and that for 984-keV γ-ray is 25.2±0.5%. The emission probabilities are used to correct the thermal neutron capture cross section of ²³⁷Np reported previously, and gives 168±6b. The neutron capture cross section is also determined as 169±6b by α-ray spectroscopic method. The measured emission probabilities and capture cross section are compared with others from references. By averaging these values deduced by different methods, the value of 169±4b is recommended as the thermal neutron capture cross section of ²³⁷Np for 2,200 m/s neutrons.     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

Analysis of Heavy-Water-Moderated,Cluster-Type Fuel Lattices by Cluster Physics Code ‘MESSIAH’

This paper presents a comprehensive analysis performed by a new cluster analysis code ‘MESSIAH’ on reactor physics constants measured in the critical facility for a pressure-tube-type, heavy-water-moderated reactor. The MESSIAH code system utilizes the method of the collision probability to solve the neutron transport equation. The effective space dependent cross sections are calculated in the thermal and resonance energy range before the eigenvalue calculation for the whole energy range. With use of these cross sections, the multi-group, space dependent transport equation is solved, and the flux distribution, spectrum and k _eff are obtained to the input bucking. In the above three steps the method of the collision probability is used consistently and extensively. The treatment of leakage neutrons from lattices in MESSIAH is also confirmed by an independent method using a Monte Carlo calculation. The calculated reactor physics constants, especially the micro-parameters and the activation traverse of Dy, agreed fairly well with the experiment. The diffusion calculation with use of the group constants calculated by MESSIAH predicts the reactivity of 0% void core excellently (<0.12%). However, for a 100% void core, the calculated reactivity was slightly lower than the experiment (~0.74%), which was attributed to over prediction of the diffusion constants.     

Application of the response matrix method to BWR lattice analysis

A lattice calculation code RESPLA has been developed for light-water reactor lattices on the basis of the response matrix method treating the heterogeneity in pin cells. The spatial dependency of neutron flux distribution along each cell boundary is taken into account by dividing the cell boundary into several subsurfaces and the anisotropy of neutron angular distribution is considered up to the P₁ component by using a relation between the P₀ and P₁ components. The RESPLA code has been applied to BWR lattice calculations and the calculational results have been compared with those obtained by the S_n method and the collision probability method. It has been found that the present response matrix method has the same accuracy as the collision probability method with fine spatial meshes and the error caused by the use of coarse meshes is much smaller than that by the collision probability method. Furthermore, the required computing time is smaller by about a factor of five than that in the collision probability method.     

Improvement of Neutron Source Introduction Method for Absolute Measurements of Low Reactor Power

The neutron source introduction method was applied to absolute measurements of low reactor power at the Static Experiment Critical Facility STACY. To obtain the effective neutron source intensity more accurately, which is a key parameter for the source introduction method, the neutron source is newly defined as fission neutrons from the first fission reaction caused by neutrons emitted from the external neutron source. To obtain the newly defined effective neutron source intensity, the probability that a neutron from the external neutron source causes a fission reaction is calculated using the Monte Carlo code MCNP. This calculation took into consideration the three-dimensional complicated core structures. Furthermore, the fission reaction distribution, fundamental mode forward and adjoint flux distribution in a critical state were calculated using the three-dimensional transport code THREEDANT. Following the principle of the neutron source introduction method, an external neutron source was inserted near the STACY core tank and the reactor power was measured. The reactor powers by the neutron source introduction method were in good agreement with the ones from the analyses of the FP activity generated by high power operation.     

Cross sections for neutron-induced reactions up to 1.6 GeV for target elements relevant for cosmochemical, geochemical, and technological applications

We determined neutron excitation functions from the respective reaction thresholds up to 1.6 GeV for almost 100 target-product combinations relevant for cosmochemical, geochemical, and technological applications. We started with thick target production rates that have been obtained by irradiating iron and stone spheres with protons having energies between 600 MeV and 1.6 GeV. From the particle spectra of primary protons, secondary protons, and secondary neutrons, and the usually well known cross sections for the proton-induced reactions we calculated the production rates only due to protons. By subtracting these data from the measured total production rates we obtained production rates only due to secondary neutrons. With the modelled neutron spectra, guess functions calculated using nuclear model codes, and sophisticated energy-dependent deconvolution procedures we were able to determine almost 100 neutron excitation functions with their uncertainties. With the thus obtained neutron cross sections we are able to describe the experimental production rates in the thick target experiments, meteorites, the lunar surface, and terrestrial surface samples usually within the uncertainties, i.e., to within 10-15%. The adjusted neutron cross sections (a posteriori) are compared to results from the theoretical nuclear model codes INCL4.5/ABLA07 and TALYS. The TALYS code usually describes the a posteriori data reasonably well, i.e., mostly within a factor of a few. The quality of the INCL4.5 + ABLA07 predictions depends on the reaction type and increases with increasing number of ejectiles, i.e., increasing target-product mass difference. The neutron cross section database, though successful in quantifying production rates in terrestrial and extraterrestrial matter, presents by no means a final step and experiments with quasi-monoenergetic neutrons are needed.     

Measurement and analysis of reaction rates in a large spherical depleted uranium pile surrounding a 14-MeV neutron source

Measurement was made of the reaction rate distributions for ²³⁸U(n, f), ²³⁵U(n, f), ²³⁸U(n, γ), ²⁷Al(n, α) and ⁵⁸Ni(n, p) in a large depleted uranium (DU) pile. The pile consisted of DU blocks forming a spherical shell of 45.72 cm radius and 40.64 cm thick. 14-MeV neutrons were generated at the center. Fast neutron leakage spectrum was also measured by an NE-213 spectrometer. In order to assess the ²³⁸U neutron cross sections of JENDL-2, the experiment was analyzed using the Monte-Carlo transport code MCNP with continuous energy cross sections. The agreement between the calculation and the experiment is generally unsatisfactory. The ratios of calculation to experiment of low energy reactions decreased with the thickness of the DU layer. The analysis of the Livermore pulsed sphere experiment for the small DU sphere revealed underestimation of leakage neutron spectrum around 10 MeV. The ²³⁸U cross sections of JENDL-2 should be improved for 14-MeV neutronics calculation.     

Indirect measurement of inelastic cross section of relativistic protons in Pb target

The inelastic cross section of relativistic protons in Pb was determined indirectly by measuring the neutron distribution along a Lead spallation neutron source. The spallation neutron source was irradiated by 1, 1.5 and 2 GeV protons. The experimental results were obtained using passive methods. By the use of the beam attenuation coefficient, deduced by a fitting procedure of experimental data, the inelastic cross section of protons in Pb was determined.     

Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix

We present a methodology to propagate nuclear data covariance information in neutron source calculations from (α,n) reactions. The approach is applied to estimate the uncertainty in the neutron generation rates for uranium oxide fuel types due to uncertainties on 1) ^17,18 O(α,n) reaction cross-sections and 2) uranium and oxygen stopping power cross sections.The procedure to generate reaction cross section covariance information is based on the Bayesian fitting method implemented in the R-matrix SAMMY code. The evaluation methodology uses the Reich-Moore approximation to fit the ^17,18 O(α,n) reaction cross-sections in order to derive a set of resonance parameters and a related covariance matrix that is then used to calculate the energy-dependent cross section covariance matrix. The stopping power cross sections and related covariance information for uranium and oxygen were obtained by the fit of stopping power data in the α-energy range of 1 keV up to 12 MeV.Cross section perturbation factors based on the covariance information relative to the evaluated ^17,18 O(α,n) reaction cross sections, as well as uranium and oxygen stopping power cross sections, were used to generate a varied set of nuclear data libraries used in SOURCES4C and ORIGEN for inventory and source term calculations. The set of randomly perturbed output (α,n) source responses, provide the mean values and standard deviations of the calculated responses reflecting the uncertainties in nuclear data used in the calculations. The results and related uncertainties are compared with experiment thick target (α,n) yields for uranium oxide.     

Analysis of neutron yield produced by high-energy proton

Comparisons have been made between computational results obtained with the BNL code system and experimental data measured by Vasfl'kov et al. for 56 x 56 x 64 cm natural and depleted uranium blocks surrounded by lead walls and primary proton energies of 400 and 660 MeV. The energetic protons from a linear accelerator are used to produce an intensive neutron source inthe uraniumblocko The computercode system prepared at BNL to perform nuclear design analyses of linear accelerator reactors consists of six main programs: NMTC for spallation-evaporation processes above 15 MeV, HIST3D for the analysis of collision event records obtained by NMTC to get P₃ neutron source distribution, DLC-2 to compile 100 energy group cross sections, TAPEMAKER for format conversion, ANISN to collapse 100 group cross sections to fewer group P₃ cross section sets, and the principal code TWOTRAN-II which performs neutron reaction and transport calculations in the energy range below 15 MeV. Our computational method gives conservative total neutron yields, i.e., underestimates of about 16.8–29.8% in comparison with measured values depending on proton energy. Radiative capture²³⁸U(n,) density distributions have been compared between the calculation and experiment. The calculated distribution has the higher peak in the central part of the target system and the steeper gradient both in the r and z directions.Brookhaven National Laboratory, Upton, New York 11973. Published in Atomnaya Énergiya, Vol. 47, No. 2, pp. 83–91, August, 1979.     

Monte Carlo study of photoneutron production in U-235 following perturbations in cross section data

Active photon interrogation systems may be employed to detect high-Z isotopes without significant spontaneous fission emissions. These systems induce photonuclear reactions with emissions (such as fission neutrons) that may be detected. However, there are inconsistencies in the literature reporting resonance photonuclear interaction data for many isotopes. Recent publications show variations as large as 20% between various measurements of photonuclear cross section data. A perturbation methodology utilizing the modular nature of the MCNPX/MCNP-PoliMi code system has been implemented and is applied here to highly-enriched uranium. Monoenergetic photon sources between 8 and 18 MeV were simulated; neutron detection was performed using the MCNP-PoliMi liquid scintillator model. At photon energies less than 12 MeV, the number of detected neutrons is approximately 70% sensitive to changes in the (γ, f) cross section and 30% sensitive to changes in the (γ, n) cross section. As gamma-ray energy increases the (γ, f) sensitivity increases and the (γ, n) sensitivity decreases. There is a small (γ, 2n) sensitivity at photon energies between 15 and 17 MeV. The ability of modern simulation tools to predict photonuclear responses is greatly limited in this energy region due to the high sensitivity of the simulated results to observed discrepancies in photonuclear cross section data.     

0.04–4.0 Mev neutron fission cross section of Pu240

A detailed investigation of the Pu²⁴⁰ nucleus cross section is of interest with regard to an experimental verification of theoretical concepts of the energy dependence of fission probability as well as regarding possible uses of Pu²⁴⁰ as a nuclear fuel in fast neutron reactors.We measured the energy dependence of the fast neutron fission cross section of Pu²⁴⁰ for neutrons with the energy E_n = 0.04–4.0 Mev. The T(p, n)He³ reaction served as the neutron source. The Pu²⁴⁰ fission cross section in the plateau region (1–4 Mev) amounts to ~ 1.6 barn and is equal to only one-half of this value for the neutron energy E_n 0.7 Mev. A sharp decrease in the fission cross section value occurs as E_n decreases to 0.3 Mev; for a further decrease in E_n, the cross section value drops less sharply, and it remains practically constant (~ 0.065 barn) for 0.04 < E_n < 0.15 Mev. The correlation between the irregularities in fission cross section values and the levels of Pu²⁴⁰ nuclei which correspond to inelastic scattering channels is discussed.The authors extend their heartfeltthanks to A. I. Leiptmskii and I. I. Bondarenko for their helpfulness and interest in the work, to L. N. Usachev for the discussion of the results, to Yu. I. Baranov and N. E. Tokmantseva for their help in measurements, and to V. A. Romanov, G. A. Strigin, and Yu. I. Parfenov, who kept the accelerator in good running order.     

Determination of critical mass and neutron flux distribution by means of physical models

The design of nuclear reactors, especially new reactors, requires experimental measurements in order to obtain accurate values of the pertinent parameters. In the present paper we present a new method for the preliminary determination of the critical mass of a reactor and the neutron flux distribution; this method is based on the use of physical models. In carrying out these experiments use is made of a model of the reactor which does not contain fissionable material. The working channels in the model are filled with a neutron absorber whose cross section simulates the absorption cross section for neutrons in the fissionable material. The production of fast fission neutrons is simulated by means of a neutron source which is moved along the channels. The distribution of thermal neutrons is measured by means of detectors which are sensitive to thermal neutrons. If the source strength and the absolute value of the neutron flux are known, it is possible to find the critical mass of the reactor.This method has been checked in a reactor with uranium hexafluoride. The value of the critical mass found experimentally was found to be in good agreement with the value obtained when the reactor was started up.The proposed method can also be useful in preliminary investigations of reactor designs, the choice of optimum lattice parameters, etc. The technique is extremely simple and does not require fissionable material or high neutron fluxes.