Measurement of Fast Neutron Induced Fission Cross Sections of Americium-243 Relative to Uranium-235

The fission cross section ratio of ²⁴³Am to ²³⁵U has been measured in the energy range of 1.1～6.8 MeV with monoenergetic neutrons. An ionization fission chamber was used to detect fission events. The quantitative analyses of the fission samples were made with a low geometry counter and a 2 π counter. Uncertainties of the measured data were analyzed considering correlations between error elements. The present result is very close to that of Fursov et al. and lower by about 20% than the values reported by Behrens & Browne.     

Measurement and Analysis of241 Am Fission Rate Ratio Relative to235 U Fission Rate in Thermal Neutron Systems Using Kyoto University Critical Assembly

Integral measurements of²⁴¹ Am fission rate ratio relative to²³⁵ U fission rate were performed at Kyoto University Critical Assembly. The fission rates were measured using the back-to-back type double fission chamber at five thermal cores with different H/²³⁵ U ratio so that the neutron spectra of the cores were systematically varied. The measured fission rate ratios, normalized to number of atoms, were 0. 0144 to 0. 0522, with a typical uncertainty of 2%. The measured data were compared with the calculated results using MVP based on JENDL-3.2, which gave the averaged calculated-to-experimental ratio (C/E) of 0.88. Obtained results of C/E using ²⁴¹Am fission cross sections from JENDL-3/2, ENDF/B-VI and JEF2.2 showed that the latter two gave larger C/E values than those by JENDL-3.2 by about 2% and 7 to 9%, respectively. It has been found that the large difference between JENDL-3.2 and JEF2.2 arises mainly from the significant cross section difference at the vicinity of resonance at 0.576 eV, whereas the difference of thermal cross sections, especially in the range of 0.01 eV to 0.2 eV also has significant contribution for well-thermalized cores.     

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.     

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.     

Measurement and Analysis of Neutron Spectrum in Spherical Pile of Thoria

The energy spectrum of neutrons in a pile of thoria was measured from several keV to several MeV by the time-of-flight method with an electron linear accelerator and the results were compared with the theoretical ones predicted with three group constants for thorium, (1) constants of Abagyan et al.'s (ABBN), (2) newly produced constants of J AERI-FAST type (NNM) and (3) constants produced from ENDF/B-H in ORNL (DLC-2D).

General agreement between measurements and calculations was seen for all cases. However, (1) the predicted spectra with ABBN and DLC-2D gave less flux by about 30% than the measured in the energy below about 100 keV, (2) similar disagreement was observed only above 10 keV for the spectrum with NNM and (3) the measured spectrum became about 20% lower than all of the predicted above 1 MeV.

In parallel with the spectrum measurement, neutron spatial distribution was measured by the activation method and the results showed spherical symmetry around a photoneutron target at the center of the pile. From this fact adequancy of using one-dimensional transport code (ANISN or DTF-IV) with spherical symmetry for analysis could be verified.     

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).     

Measurement and Analysis of Fission Rate Distributions of 237Np and 238U in a Polyethylene System with a 65 MeV Neutron Source

Fission rates of ²³⁷Np and ²³⁸U in a polyethylene (CH₂) system were measured with a 65MeV quasi monoenergetic neutron source. Relative fission rate distributions dependent on polyethylene thickness up to about 70 cm were obtained for both nuclides with the experimental error within 7%. The present experiment was analyzed by the NMTC/JAERI code that has been employed for designing accelerator-driven transmutation systems. The fission rates of both ²³⁷Np and ²³⁸U calculated by the NMTC/JAERI did not agree with the experimental ones. The C/E values for both were about 2.0 at 71.8cm of polyethylene thickness when both experimental and calculated values were normalized to 1.0 at 0.0 cm of polyethylene thickness. A sensitivity analysis of the NMTC/JAERI was performed by changing cross sections and angular distributions of hydrogen and carbon and by employing three options of the intra-nuclear cascade/evaporation calculation of the NMTC/JAERI. The disagreement of the NMTC/JAERI calculation with the experimental values was partially improved by increasing the nonelastic-scattering cross section of carbon and by broadening the elastic-scattering angular distribution of carbon.     

Measurement of the 93Nb(n,n')93mNb Reaction Cross Section at 14.5 and 14.9MeV

Abstract

Cross sections at 14.5 and 14.9 MeV for the low threshold energy reaction of ⁹³Nb(n, n')^93mNb were measured by the activation method at the FNS (Fusion Neutronics Source) of JAERI. Although there are multiple experimental data for this reaction cross section from the threshold to 9 MeV, only one data at 14.3 MeV has been reported by Ryves et at. The correction for low energy neutron contribution was found very sensitive on the reaction rate determination. Careful treatments were performed to arrive at final error evaluation considering neutron spectrum calculated with a Monte Carlo Code and cross section curves available. The cross sections measured in the present work are larger by more than 30% than those in both IRDF-90 and JENDL Dosimetry File, which are based on the data of Ryves et al. On the other hand, the present data are lower by 10-15% than the evaluation by Odano et al. It is highly recommended to re-evaluate the cross section by taking the present data into account.     

Neutron Emission,Mass Distribution and Energetics in 14.5 MeV-Neutron Induced Fission of Uranium-238

A study is performed on 14.5 MeV-neutron induced fission of ²³⁸U by means of three-parameter experiment in which the energies of both fragments and the time-of-flight of one fragment are measured. A mosaic-arrayed surface barrier detector of large sensitive area is used at the remote end of a flight tube. The pre- and post-neutron-emission fragment mass distributions are obtained, together with the average total kinetic energy of fragment as a function of its mass. The average number of neutrons emitted from an individual fragment and the average total number of emitted neutrons are also derived as a function of fragment mass. The results agree well with those calculated by the method developed in our laboratory for medium-excitation fission. The average number of emitted neutrons and the mass distribution of fission fragment are derived for the respective reactions of first-, second- and third-chance fission.     

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.     

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.