Measurement of Nuclear Inelastic Scattering Cross Section of Thorium-232 for 144keV Si-Filtered Neutrons

The nuclear inelastic scattering cross section of ²³²Th has been measured at 144 keV using a Si-filtered neutron beam from the University of Missouri Research Reactor (MURR). The energy spectrum of the scattered neutrons was measured with a spherical hydrogen-gas counter and a pulse-height-unfolding procedure. Data were taken at nine scattering angles from 30° to 150° in 15° increments. The results provide new experimental information below 250 keV. The angle-integrated cross section is 0.74±0.05 barn, which agrees with the evaluation JENDL-2 and with a coupled-channel calculation where the inelastic scattering through the direct-excitation process of the collective rotational motion of a deformed nucleus is included as well as that through the compound-nucleus-formation process. Experimental results have been obtained also for the angular dependence of the elastically scattered neutrons at 144 keV.     

Effect of Cold Neutrons on the Neutron Pulse and Wave Propagations in Crystalline Media

With the view to contributing information on the effect of cold neutrons on neutron pulse propagation and on die-away phenomena and phenomena related to neutron wave propagation, measurements of propagating pulse shapes in graphite and lead prisms were carried out with 300°K water moderated and 77°K ice moderated sources. Corresponding theoretical analyses were also performed, with use made of the successive iteration method, and a coupled two- group theory based on the distributed source diffusion equation was developed.

As a result, it was established that: (a) infiltrating cold neutrons below Bragg-cut-off energy, which has a sharply peaked pulse head, produce peculiarly shaped pulse propagation responses from detector; (b) the resonance phenomena observed in wave propagation can be fairly well explained as due to the interference between the thermal neutron group and the infiltrating cold neutron group, the cold neutrons in question being those below the Bragg cutoff energy in the case of graphite, and in the case of lead those below 0.01 eV; and (c) in the analyses related to resonance phenomena in graphite, the experimental characteristics are more consistently represented in the results of calculations using the BNL-325 data as compared to those using the UNCLE data for the total cross section below the Bragg cut-off.     

The effect of temperature on the diffusion length and the scattering cross section of thermal neutrons in graphite

The effect of heating graphite on the diffusion length and effective scattering cross section of thermal neutrons was investigated. It was established that in the 15–350 °C range the diffusion length changes mainly in accordance with the law 1/v for the absorption cross section. The slight deviation from this law is due to the increase of 0.5 mb/deg in the scattering cross section with an increase in temperature.     

Effects of Reflector on Intensity of Thermal Neutrons Emitted from Moderator for Pulsed Neutron Source

Intensity of the thermal neutrons emitted from the moderator with a reflector was calculated to study the effects on the intensity caused by a macroscopic total neutron cross section and an average logarithmic energy loss of the reflector materials.

A reflector with a large macroscopic total neutron cross section produced higher thermal neutron intensity than that with a small cross section if they had the same average logarithmic energy loss. Among the reflectors with the same total macroscopic neutron cross section, the thermal neutron intensity was not changed by decreasing the average logarithmic energy loss to a range less than about 0.1 but above this value the intensity was weakened. From this result it was found that a large macroscopic total cross section and a small average logarithmic energy loss are preferable characteristics for reflector materials.

As actual reflector materials, three reflector materials were examined, namely beryllium, graphite and lead, which are now considered to be candidates. The lead reflector was effective for the moderator with a large emission-surface and the beryllium reflector for the moderator with a small one. This result indicates that the moderator size is important for choosing the best reflector material to produce the highest beam intensity.     

Development of high-frame-rate neutron radiography and quantitative measurement method for multiphase flow research

In relation to the establishment of thermal neutron radiography as a measurement method with high accuracy and reliability, this paper reviewed the present status on the development of high-frame-rate neutron radiography with a steady thermal neutron beam and its application to multiphase flow researches. This review included also the present progress on the quantification of neutron radiographic image at Kyoto University, i.e. (1) quantitative method to measure void fraction of two-phase flow with thermal neutron radiography (Σ-scaling method), (2) influence of scattered neutrons on void fraction measured by neutron radiography, (3) measurement error of neutrons in a low neutron flux field, (4) error in void fraction measurement due to low gray level, and (5) measurement error due to low imaging speed Moreover, a new experimental approach on a total macroscopic cross section for thermal neutrons measurement by neutron radiography was presented. This paper revealed neutron radiography to be a promising visualization and measurement method in thermal hydraulic research.     

Geometrical Interpretation of Mutual Relatonship between Ordinary Coherence and Noise Power Contribution Ratio

Angular neutron fluxes leaking from the surface of lithium-oxide and graphite slab assemblies have been measured with irradiation of D-T neutrons. The spectrum measurement was performed using the time-of-flight technique with an NE213 scintillation detector. The thicknesses of the slabs were 0.6 to 5 mean free path for 14.8 MeV neutrons, and the measured leaking angles of the angular fluxes were 0.0°, 12.2°, 24.9°, 41.8° and 66.8°. The experimental results have been compared with the results calculated by the continuous energy Monte Carlo transport code MCNP, using the data in the JENDL-3PR1, ?3PR2, and ENDF/B-V nuclear data files. The comparisons between the experimental and calculated results show that the data of ⁷Li in JENDL-3PR2 is improved for the secondary emission spectra of the 4.63 MeV level and (n, 2n) reactions; the angular distributions of 3rd-and 4th-level inelastic reactions of C in the JENDLs are questionable. The thickness dependences for high energy neutrons also suggest that the total cross section of ⁷Li and the elastic cross sections of C are slightly inadequate.     

Angular distribution of elastically and inelastically scattered 2.9-Mev neutrons

A hydrogen ionization chamber and annular geometry are used to measure the angular distribution of elastically and inelastically scattered 2.9-Mev neutrons, as well as inelastically scattered neutrons associated with excitation of various levels or groups of levels of iron, copper, lead, and bismuth nuclei. The integral cross sections for elastic and inelastic scattering are presented, as are the transport cross sections. The experimental results are compared with theoretical calculations based on the optical model. It is noted that the angular distribution of elastically scattered neutrons from atoms with almost equal atomic weights may be quite different.     

Measurements of Pseudo-Decay-Constants of Trapped Neutrons in Small Graphite Systems at 77 and 194°K

Abstract

Neutron pulse die-away experiments for small graphite assemblies by using a pulsed “cold” neutron source technique were carried out for the purpose to measure the pseudo-decay-constants of trapped neutrons at the lowest Bragg-peak energy E _B of coherent scattering in graphite, and also to estimate the temperature dependence of inelastic scattering cross section of neutrons at E _B . Experiments were carried out for 77 and 194°K graphite systems with dimensions from 40x40x40 cm³ to 30x30x20 cm³.

The experimentally determined pseudo-decay-constants showed distinct temperature dependence and good agreement with the theoretical inelastic scattering cross sections by Young-Koppel model of phonon frequency distribution of graphite at these low temperatures.     

基于白光中子源的核石墨硼当量测量

核材料中热中子吸收截面高的杂质会引起堆芯反应性的变化,一般用硼当量表示这些杂质对热中子的吸收,硼当量是衡量核材料纯度的重要指标之一。热中子宏观吸收截面法是硼当量测量的方法之一,测量时采用同位素中子源则精度低,而白光中子源产生的中子强度高、方向性好,且可慢化为热谱,能有效提高硼当量测量精度。本文基于15 MeV电子加速器驱动的白光中子源开展核石墨硼当量测量的研究,利用蒙特卡罗模拟并优化实验方案,对实验数据进行检验与修正,建立核石墨硼当量测量定量分析方法。该方法能快速、准确检测核材料的硼当量,对反应堆的物理设计、安全性评估等具有重要意义。     

Boron Equivalent Measurement of Nuclear Graphite with Photoneutron Source

Impurities in nuclear materials with high thermal neutron absorption cross section will change the reactivity. The absorption of thermal neutrons by these impurities is represented by boron equivalent, which is one of the important factors to measure the purity of nuclear materials. Boron equivalent can be determined directly via the measurement of macroscopic thermal neutron absorption cross section based on an isotopic neutron source, but with lower accuracy. The photoneutron source, which can generate neutrons with higher intensity, better direction and lower energy, can effectively improve the accuracy of boron equivalence measurement. Therefore, the boron equivalent measurement of nuclear graphite was carried out with the photoneutron source driven by 15 MeV electron LINAC. Monte Carlo simulation method was used to optimize the experimental scheme, and the experimental data were tested and modified. Finally, the quantitative analysis method was established for the measurement of graphite boron equivalent. This method can quickly and accurately measure the boron equivalent of nuclear materials, which is of great significance for the physical design and safety assessment of the reactor.     

Angular distribution of elastically and inelastically scattered 2.9-Mev neutrons

A hydrogen ionization chamber and annular geometry are used to measure the angular distribution of elastically and inelastically scattered 2.9-Mev neutrons, as well as inelastically scattered neutrons associated with excitation of various levels or groups of levels of iron, copper, lead, and bismuth nuclei. The integral cross sections for elastic and inelastic scattering are presented, as are the transport cross sections. The experimental results are compared with theoretical calculations based on the optical model. It is noted that the angular distribution of elastically scattered neutrons from atoms with almost equal atomic weights may be quite different.In conclusion the author thanks Acting Member of the Academy of Sciences, USSR, A. I. Leipunskii for directing the work, Candidate of Physical-Mathematical Sciences, A. N. Serbinov who operated the high-voltage apparatus, and V. S. Stavinskii for discussing the results.     

Inelastic Scattering of 14MeV Neutrons by Medium Weight Nuclei

Continuous spectra of neutrons inelastically scattered from Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ge and As have been measured at an incident energy of 14 MeV with good statistics. Time-of-flight technique was used. In the analyses there were used three sets of inverse-reaction cross section σ_i: (1) calculated with Perey-Buck potential, (2) with Bjorklund-Fernbach potential and (3) cross section assumed to be constant. The inverse-reaction cross section does not affect appreciably the values of nuclear temperature or nuclear level density parameter for these nuclei except Ti, V and Cr. The nuclear temperature was found to be nearly constant in the mass number region of 48–75. The level density parameter as a function of mass number follows the general trend suggested by Newton. Newton's coefficient is found to be 0.092, in agreement with values obtained in other experiments. The total inelastic scattering cross sections are derived.     

The moderation length of neutrons

The neutron moderation length is an important constant without which is is impossible to deal with the design of nuclear reactors in all their aspects. A knowledge of the moderation lengths is especially necessary for the determination of the space distribution of the neutrons in a reactor, and for the calculation of the energy spectrum of the moderated neutrons.In the present work is given an approximate solution of the integral equations satisfied by the space moments of the neutron distribution function inan infinite medium with an infinite, plane, isotropic source. The energy-angle moments of the neutron scattering function are expressed in terms of experimentally determined angular distributions of neutrons of various energies in the case of anisotropic elastic scattering from nuclei. By using experimental results for the total cross section and for the angular distribution in the case of elastic scattering from the nuclei H¹, D², Be⁹, C¹², O¹⁶ the neutron moderation lengths were calculated for the moderators: water, heavy water, graphite, beryllium, and beryllium oxide.The results of the calculation were found to be in satisfactory agreement with experimental values.In conclusion the authors wish to thank Doctor of Physical and Mathematical Science G. I. Marchuk for useful discussions, and also B. S. Gudkov, Z. P. Drobyshev, and Z. I. Shemetenko for carrying out the calculations.     

Measurement of neutron spectra in the thermal column of the reactor of a power station

Knowledge of neutron spectra In nuclear reactors allows comparison of various theories of the slowing down of neutrons with experiment, and also allows carrying out reactor calculations which are based on actual neutron distributions therein. In this paper is described a neutron intensity monochtomator Intended for the measurement of neutron spectra in the energy interval 0 to 0.5 ev.Results are given for measurements for neutron spectra in the thermal column of the reactor of an atomic power station. Discontinuities in the neutron flux were discovered at neutron velocities of 600, 1000 and 1650 m/sec; an analysis is given of the causes of discontinuities of the neutron flux; an evaluation is made of the inelastic scattering cross section for neutrons in graphite. By the method of least squares, the temperature of the neutron gas was found, being equal to 354 ° K at a graphite temperature of 304 ° K.In conclusion we consider it our duty to express gratitude to A. K. Krasin and B. G. Dubovskii for interest and help in the work and F. L. shapiro for valuable interpretation of previous results.     

Calculation of the passage of fast neutrons through graphite

Using the Monte-Carlo method, the author calculates the results of passing neutrons from a plane unidirectional source, with E₀=3.3 or E₀=8.0 MeV, through graphite. The angle of incidence is taken as 0°. The graphite layer thicknesses considered are from 0.9 to 6 times the free-path length. The author calculates the dose, energy and numerical albedos of graphite and also the angular and energetic distributions of the reflected neutrons and the angular distribution of the energy of scattered neutrons. He plots the mean cosine of the angle of scattering vs. the albedo, and also vs. the value of the energy angulardistribution constant. The data obtained may be found useful in the design of shadow and labyrinth shielding.Translated from Atomnaya Énergiya, Vol. 22, No. 2, pp. 97–100, February, 1967.     

Evaluation of Neutron Nuclear Data for Uranium-233

Neutron nuclear data of ²³³U have been evaluated in the energy range from 10-⁵ eV to 20 MeV. Evaluated quantities are the total, fission, capture, elastic and inelastic scattering, (n,2n) and (n,3n) reaction cross sections, and the average numbers of prompt and delayed neutrons emitted per fission. The thermal and resonance cross sections have been evaluated on the basis of the measured data. The resolved resonance parameters are given up to 100 eV and the unresolved resonance parameters between 100 eV and 30keV. The total and fission cross sections have been evaluated in the higher energy region on the basis of the recently measured data, while the theoretical calculation with the optical, statistical and evaporation models has been used for evaluation of the other cross sections. The presently adopted optical potential parameters have reproduced well the experimental total cross section in the entire energy range as well as the measured data of the s-wave strength function. The structure observed in the vp values below 1 MeV is reproduced by the semi-empirical formula based on the fission fragment kinematics. The presently evaluated fission cross section is considerably lower than that of ENDF/B-IV between 10 and 50keV. This low fission cross section is expected to resolve the ^Keff discrepancy pointed out from the benchmark tests in ²³³U critical assemblies.     

各向异性散射截面对快堆敏感性系数计算影响研究

快能谱反应堆由于中子能量较高,中子各向异性散射会对计算结果有重要影响。本文在计算弹性散射和非弹性散射截面敏感性系数时,研究了高阶散射截面扰动对弹性散射和非弹性散射截面敏感性系数计算的影响。从理论上分析了隐式敏感性产生的原因和相关近似条件,采用直接扰动方法计算了ZPR-6/7快能谱反应堆主要核素的主要反应道的敏感性系数。研究结果表明,对于ZPR-6/7快能谱反应堆,不扰动238U高阶散射截面,总的弹性散射截面的敏感性系数比考虑高阶散射截面时的敏感性系数高44.3%,不考虑56Fe高阶非弹性散射截面的扰动,会造成非弹性散射截面敏感性系数偏高28.9%,而对其他核素的弹性散射和非弹性散射的敏感性系数影响较小。考虑到高阶散射截面后,自主开发的程序SUFR计算的总的敏感性系数结果与国际同类程序ERANOS和MCNP的计算结果吻合很好,最大偏差不超过3.22%,同时238U的弹性散射反应道和56Fe的非弹性散射反应道对有效增殖因子不确定度分析的精度也有了很大提高。因此,快堆敏感性系数计算需要考虑高阶散射截面影响,同时敏感性和不确定度分析程序SUFR开发正确,针对于快能谱反应堆进行敏感性系数的技术路线可行,计算精度同国际同类程序的计算精度相当。      

Neutron-capture therapy with ultra-cold neutrons

The status of neutron-capture therapy of malignant tumors and its problems – damage to healthy tissue as a result of neutron transport to the irradiation location and presence in the therapeutic beam of a background consisting of γ rays and fast neutrons – are presented. To solve these problems, the authors have proposed using ultracold neutrons with energy less than 10^–7 eV, whose unique capability is to undergo total reflection from the surface of a condensed substance at any angle of incidence. Numerous works have demonstrated that such neutrons can be transported along neutron guides. The cross section for inelastic scattering of neutrons by hydrogen-containing substances – water, ethyl alcohol, and biological tissue – has been measured in an IR-8 beam of ultracold and very cold neutrons. At temperature 200–300 K, the experimental data are in very good agreement with calculations, but as temperature decreases further a discrepancy appears, which could be due to the inaccuracy of the model spectra of the oscillations hydrogen-containing substances used in the calculations. The use of ultracold neutrons opens up new possibilities of neutron-capture therapy for treating malignant tumors localized in body cavities or organs.     

Theoretical estimates of cross sections for neutron–nucleus collisions

We construct an analytical model derived from nuclear reaction theory and having a simple functional form to demonstrate the quantitative agreement with the measured cross sections for neutron induced reactions. The neutron–nucleus total, reaction and scattering cross sections, for energies ranging from 5 to 700 MeV and for several nuclei spanning a wide mass range are estimated. Systematics of neutron scattering cross sections on various materials for neutron energies up to several hundred MeV are important for ADSS applications. The reaction cross sections of neutrons are useful for determining the neutron induced fission yields in actinides and pre-actinides. The present model based on nuclear reaction theory provides good estimates of the total cross section for neutron induced reaction.     

Neutron Nuclear Data Evaluation of Cesium Isotopes for JENDL-4.0

For the development of JENDL-4.0, neutron nuclear data for fission product nuclides, ^{133,134,135,136,137}Cs, were revised in the incident neutron energy range from 1 eV to 20MeV by using a coupled-channels optical model (OM), and nuclear reaction models. The OM potential parameters were determined for stable ¹³³Cs to reproduce the experimental data of total and elastic scattering cross sections and angular distributions of elastically scattered neutrons. The present results reasonably reproduce measured data for (n; 2n), (n; p), (n; α), and capture reactions on ¹³³Cs. Important differences between the present results and JENDL-3.3 are found for the capture cross sections of ^134,137Cs. The cross section obtained for ¹³⁷Cs was smaller than that in JENDL-3.3. This result makes the transmutation of medium-lived ¹³⁷Cs increasingly difficult. The production probabilities of metastable states for ^134,138Cs via capture reactions on ^133,137Cs are compared with experimental values. The present result for ^134m Cs production is marginally consistent with measured data. However, a large discrepancy is recognized for ^138m Cs production. The γ-ray emission data were evaluated with available measurements, and newly compiled in JENDL-4.0. Maxwellian-averaged capture cross sections were calculated in the energy range from 1 to 103 keV, and are compared with other derived data.