Neutron Resonance Parameters of Silver-107 and Silver-109

Neutron transmission measurements were carried out on the separated isotopes of silver using the time-of-flight facility at the Japan Atomic Energy Research Institute electron linear accelerator. Neutrons were detected with the ⁶Li-glass detectors at 56 and 191 m. The samples used were metallic powder enriched to 98.2% for ¹⁰⁷Ag and 99.3% for ¹⁰⁹Ag. Transmission data were analyzed with the multi-level Breit-Wigner formula incorporated in a least squares fitting program. Resonance energies and neutron widths were determined for the large number of resolved resonances in the neutron energy region of 400 eV~7 keV. The s-wave strength functions and average level spacings were obtained to be; S₀= (0.43±0.05) × 10^?4, D₀ = 20±2 eV for ¹⁰⁷Ag and S₀= (0.45 ± 0.05) × 10^?4, D₀ = 20 ± 2eV for ¹⁰⁹Ag.     

Steady-State Power Coefficients of Fast Reactor Considering the Effects of Partial Load and Flow Control

High resolution neutron transmission and low background capture measurements were carried out on the separated rubidium isotopes, using the time-of-flight facility of the linear accelerator of Japan Atomic Energy Research Institute. Resonance parameters and associated quantities were deduced as follows:

For ⁸⁵Rb, gΓ _n values were determined for 138 resonance levels in the energy region below 18.5 keV. s-wave strength function was obtained to be S₀=(0.94±0.11)×10^?4, average level spacing ≤D>=133±11 eV and average radiative width ≤Γ _r >=328±18 me V. For ⁸⁷Rb, gΓ _n values were determined for 30 resonance levels in the energy region below 48.6 keV and the following quantities were deduced:

S₀=(1.15±0.3)×10^?4, D=1,380±250 eV and ≤Γγ>=166±30 meV.

For ⁸⁵Rb average properties of resonances are in good agreement with the prediction of the statistical model. On the other hand, for ⁸⁷R9b the average properties of resonances deviate from the prediction of the statistical model; four strong s-wave resonances cluster within an energy interval of 5 keV, and they carry about 37% of s-wave strength below 48.6 keV.     

Design of an epi-thermal neutron flux intensity monitor with GaN wafer for boron neutron capture therapy

Boron neutron capture therapy (BNCT) is a promising cancer therapy. Epi-thermal neutron (0.5 eV < E_n < 10 keV) flux intensity is one of the basic characteristics for modern BNCT. In this work, based on the ⁷¹Ga(n,γ)⁷²Ga reaction, a new simple monitor with gallium nitride (GaN) wafer as activation material was designed by Monte Carlo simulations to precisely measure the absolute integral flux intensity of epi-thermal neutrons especially for practical BNCT. In the monitor, a GaN wafer was positioned in the center of a polyethylene sphere as neutron moderator covered with cadmium (Cd) layer as thermal neutron absorber outside. The simulation results and related analysis indicated that the epi-thermal neutron flux intensity could be precisely measured by the presently designed monitor.     

Monte Carlo optimization of an epithermal neutron flux monitor for BNCT

An epithermal neutron (0.5 eV < E_n < 10 keV) flux monitor developed for boron neutron capture therapy (BNCT) was optimized by Monte Carlo simulations. Based on this optimization study, the optimization results for each component of the epithermal neutron flux monitor were obtained. The simulation results indicated that the epithermal neutron flux monitor with optimal configuration was more efficiently applicable to precisely measure the epithermal neutron fluxes of BNCT neutron sources.     

Neutron capture cross section measurements of 151,153Eu using a pair of C6D6 detectors

We have measured the neutron capture cross sections of ¹⁵¹Eu and ¹⁵³Eu by the time-of-flight (TOF) method in the range from 0.005 eV to keV region using the Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC). We employed a pair of C₆D₆ liquid scintillators for the prompt capture γ-ray measurement. The pulse-height weighting technique was employed to obtain the capture yields from the γ-ray spectra of ^151,153Eu. The obtained thermal cross sections at 0.0253 eV are 9051 ± 683 b for ¹⁵¹Eu and 364 ± 44 b for ¹⁵³Eu, respectively. The resonance integrals have been derived as 3490 ± 162 b for ¹⁵¹Eu and 1538 ± 106 b for ¹⁵³Eu.

The obtained capture cross sections were compared with the previously reported experimental data and the evaluated data. The evaluated data in JENDL-4.0 and JEFF-3.2 show good agreement with the present experiment results of ¹⁵¹Eu, however, the evaluated data in ENDF/B-VII.1 are larger than the present experiment results of ¹⁵¹Eu about 10% to 20% in the energy region from 0.03 to 0.2 eV. For the neutron capture cross sections of ¹⁵³Eu, the evaluated data in ENDF/B-VII.1 and Widder's data are in good agreement with the present results in the energy region below 0.35 eV.     

Measurements of Energy Dependent Doppler Factor for Uranium Dioxide Using a Lead Slowing-Down Time Spectrometer

The energy dependent Doppler factor of resonance absorption in U0₂ was measured in the region of neutron energy between 1 eV and 10 keV using a U0₂ cylindrical sample and a lead slowing-down time spectrometer.

The Doppler factor is herein defined as the ratio of neutron capture rate in the sample at the high temperatures T = 450 and 300° C to that at the low temperature T = 20°C. The Doppler factor measured at T = 450 and 300°C showed the maximum values of 1.225± 0.044 and 1.186±0.040, respectively around the resonance of 66.3 eV. In the energy region above about 200 eV, the Doppler factor decreased with increasing neutron energy. In the energy region below about 70 eV except for region around the resonance of 6.68 eV, the Doppler factor decreased with increasing neutron energy.

The energy dependent Doppler factors were calculated by the collision probability method and the resonance parameters of ENDF/B-IL In the energy region of 1~454 eV, the calculation agreed with the experiment within the error of 5A% which was not unreasonable when one considered the experimental errors of 3.1~4.7%. In the energy region of 454 eV~3 keV, the calculation was systematically 1~6.5% larger than the experiment, though the experimental errors were 2~3.6% in this energy region.     

Neutron Capture Cross Section Measurement of Praseodymium in the Energy Region from 0.003 eV to 140 keV

The neutron capture cross section of praseodymium (¹⁴¹Pr) has been measured relative to the ¹⁰B(n,αγ) standard cross section in the energy region from 0.003 eV to 140 keV by the neutron time-of-flight (TOF) method with a 46-MeV electron linear accelerator (linac) of the Research Reactor Institute, Kyoto University (KURRI). An assembly of Bi₄Ge₃O₁₂ (BGO) scintillators was used for the capture cross section measurement. In addition, the thermal neutron cross section (2,200 m/s value) of the ¹⁴¹Pr(n, γ)¹⁴²Pr reaction has been also measured by an activation method at the heavy water thermal neutron facility of the Kyoto University Reactor (KUR). The thermal neutron flux was monitored with the ¹⁹⁷Au(n, γ)¹⁹⁸Au standard cross section. The above TOF measurement has been normalized to the current activation data (11.6±1.3 b) at 0.0253 eV.

The evaluated data in JENDL-3.3, ENDF/B-VI, and JEF-2.2 have been in general agreement with the current result, except that the JENDL-3.3 and the JEF-2.2 values are clearly lower than the measurement in the cross section minimum region from about 10 to 500 eV.     

Measurement of keV-neutron capture cross-sections for 164Dy

The neutron capture cross-sections of ¹⁶⁴Dy were measured in the neutron energy region of 10 to 90 keV using the 3-MV Pelletron accelerator of the Research Laboratory for Nuclear Reactors at the Tokyo Institute of Technology. Pulsed keV neutrons were produced from the ⁷Li(p,n)⁷Be reaction by bombarding a lithium target with the 1.5-ns bunched proton beam from the Pelletron accelerator. The incident neutron spectrum on a capture sample was measured by means of a TOF method with a ⁶Li-glass detector. Capture γ-rays were detected with a large anti-Compton NaI(Tl) spectrometer, employing a TOF method. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to derive capture yields. The capture cross-sections were obtained by using the standard capture cross-sections of ¹⁹⁷Au. The present results were compared with the previous measurements and the evaluated values of ENDF/B-VI.     

Design of the thermal neutron beam for neutron radiography at the Syrian MNSR

Neutron beam design was studied at the Syrian reactor (MNSR, 30 kW) with a view to generating thermal neutron beam in the vertical irradiation sites for neutron radiography. The design of the neutron collimator was performed using MCNP4C and the ENDF/B-V cross-section library. Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, >10 keV, respectively. To produce a good neutron beam quality, bismuth was used as photon filter. In this design, the L/D ratio of this facility had the value of 125. The thermal neutron flux at the beam exit was about 2.548 × 10⁵ n/cm² s. If such neutron beam were built into the Syrian MNSR many scientific applications would be available using the neutron radiography.     

Measurement of neutron capture cross-section of the Ga(n, γ)Ga reaction at 0.0536 eV energy

The neutron capture cross-section for the ⁷¹Ga(n,  γ)⁷²Ga reaction at 0.0536 eV energy was measured using activation technique based on TRIGA Mark-II research reactor. The ¹⁹⁷Au(n, γ)¹⁹⁸Au monitor reaction was used to determine the effective neutron flux. Neutron absorption and γ-ray attenuation in gallium oxide pellet were corrected in determination of cross-section. The cross-section for the above reaction at 0.0536 eV amounts to 2.75 ± 0.14 b. As far as we know there are no experimental data available at our investigated energy. So far we are the first, who carried out experiment with 0.0536 eV neutrons for cross-section measurement. The present result is larger than that of JENDL-3.3, but consistent within the uncertainty range. The value of ENDF/B-VII is higher than this work. The result of this work will be useful to observe energy dependence of neutron capture cross-sections.     

Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10¹⁷, 6.1 × 10¹⁶ (E < 0.55 eV), 7.6 × 10¹⁶ (E > 0.1 MeV) and 4.0 × 10¹⁶ (E > 1 MeV) m⁻² s⁻¹. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.     

Measurement of neutron capture cross-section of indium in the energy region from 0.003 eV to 30 keV

The neutron capture cross-section of indium (In) has been measured in the energy region from 0.003 eV to 30 keV by the neutron time-of-flight (TOF) method with a 46-MeV electron linear accelerator (linac) of the Research Reactor Institute, Kyoto University. An assembly of Bi₄Ge₃O₁₂ (BGO) scintillators, which was composed of 12 pieces of BGO and placed at a distance of 12.7±0.02 m from the neutron source, was employed as a total energy absorption detector for the prompt capture γ-ray measurement from the sample. In order to determine the neutron flux impinging on the capture sample, a plug of ¹⁰B powder sample and the ¹⁰B(n,α) standard cross-section were used. The data measured by Haddad et al. (Haddad, E., Friesenhahn, S., Lopez, W.M., 1963. Report of Gulf Energy and Environmental Systems, p. 3874) seem to be in good agreement with the present measurement. Popov et al. obtained the poor energy resolution data in the resonance region with a lead slowing-down spectrometer and the consistent data with the present above about 300 eV. The experimental data measured by Kononov et al. (Kononov, U.N., Jurlov, B.D., Poletaev, E.D., Timokhov, V.M., Manturov, G.N., 1977. Report of Obninsk, pp. 22–29) and Gibbons et al. (Gibbons, J.H., Macklin, R.L., Miller, P.D., Neiler, J.H., 1961. Phys. Rev. 122, 182) showed good agreement with the present values in the higher energy region. However, the data measured by Block et al. (Block R.C., Kaushal, N.N., Hockenbury, R.W., 1972. Conference on New Developments in Reactor Physics and Shielding at Kiamesha Lake, Vol. 2, p. 1107) seem to be a little higher than the present measurement above 800 eV. The evaluated data in ENDF/B-VI, JENDL-3.2, and JEF-2.2 have been in general agreement with the present result in the relevant energy region, although the JENDL-3.2 are higher than the measurement, the ENDF/B-VI and the JEF-2.2 from 2 to about 10 keV. Most of the previous experimental and the evaluated thermal neutron cross-sections are generally close to the present value of 199.6±5.6 b at 0.0253 eV.     

Measurement of keV-neutron capture cross-sections and capture γ-ray spectra of Fe and Fe

The neutron capture cross-sections and the capture γ-ray spectra of ⁵⁶Fe and ⁵⁷Fe have been measured in the neutron energy range from 10 to 90 keV. Pulsed keV-neutrons were produced from the ⁷Li(p,n)⁷Be reaction by bombarding a lithium target with a 1.5-ns bunched proton beam from a 3 MV Pelletron accelerator. The incident neutron spectrum on the capture sample was measured using a time-of-flight method with a ⁶Li-glass detector. The capture γ-rays emitted from an iron or standard gold sample were detected with a large anti-Compton NaI(Tl) spectrometer. The capture yield of the iron or gold sample was obtained by applying a pulse-height weighting technique to the corresponding capture γ-ray pulse-height spectrum. The capture cross-sections of ^56,57Fe were derived with errors less than 5% using the standard capture cross-sections of ¹⁹⁷Au. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. The present results for the capture cross-sections were compared with the previous measurements and the evaluated values of ENDF/B-VII.0 and JENDL-3.3. The Maxwellian-averaged capture cross-sections of ⁵⁶Fe and ⁵⁷Fe at 30 keV are derived as 12.22 ± 2.06 mb and 44.48 ± 7.56 mb, respectively.     

Neutron Resonance Parameters of Ba-135, Ba-137 and Ba-138

Neutron transmission measurements were carried out on the separated isotopes of Ba at the JAERI electron linac. Resonance energies and neutron widths were determined for a large number of resonances in the neutron energy range from 400 eV to 4.6 keV for ¹³⁵Ba, 15 keV for ¹³⁷Ba and 63 keV for ¹³⁸Ba. Many of these resonances were newly observed in this experiment. The s-wave strength functions obtained are S ₀= (1.33±0.22) x ^10-4 for ¹³⁵Ba, and S ₀= (0.51±0.12) x ^10-4 for ¹³⁷Ba. An apparent energy dependence of the strength function was observed for ¹³⁵Ba. New resonance parameters of ¹³⁸Ba were also obtained for several weak P-wave levels.     

Measurement of thermal neutron cross-section and resonance integral for 186W(n,γ)187W reaction by the activation method using a single monitor

The thermal neutron cross-section (σ₀) and the resonance integral (I₀) of the reaction ¹⁸⁶W(n,γ)¹⁸⁷W were measured by the activation method using ⁵⁵Mn as a single comparator. The diluted MnO₂ and WO₃ samples within and without a cylindrical Cd shield case were irradiated in an isotropic neutron field of the ²⁴¹Am–Be neutron source. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated high purity Ge detector. The necessary correction factors for gamma ray attenuation, thermal and resonance neutron self-shielding effects, and the shape factor (α) for epithermal neutron spectrum were taken into account in the determinations. The thermal neutron cross-section for ¹⁸⁶W(n,γ)¹⁸⁷W reaction has been determined to be 39.5±2.3 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross-section value of 13.3±0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. The present value of 39.5±2.3 b for ¹⁸⁶W(n,γ)¹⁸⁷W, in general is in good agreement with most of experimental data and evaluated ones in ENDF/B-VI and JENDL-3.2 within the limits of error. The resonance integral has also been measured relative to the reference value of 14.0±0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction using a 1/E^1+α epithermal neutron spectrum of the ²⁴¹Am–Be neutron source. By defining Cd cut-off energy 0.55 eV, the resonance integral obtained was 493±40 b. The existing experimental and evaluated data for the resonance integral are distributed from 290 to 534 b. The present resonance integral value agrees with some previously reported values.     

Measurement of thermal neutron cross section and resonance integral for 165Ho(n, γ)166gHo reaction by the activation method

The thermal neutron cross section and the resonance integral of the reaction ¹⁶⁵Ho(n, γ)^166gHo were measured by the activation method using ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction. The sufficiently diluted MnO₂ and Ho₂O₃ samples with and without a cylindrical Cd case were irradiated in an isotropic neutron field of the ²⁴¹Am–Be neutron sources. The γ-ray spectra from the irradiated samples were measured with a calibrated n-type high purity Ge detector. Thus, the thermal neutron cross section for ¹⁶⁵Ho(n,γ)^166gHo reaction has been determined to be 59.2 ± 2.5 b relative to the reference thermal neutron cross section value of 13.3 ± 0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction, and it generally agrees with the recent measurements within about 1 to 12%. The resonance integral has also been measured relative to the reference value of 14.0 ± 0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction using an epithermal neutron spectrum of the ²⁴¹Am–Be neutron source. The resonance integral for ¹⁶⁵Ho(n, γ)^166gHo reaction obtained was 667 ± 46 b at a cut-off energy of 0.55 eV for 1 mm Cd thickness. The existing experimental and evaluated data for the resonance integral are distributed from 618 to 752 b. The present resonance integral value agrees with most of the previously reported values obtained by ¹⁹⁷Au standard monitor within the limits of error.     

Design of moderator and multiplier systems for D–T neutron source in BNCT

Boron Neutron Capture Therapy (BNCT) is an outstanding way to treat Glioblastoma Multiforme. Epithermal neutrons with energy from 1 eV to 10 keV represent the most effective range for brain tumor therapy. In this research we have focused on ³H(d, n)⁴He reaction as a neutron source using Cock Craft Walton accelerator. High neutron yield with 14.1 MeV energy can be generated via accelerating a deuteron beam with 110 keV energy.A Monte Carlo simulation code (MCNP4C) was used to design the D–T source. Pb and ²³⁸U are suggested as neutron multipliers; AlF₃ and BeO as a moderator and reflector, respectively. An Al layer is used for decreasing the ratio of fast to total neutron fluxes. Epithermal neutron flux in the suggested system is 10⁸ n/cm² s and is a suitable flux for BNCT applications. Finally the suggested configuration is compared to the most recent works and it is shown that the proposed configuration works better.     

Resonance Parameters of Tantalum-181 in Neutron Energy Range from 100 to 4,300eV

Neutron transmission measurements were performed on natural tantalum (abundance ratio 99.988% for ¹⁸¹Ta) in the energy range of 100–4,300 eV using the Japan Atomic Energy Research Institute linac. The transmissions were measured using 55 and 190 m time-of-flight spectrometers for two and three samples of different thicknesses, respectively. These transmission data were simultaneously analyzed with a least squares fitting program based on a multl-level Breit-Wigner formula, and resonance energies and neutron width were obtained for 696 resonances of ¹⁸¹Ta.

The statistical analysis of these parameters gave the s-wave average level spacing of D=4.10±0.14 eV and s-wave neutron strength functions of (1.67±0.13) × 10^?4, (1.09 ± 0.09) × 10^?4 and (1.42 ± 0.20) × 10^?4 for the energy intervals from 100 ? 1,700 eV, 1,700–3,400 eV and 3,400–4,300 eV, respectively. This significant difference among the neutron strength function for each energy interval is a prominent result of the present experiments and is of great interest.     

Capture cross sections in construction materials for neutrons with energies up to 50 keV

The capture cross sections for neutrons with energies up to 50 keV in nickel, copper, molybdenum, and tungsten were measured with a lead slowing-down-time neutron spectrometer. Previously unreported resonances were detected in nickel (2.3 keV), Cu⁶³ (–100 eV), and Mo⁹³ (12 eV), and their parameters were estimated. The radiation widths of the copper resonances at 227 and 580 eV were 0.6 eV, Resonance absorption integrals are given for the materials studied.Translated from Atomnaya Énergiya, Vol. 15, No. 2, pp. 120–126, August, 1963