Measurement and Analysis of Neutron Spectra in Lithium Fluoride and Polytetrafluoroethylene Piles

For the assessment of neutron cross section data for fluorine, angular neutron spectra in the lithium fluoride (LiF) and polytetrafluoroethylene ((CF₂)n) piles were measured in the energy range from a few keV to a few MeV by the time-of-flight method with an electron linac, and the results were compared with those calculated by using nuclear data from JENDL-2 and ENDF/B-IV. Spatial distributions of neutron and X-ray fluxes were also measured in the test piles by the activation method, and the influence of photoneutrons generated in the sample material on the neutron spectrum in each pile was estimated. As a result, it was found that their influence on the neutron spectrum shape below 1 MeV was not so large as was necessary to be taken into account for the present assessment.

The calculated spectra using the JENDL-2 data and the ENDF/B-IV data show generally good agreement with those measured in both piles. However, both calculations underestimate the neutron fluxes around several 100 keV, and overestimate those below 100 keV, when they are normalized in the energy range of 10 keV～1 MeV. Large discrepancies are found between the shapes of the measured and calculated spectra around the resonances of fluorine cross section below 100 keV. The present measurements and analyses suggest that the reevaluations of the inelastic and elastic scattering cross sections below 1MeV and the resonance cross sections below 100 keV are necessary to reduce the observed discrepancies.     

Spectrum-Averaged One-Group Cross Sections of Actinides Based on JENDL-3

When uranium vapor is generated with an electron beam evaporator, a uranium plasma is formed on the evaporating surface. This plasma rises and expands with the vapor. Propagation behavior of this plasma was investigated by measuring plasma parameters, drift energy of ions and vapor flux along the propagation path. Over the range of 20-50 cm from the evaporation surface, the plasma density decreased from 3 × 10⁹ cm^?3 to 3 × 10⁸ cm^?3, while the electron temperature had a constant value of 0.29 eV. When the space potential was lowered from 1.48 to 0.80 V, the plasma ions were accelerated to increase the drift energy from 1.50 to 2.14 eV. Validity of the Boltzmann electron distribution was checked by comparing the space potential distribution with the plasma density distribution, and also the floating potential distribution with the ion flux distribution. These results confirm that the ambipolar diffusion governs the plasma propagation behavior. The change in the plasma density during its propagation occurred not only by an increase of plasma volume, but by the ion acceleration toward the propagation direction as well.     

Measurement of Neutron Total Cross Section of Silicon at 146-and 53.5-keV Windows

The process constituting the elements of the method for treating high-level liquid waste with addition of reducing agent at temperatures above 1,873 K were examined applying thermogravimetry and X-ray diffractometry to sample mixtures of the platinum metal oxides RuO₂, Rh₂O₃, PdO and of Re₂O₇, with and without addition of TiN as reducing agent. The addition of TiN reducing agent proved to induce reduction of the platinum metal oxides at temperatures far below those of thermal decomposition occurring in the absence of reducing agent. The presence of reducing agent further proved to reduce Re₂O₇ to ReO₃ at 373K, to ReO₂ at 773 K and to Re metal at 1,073 K. At around 773 K disproportionation reconverted part of the ReO₃ formed at 373K to ReO₂ and Re₂O₇. Thus, sublimation of the Re₂O₇ in starting material which occurs at 523 K in the absence of TiN was eliminated, and occurred only at 773 K on the Re₂O₇ that was regenerated at that temperature by disproportionation of the converted ReO₃. Alloys of Ru, Rh and Pd resulting from the above treatment with TiN proved to agree with what is indicated from phase diagrams. Among the alternative compounds existing in the Rh-Ti and Rh-Ti binary systems, however, solely Ru-Ti proved to have been produced.     

Equivalence of Taylor Series Approximation with Transcendental Equation of Neutron Slowing Down Theory under Constant Cross Sections

Abstract

Neutron moderation in an infinite homogeneous medium with constant cross sections, which is traditionally treated by Taylor series approximation methods, bears a close correspondence with the exact transcendental equation approach, as has been demonstrated by reducing the exact result to approximate forms. In this paper we give an alternative approach in which the convergence of the approximations to the exact result is derived by allowing the order of approximation to increase indefinitely. The approach is based on the fact that the usual approximations do not give a self consistent definition of the effective mass parameter. The present result can hence be considered as the slowing down counterpart of Davison's treatment of the one speed equation in the P_l approximation.     

Radiochemical Determination of Neutron Capture Cross Sections of 241Am

The thermal neutron capture cross sections and the neutron capture resonance integrals of ²⁴¹Am leading to the production of the isomer ²⁴²Am and the ground-state ^242gAm were measured radiochemically by the Cd-ratio technique with neutron flux monitors of Co/Al and Au/Al alloy. Highly-purified ²⁴¹Am targets were irradiated in an aluminum capsule by using JMTR. The neutron fluxes and their epithermal neutron fractions were determined by measuring γ-rays of ⁶⁰Co and ¹⁹⁸Au. The yields of ^242mAm and ^242gAm were decided by analyzing growth and decay curves of the α-ray activity ratios ²⁴²Cm/²⁴¹Am. The resultant thermal neutron capture cross sections are 85.7 ± 6.3 b and 768 ± 58 b for ^242mAm and ^242gAm, and the resonance integrals 114±7 b and 1,694±146 b, respectively. The differences between the present results and the evaluated values by Mughabghab are 38–59%. The isomeric ratios, g/(m+g), of 0.90±0.09 for thermal neutrons and 0.94±0.11 for epithermal neutrons are, however, almost consistent with evaluated values.     

Average Neutron Capture Cross Sections of 151Eu and 153Eu from 3 to 100 keV

Neutron capture cross sections of europium isotopes ¹⁵¹Eu and ¹⁵³Eu were measured in the neutron energy range of 3~100keV. Experiments were carried out with the time-of-flight facility at the 52 m station of the JAERI Electron Linear Accelerator. Prompt capture γ-rays were detected by a large liquid scintillation detector and the neutron flux shape was determined with a ⁶Li glass scintillation detector. The average capture cross sections were examined in terms of energy independent strength functions for ¹⁵¹Eu and ¹⁵³Eu.     

Measurement of Neutron Capture Cross Sections with Fe-Filtered Beam

The neutron capture cross sections of ⁹³Nb, ¹¹⁵In, ¹²⁷I, ¹⁶⁵Ho, ¹⁸¹Ta, ²³²Th and ²³⁸U were measured using the Fe-filtered beam. A 15-cm thick Fe filter was placed in the neutron beam produced by the KUR 46-MeV electron Linac and capture prays were detected by two C₆F₆ scintillation detectors located at an 11.7 m-flight path. The pulse-height weighting technique was used to determine the relative capture pray detection efficiency. The neutron flux was measured by the same detectors, whose detection efficiency for the 480-keV pray from the ¹⁰B(n, α₁) reaction was calibrated by the saturated resonance capture in Ag at 5.2-eV. Self-shielding and multiple scattering corrections were applied to the data. The results of 24-keV capture cross sections are 340, 770, 780, 1,280, 880, 520 and 520 mb for ⁹³Nb, ¹¹⁵In, ¹²⁷I, ¹⁶⁵Ho, ¹⁸¹Ta, ²³²Th and ²³⁸U, respectively. Total errors are 5 to 8%, with an estimated systematic error of 4%. The discrepancy between the present results and other data measured recently is within 10%.     

R-Matrix Analysis of 239Pu Neutron Transmissions and Fission Cross Sections in Energy Range from 1.0 keV to 2.5 keV

Abstract

The results of the ²³⁹Pu high resolution neutron transmission measurements of Harvey et al. and of the ²³⁹Pu high resolution fission cross section measurements of Weston & Todd performed at the Oak Ridge Electron Linear Accelerator (ORELA) were analysed in the energy range from 1 keV to 2. 5 keV by the Bayesian code SAMMY using the Reich-Moore approximation of the R-matrix theory. The results obtained in a previous analysis in the energy range from thermal to 1 keV were updated by taking into account the recent renormalization of the experimental fission data by Weston & Todd. The statistical properties of the parameters of the resonances identified in the energy range from thermal to 2. 5 keV were examined and improved values of the average parameters were obtained. The resonance parameters are given in an ENDF-6 format file available from JAERI Nuclear Data Center and from NEA Data Bank (OECD).     

Proton-Recoil Counter Technique for Measurement of Fast Neutron Spectrum

Fast neutron spectra were measured at the core centers of two plutonium-fuelled assemblies of FCA by the proton-recoil counter technique. Small cylindrical counters with field tubes made by metallizing a ceramic-seal surface were constructed. The filling gases, hydrogen and methane, were purified to reduce the concentration of electronegative impurites. The highest resolutions measured by ³He(n, p,^3T reaction were 3.8 and 5.5% (FWHM) for the hydrogen-filled and methane-filled counters respectively. Gamma-induced background was rejected by the pulse-shape discrimination technique, using a digital processing arithmetic unit and a two-dimensional pluse height analyzer. The spectrum measurement over a wide energy range was carried out by different settings of the main amplifier gain and of the gas multiplication of the counters; its energy ranged from 2.5 keV to 2 MeV.

Measurements were made with the energy resolution about 10% (FWHW) or less at energies down to 5 keV. The measured neutron spectra are compared with calculated spectra, obtained by the cell program SP-2000 with the fine group cross section library AGLI. The comparisons indicate that the measured spectra are in good agreement with the calculated spectra except near the large iron and oxygen resonances.     

Application of BGO Scintillators to Absolute Measurement of Neutron Capture Cross Sections between 0.01 eV and 10 eV

Applying a total energy absorption γ-ray detector composed of 12 bricks (5 × 5 cm², 7.5 cm thick) of BGO scintillators, the absolute measurement of capture cross sections for Au and Sb has been made in an energy region between 0.01 and 10eV using the linac time-of-flight method. Incident thermal neutron flux was absolutely determined by using the BGO detection system with a Sm sample. To extend the neutron flux measurement from the thermal neutron region to higher neutron energies, the ¹⁰B(n, αγ) reaction was applied. Absolute capture yield for the relevant capture sample was obtained by the saturated capture yield at a large resonance of the sample.

Gold was selected to investigate the application of the BGO detection system to the absolute measurement of the capture cross sections, since the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section is a well known standard one. The result of the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section showed good agreement with the evaluated data in JENDL Dosimetry File and ENDF/B-VI. Then, the detection system was applied to the Sb(n, γ) cross section measurement. Antimony has a large scattering-to-capture cross section ratio comparing to that of gold. The result showed good agreement with the evaluated data in JENDL-3.2 and ENDF/B-VI.     

Calculation of Neutron-Induced Reaction Cross Sections of Manganese-55

Neutron-induced reaction cross sections of ⁵⁵Mn are calculated for the evaluated nuclear data libraries, ENDF/B-VI and JENDL-3. Simultaneously calculated are the inelastic scattering, (n,2n), (n,p), (n,α), (n,np), (n,nα) and (n,γ) reaction cross sections, the angular distributions of emitted neutrons and protons, and the energy distributions of emitted particles and γ-rays. A unified Hauser-Feshbach code is applied to calculate these quantities. Precompound and direct-reaction processes are taken into consideration, in addition to the compound process. The calculated results reproduce the experimental data very well. Forward-peaked angular distributions of continuous neutrons are obtained from the calculation, and found to be consistent with the measurements at 14MeV.     

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.     

Measurements of Double-Differential Neutron Emission Cross Sections of 6Li, 7Li and 9Be for 11.5 MeV and 18.0 MeV Neutrons

Double-differential neutron emission cross sections (DDXs) of ⁶Li, ⁷Li and ⁹Be were measured for 18.0 MeV and 11.5 MeV incident neutrons produced by the T(d, n) and ¹⁵N(d, n) reactions respectively, using the Tohoku University Dynamitron time-of-flight (TOF) spectrometer. The data were obtained at 13 laboratory angles, and angular-differential cross sections (ADXs) of elastic and inelastic scattering neutrons were derived from the DDXs. For 11.5 MeV neutrons, we obtained the neutron emission spectra over the secondary neutron energies by newly employing the double TOF method as well as the conventional one. In the measurements at 18.0 MeV, we achieved better energy resolution than in our previous studies by using a neutron detector that has a larger solid angle and a thinner tritium target. The experimental results of DDXs and ADXs were compared with our previous results and the evaluated data given in JENDL-3.2, JENDL Fusion File and ENDF/B-VI. It is found that the JENDL data reproduce the experimental ones very well.     

Examination of Various Kinds of Systematics of Double-Differential Particle Emission Cross Sections for Medium-Heavy Nuclei Important to Fusion Neutronics

Various kinds of systematics used to calculate the double-differential light particle emission cross sections from nuclear reactions induced by light particles are examined for medium-heavy nuclei important in fusion neutronics applications. Fixing the incident and outgoing particles to neutrons, and the incident energy at 14 MeV, results calculated by the systematics, supplemented by the statistical model calculation, are compared with experimental data measured by two Japanese groups. It is concluded that systematics derived by Kumabe et at. and Kalbach has good accuracy in reproducing these data. Discrepancies in the experimental data are pointed out, and suggestions to future compilation of a special purpose file of JENDL for fusion neutronics are made.     

Measurement of Fast Neutron Induced Fission Cross Section Ratios of Pu-240 and Pu-242 Relative to U-235

Fission cross section ratios of ²⁴⁰Pu and ²⁴²Pu relative to ²³⁵U were measured by using the 4.5 MV Dynamitron accelerator of Tohoku University. The measurement using mono-energetic neutrons was performed in the neutron energy range of 0.6–7 MeV with the time-of-flight method. Prior to the measurement, a fast timing back-to-back fission chamber was developed with good time resolution to reduce the backgrounds due to α-particles and spontaneous fissions. Furthermore, we took account of the effect of the nonuniformity of fission sample thickness for accurate determination of fission cross section ratio. The uncertainty was estimated by analyzing the correlation between the error sources. The correlation matrix between the measured data was given. The overall uncertainty of the present results is about 2%. For both nuclides, the present results agree well with those by Meadows and by Kuprijanov et al. The JENDL-3 evaluation generally has good agreement with the present results. However, the evaluated data are slightly higher around 1 MeV and lower above 6 MeV than the present results.     

Calculation of Neutron Albedos for Slab with Invariant Imbedding Method

The albedo data are important for use in solving the radiation streaming problem with albedo techniques, such as albedo Monte Carlo and albedo-Sn methods. This paper describes a method for calculating the energy-angle dependent doubly differential albedos for slab geomery with one-dimensional transport theory, based on the invariant imbedding method as well as the Sn method. Neutron albedo data calculated by the invariant imbedding method, are compared with those calculated by the Sn method and with the experimental data. It is found that the invariant imbedding method can be used to calculate the albedo data for a semi-infinitely thick slab several dozens of times faster than when using the Sn method. The calculated results have excellent agreement with the measured values.     

Measurement of Leakage Neutron Spectra from a Spherical Pile of Zirconium Irradiated with 14MeV Neutrons and Validation of Its Nuclear Data

In order to make a benchmark validation of the nuclear data for Zr, the leakage neutron spectrum from a Zr sphere of a 61-cm diameter was measured between 0.1 and 16MeV using a time-of-flight technique with a 14MeV neutron source facility, OKTAVIAN. The result was compared with the calculation using the Monte Carlo code MCNP-4A. To investigate the spectrum dependence on the individual neutron reactions, test calculations were carried out with the MCNP-4A code using the JENDL-3.2-based libraries, in which partial cross section values were reduced from the original values. From the comparison between the measured and the calculated spectra, it was found that each of the results could predict well the experiment in general. However, in detail, both ENDF/B-VI and EFF-2.4 gave considerable overestimation above 1 MeV. The JENDL-3.2 predicts the spectrum almost satisfactorily except below 0.8 MeV and around 10 MeV. The discrepancy found in JENDL-3.2 calculation is considered due to the cross section values of the (n, 2n) reaction and its secondary energy distributions (SED). The modified JENDL-3.2 library with the reduced (n, 2n) reaction values and the lower SED below 1 MeV reproduced the experiment with better agreement over the whole energy range.     

Measurement of Gamma-Ray Production Cross Sections of Iron for Incident Neutron Energies between 6 and 33 MeV

Measurement of differential γ-ray production cross sections, i.e. (n, x γ) cross sections, of Fe was made for neutron energies from 6 to 33 MeV. Neutrons used in the experiment were white neutrons produced with (p, n) reactions by 35 MeV protons using a thick Be target. The neutron energy was analyzed by the time-of-flight method and bunched into 3 MeV wide energy bins, for each of which the spectrum of secondary γ-rays produced in an Fe sample was measured by a BGO scintillator at an angle of 144° to the neutron beam direction.

The obtained (n, xγ) cross sections agreed well with other data and the evaluated data file of ENDF/B-IV at neutron energies below 15 MeV where data were existing. The JENDL-3 file overestimated the γ-ray spectra at γ-ray energies of 3 to 7 MeV. The present work newly provided the data in the neutron energy range above 20 MeV. The GNASH calculation made by Young reproduced the measured data fairly well even at these higher energies.