Thermal-Neutron Capture Cross Section and Resonance Integral of Americium-241

The thermal-neutron capture cross section (σ_0;g) and the resonance integral (I _0,g) leading to the ground state of 242Am were measured by an activation method for neutron capture by²⁴¹ Am. A method with gadolinium, which was similar to the cadmium difference method, was used to measure the cross section σ_0;g with attention to resonances of²⁴¹ Am. Americium chloride samples containing²⁴¹ Am radioisotope were irradiated for 68 h in the long-irradiation plug of the Kyoto University Research Reactor, KUR. Wires of Co/Al and Au/Al alloys were used as monitors to determine thermal-neutron fluxes and epithermal Westcott's indexes at the irradiation positions. An α-ray spectrometer was used to measure the activity ratios of²⁴² Cm to²⁴¹ Am. On the basis of Westcott's convention, the σ_0;g and I _0,g values were determined as 628 ± 22 b and 3:5 ± 0:3 kb, respectively.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

Measurements of neutron total and capture cross sections of 241Am with ANNRI at J-PARC

Neutron total and capture cross sections of ²⁴¹Am have been measured with a new data acquisition system and a new neutron transmission measurement system installed in Accurate Neutron Nucleus Reaction measurement Instrument at Materials and Life Science Experimental Facility of Japan Proton Accelerator Research Complex. The neutron total cross sections of ²⁴¹Am were determined by using a neutron time-of-flight (TOF) method in the neutron energy region from 4 meV to 2 eV. The thermal total cross section of ²⁴¹Am was derived with an uncertainty of 2.9%. A pulse-height weighting technique was applied to determine neutron capture yields of ²⁴¹Am. The neutron capture cross sections were determined by the TOF method in the neutron energy region from the thermal to 100 eV, and the thermal capture cross section was obtained with an uncertainty of 4.1%. The evaluation data of JENDL-4.0 and JEFF-3.2 were compared with the present results.     

Measurement of Effective Capture Cross Section of Americium-243 for Thermal Neutrons

The effective capture cross section of ²⁴³Am for thermal neutrons was measured with an activation method. A sample of ²⁴³Am was irradiated for 10 hrs at Kyoto University Reactor, KUR. After the irradiation, the sample was cooled for one month. In the cooling time, ^244mAm and ^244gAm produced by the irradiation decayed out to ²⁴⁴Cm. The α rays emitted from ²⁴³Am and ²⁴⁴Cm were measured with a silicon surface barrier detector. The neutron flux at the irradiation position was monitored using Au/Al and Co/Al wires. The effective capture cross section was deduced as 174.5±5.3b from the α-ray counts and the neutron flux. The quantity r√T/T₀ in Westcott's convention was 0.037±0.004. The present result was compared with the effective capture cross sections from the JENDL-3.3 and the Mughabghab evaluations.     

Measurement and Analysis of Reactivity Worth of 241Am Sample in Water-Moderated Low-Enriched UO2 Fuel Lattices at TCA

The reactivity worths of 22.82 grams of ²⁴¹Am oxide sample were measured and theoretically analyzed in water-moderated UO2 fuel lattices in seven cores of the Tank-Type Critical Assembly (TCA) at the Japan Atomic Energy Agency for an integral test of ²⁴¹Am nuclear data. These cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The sample reactivity worth was measured with an uncertainty of 2.1% or less. The theoretical analysis was performed using the JENDL-3.3 nuclear data by a Monte Carlo calculation method. Ratios of calculation to experiment (C/Es) of the reactivity worth were between 0.91 and 0.97, and showed no apparent dependence on the neutron spectrum. In addition, sensitivity analysis based on the deterministic calculation method was carried out to obtain the impact of changing the ²⁴¹Am capture cross section on the sample reactivity worth. The result of this analysis showed that the C/E could be significantly improved by almost uniformly increasing the ²⁴¹Am capture cross section of JENDL-3.3 by 25–30%.     

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 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.     

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.     

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.     

Neutron cross section sensitivity analysis of UO2 and MOX core physics experiments on light water reactors

The perturbation theory based on the transport calculation has been applied to study sensitivity of neutron multiplication factors (k_eff's) to neutron cross sections used for the reactivity analysis of UO₂ and MOX core physics experiments on light water reactors. The studied cross sections were neutron capture, fission and elastic scattering cross sections, and a number of fission neutrons, ν. The obtained sensitivities were multiplied to relative differences in the cross sections between JENDL-4.0 and JENDL-3.3 in order to estimate the reactivity effects. The results show that the increase in k_eff, 0.3%Δk/kk′, from JENDL-3.3 to JENDL-4.0 for the UO₂ core is mainly attributed to the decreases in the capture cross sections of ²³⁸U. On the other hand, there are various contributions from the differences in the cross sections of U, Pu, and Am isotopes for the MOX cores. The major contributions to increase in k_eff are decreases in the capture cross sections of ²³⁸U,²³⁸Pu, ²³⁹Pu, and those to decrease in k_eff are decreases in ν of ²³⁹Pu and increases in the capture cross sections of²⁴¹Am. They compensate each other, and the difference in k_eff between JENDL-3.3 and JENDL-4.0 is less than 0.1%Δk/kk′ and relatively small.     

Measurement of Thermal Neutron Capture Cross Section and Resonance Integral of the Reaction 133Cs(n, γ)134m,134gCs

The measurements of the thermal neutron (2,200 m/s neutron) cross section (σ₀) and the resonance integral (I ₀) of the ¹³³Cs(n, γ;) reaction were performed by an activation method to obtain fundamental data for research on the transmutation of nuclear wastes. The cross sections for the formations of the isomeric state ^134mCs and the ground state ^134mCs were measured respectively by following the behavior of the γ-ray counting rate after the irradiation.

The thermal neutron capture cross sections and the resonance integrals of the ¹³³Cs(n, γ) reaction were determined to be 2.70±0.13 b and 23.2±1.8 b for the formation of the isomeric state ^134mCs, and 26.3±1.0 b and 275±16 b for the formation of the ground state of ^134gCs. The results for the reaction ¹³³Cs(n, γ)^134m+gCs were 29.0±l.0 b and 298±16 b, respectively. As for the thermal neutron capture cross section for the formation of ^134m+gCs, the evaluated value (29.00 b) of JENDL-3.2 agreed with the present result. The reported value by Baerg et al. was in good agreement with the present result within the limits of error on the thermal neutron capture cross section for ^134mCs. On the other hand, the resonance integral for ^134m+g Cs was 32% smaller than the experimental value by Steinnes et al.     

<Superscript>243</Superscript>Am neutron fission cross section and resonance parameters

The SVZ-100 lead moderation time neutron spectrometer at the Institute of Nuclear Physics of the Russian Academy of Sciences was used to measure the fission cross section for ²⁴³Am in the neutron energy range E_n = 0.3 eV – 10 keV. The resonance fission integrals and the area and fission width of the resolved resonances were calculated. The properties of the intermediate-structure resonances were evaluated. The results were compared with existing data and recommended evaluations.     

Slow Neutron Resonances in Terbium-159

An experiment of neutron resonances in ¹⁵⁹Tb was carried out using the Japan Atomic Energy Research Institute linac time-of-flight facility. Transmission and capture measurements were made on terbium samples of two thicknesses, using ⁶Li-glass and Moxon Rae detectors at the 47 m station; the neutron flux was monitored with a ⁶Li-glass transmission type flux monitor. Transmission data were analyzed with an area program up to 1.2 keV, and capture data with Monte-Carlo program CAFIT, to obtain 2gΓ⁰ _n Γ and Γ_γ Resonance parameters of 209 levels below 1.2 keV are obtained, and 52 levels between 754eV and 1.2keV are new ones. The results are; average level spacing <D>=4.4±0.4eV below 600 eV, s-wave strength function S₀=(1.55 ±0.15)10⁴ below 1.2 keV, and average radiation width <Γ_γ>=107±7 meV for lower 25 levels. Average capture cross section <σ_c> were obtained from 50 eV to 30keV.     

Thermal neutron cross section determination of short-to-medium lived nuclides using a 20 Ci Am–Be neutron source

While there are growing demands for the nuclear data at higher energy regions than keV for up-to-date scientific and technological development, accurate capture cross sections at thermal energy are still needed. The thermal neutron capture cross sections for the reactions ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm,¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U were determined by the method of foil activation using ⁵⁵Mn(n,γ)⁵⁶Mn as a reference reaction. The experimental samples with and without a Cd cover were irradiated in an isotropic neutron field of a 20 Ci ²⁴¹Am–Be neutron source facility. A high purity Ge detector was used to measure the induced gamma-rays from the samples and the monitor. The thermal neutron capture cross sections of the reactions ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm, ¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U were deduced from the analysis of obtained gamma-ray spectra. The thermal neutron capture cross section values for ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm, ¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U reactions are (5.93 ± 0.52), (207.3 ± 9.4), (7.7 ± 0.3), and (2.79 ± 0.09) barns respectively. The obtained results have been discussed and compared with the available experimental data and were found to be in agreement with each other.     

Neutron Nuclear Data Evaluation for Thorium-232

An evaluation was made on the neutron cross sections, resonance parameters and average neutron yield in fission for ²³²Th in the energy range from thermal energy to 20 MeV. The fission and capture cross sections were evaluated on the basis of the experimental data by converting the relative ratio data into cross section values by making use of recent evaluations for reference cross sections. The total cross section was determined from experimental data in the region from 24 keV to 15 MeV and then extrapolated to lower and higher energies by using the optical model whose parameters had been adjusted as so to reproduce the measured data. The elastic and inelastic scattering, (n, 2n) and (n, 3n) reaction cross sections were calculated by means of the statistical model combined with the optical model. A set of resonance parameters were recommended in the energy range below 3.5 keV and average resonance parameters were deduced in the unresolved resonance region. A value of 7.40 b was chosen for the capture cross section at 0.025 eV, and the picket-fence negative-energy levels were introduced so as to reproduce the non-l/v behavior of the capture cross section in the epithermal region.

The results were incorporated in the Japanese Evaluated Nuclear Data Library, Version 2 (JENDL-2). Comparison was made between the present and other evaluations such as ENDF/B-V and possible reasons for the discrepancy were discussed.     

Measurement of thermal neutron and resonance integral cross sections of the reaction V(n,γ)V using a 20 Ci Am–Be isotopic neutron source

The thermal neutron capture cross section (σ_o) and the resonance integral (I_o) of the ⁵¹V(n,γ)⁵²V reaction were measured with an activation method to provide fundamental data for reactor calculation, activation analysis, and other theoretical and experimental uses concerning the interaction of neutron with matter. The vanadium and manganese samples were irradiated within and without a Cd shield case using a 20 Ci Am–Be neutron source. The activities of the samples were measured using gamma-ray spectroscopy. The thermal neutron capture cross section and the resonance integral were determined relative to the reference reaction ⁵⁵Mn(n,γ)⁵⁶Mn and the values obtained are 5.16 ± 0.19 barns and 2.53 ± 0.1 barns respectively. The previous measurements of the σ_o and I_o of the reaction ⁵¹V(n,γ)⁵²V were reviewed and the difference between the present values and the previous results were discussed.     

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.     

Resonance analysis of 151,153Eu from neutron capture cross section measurements in the energy range from 1 to 20 eV

The neutron capture cross sections of Europium-151 and Europium-153 have been measured by the time-of-flight method in the energy range from 0.005 to 100 eV using the Kyoto University Research Reactor Institute-Linear Accelerator (KURRI-LINAC). An assembly of Bismuth Germanate (BGO) scintillators was used to detect the prompt capture of γ rays. The absolute values of the neutron capture cross sections of ¹⁵¹Eu and ¹⁵³Eu were deduced by normalizing the thermal capture cross sections in JENDL-4.0 and ENDF/B-VII.1, respectively. Then, we have obtained the resonance parameters of 20 resonances in ¹⁵¹Eu and 17 resonances in ¹⁵³Eu using the code SAMMY.

For the 3.36-eV resonance of ¹⁵¹Eu, the evaluated resonance peak area in JENDL-4.0 is about 95% smaller than the present result. For the 7.00-, 7.22-, and 7.42-eV resonance; we confirmed that there are significant differences between the measured peaks and evaluated peaks in JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2. For the ¹⁵³Eu, the evaluated resonance peak areas in JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2 are about 15% larger than the measured resonance peak areas at the 2.46-, 3.29-, and 3.94-eV resonances.     

Measurements of thermal-neutron capture cross-section and resonance integral of neptunium-237

The thermal-neutron capture cross-section (σ₀) and resonance integral (I₀) were measured for the ²³⁷Np(n,γ) ²³⁸Np reaction by an activation method. A method with a Gadolinium filter, which is similar to the Cadmium difference method, was used to measure the σ₀ with paying attention to the first resonance at 0.489 eV of ²³⁷Np, and a value of 0.133 eV was taken as a cut-off energy. Neptunium-237 samples were irradiated at the pneumatic tube of the Kyoto University Research Reactor in Institute for Integral Radiation and Nuclear Science, Kyoto University. Wires of Co/Al and Au/Al alloys were used as monitors to determine thermal-neutron fluxes and epi-thermal Westcott’s indices at an irradiation position. A γ-ray spectroscopy was used to measure activities of ²³⁷Np, ²³⁸Np and neutron monitors. On the basis of Westcott’s convention, the σ₀ and I₀ values were derived as 186.9 ± 6.2 barn, and 1009 ± 90 barn, respectively.     

Reactor Neutron Capture Cross Section of 60.3-day 124Sb

The radioactive nuclides ¹²⁴Sb (T _1/2=60.3d) and ¹²⁵Sb (T_1/2=2.77yr) were produced from natural antimony by JRR-3 reactor irradiation of 283.5 h through the single and double capture processes. After cooling of 3.50 yr, the γ-ray spectrum of the antimony sample irradiated was measured by a 50 cc coaxial type Ge(Li) detector, and the photo-peak yield ratio of ¹²⁵Sb (E _r=428keV) to ¹²⁴Sb (E _r=1.691 MeV) was obtained. By using a relation between this photo-peak yield ratio and the ¹²⁴Sb (n, γ) ¹²⁶Sb cross section, the reactor neutron capture cross section of 60.3-day ¹²⁴Sb was obtained as 17.4:5:^+2.8 _?2.5b. The thermal neutron flux at the position of antimony sample irradiated was estimated as (4.92±0.38) ×10¹²n/cm²·s by measuring the 1.333-MeV photo-peak yield of ⁶⁰Co, which was activated by reactor irradiation of cobalt impurity contained in the antimony sample.