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.     

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 of Thermal Neutron Capture Cross Section and Resonance Integral of the 129I(n, γ) 130I Reaction

To obtain fundamental data for research on the transmutation of nuclear wastes, the thermal neutron cross section and the resonance integral of the ¹²⁹I(n, γ)¹³⁰I reaction have been measured using an activation method. The neutron cross sections for the formation of the ground (5⁺) state and the isomeric (2⁺) state of ¹³⁰I were measured separately.

Six ¹²⁹I targets were irradiated for 10 min with thermal reactor neutrons; three of them containing 2.55- 2.61 kBq of ¹²⁹I were irradiated within a Cd capsule, and the other three targets containing 259–261 Bq of ¹²⁹I were irradiated without it. The Co/Al and Au/Al alloy wires were used to monitor the neutron flux and the fraction of the epithermal part (Westcott's epithermal index). The gamma-ray spectra from the irradiated samples were measured with a Ge detector.

The thermal neutron capture cross section (the 2,200 m/s neutron cross section) and the resonance integral of the ¹²⁹I(n, 7)¹³⁰I reaction were determined to be 12.5±0.5b and 15.6±0.7b for the formation of the ground state ^130gI(5+), 17.8±0.7b and 18.2±0.8b for the formation of the isomeric state ^130mI(2⁺), and 30.3±1.2b and 33.8±1.4b for the formation of ¹³⁰I (the sum of the 2⁺ and the 5⁺ states), respectively. The sum of the thermal neutron capture cross sections forming the 2⁺ and the 5⁺ states was 12% larger than the evaluated values of JENDL-3.2 and ENDF/B-VI and that reported by Roy et al. This discrepancy is explained by the population of the isomeric level.     

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 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.     

Estimation of Covariance Matrices for Nuclear Data of 237Np, 241Am and 243Am

Covariance matrices were estimated for the fission and capture cross sections and the numbers of neutrons per fission of ²³⁷Np, ²⁴¹Am and ²⁴³Am given in JENDL-3.3. GMA and KALMAN codes were applied to estimate them for the fission and capture cross sections, respectively. In the low energy region, the errors of resonance parameters were given. The covariance matrices for the numbers of prompt neutrons per fission (V_p ) were evaluated by assuming a linear equation. For the delayed neutrons (v_d ), only their standard deviations were estimated. The results were compiled in the ENDF-6 format and merged with JENDL-3.3.     

Neutron Economy of Transmutation of TRU in Thermal and Fast Neutron Fields

Neutron economy of the transmutation of TRU was examined in well thermalized, thermal and fast neutron fields. Burn-up chains of ²³⁷Np, ²⁴¹Am and ²⁴³Am, which are the main TRU nuclides in the high level waste, were calculated in the flux region from 10¹⁴ to 10¹⁷ n/cm².s. Numbers of neutrons absorbed and produced of each chain were calculated using JENDL-3. The net number of neutron produced n _net, which was obtained by the difference of the two numbers, largely varied with the neutron fields, the nuclides and the flux levels. The n _net value in the fast neutron field was positive (0.0–1.0) for ²³⁷Np, ²⁴¹Am, ²⁴³Am and TRU with the nuclide composition in the high-level waste generated by the conventional PWR. The transmutation of TRU by fission can be performed with producing neutrons in the fast neutron field. On the other hand, the n _net value was negative for the well thermalized and thermal neutron fields. For TRU in the high-level waste, the values in those fields were —1.0 at 10¹⁴ n/cm².s and 0.0 at 10¹⁶ n/cm².s. In the high flux region of 10¹⁶ n/cm².s, TRU in the high-level waste can be transmuted by fission without consuming additional neutrons. In the flux region about 10¹⁴ n/cm².s, the transmutation of TRU in the high-level waste by fission requires about one neutron.     

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.     

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.     

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.     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

Emission Probabilities of Gamma Rays from the Decay of 233Pa and 238Np,and the Thermal Neutron Capture Cross Section of 237Np

In order to determine the thermal neutron capture cross section of ²³⁷Np, the relevant γ emission probabilities of the 312-keV γ-ray from the decay of ²³³Pa and the 984-keV γ-ray from the decay of ²³⁸Np are deduced from the ratio of the emission rate to the activity. The emission rate and activity are measured with a Ge detector and a Si detector, respectively. The measured emission probability for 312-keV γ-ray is 41.6±0.9% and that for 984-keV γ-ray is 25.2±0.5%. The emission probabilities are used to correct the thermal neutron capture cross section of ²³⁷Np reported previously, and gives 168±6b. The neutron capture cross section is also determined as 169±6b by α-ray spectroscopic method. The measured emission probabilities and capture cross section are compared with others from references. By averaging these values deduced by different methods, the value of 169±4b is recommended as the thermal neutron capture cross section of ²³⁷Np for 2,200 m/s neutrons.     

Production of the Isomeric State of 138Cs in the Thermal Neutron Capture Reaction 137Cs(n, γ)138Cs

In order to obtain precise data of the neutron capture cross section of the reaction ¹³⁷Cs(n, γ)¹³⁸Cs, the production probability of isomer state ^138mCs was measured in this work. Targets of about 0.37MBq ¹³⁷Cs were irradiated for 3 min in. the pneumatic tube facility (Pn-3) of Kyoto University Reactor (KUR). The 1,436 keV gamma;-ray emitted from both of ^138gCs and ^138mCs was measured. A ratio of the production probability between ^138gCs and ^138mCs was deduced from time dependence of peak counts of 1,436 keV γ-ray by making use of difference of half-lives of ^138gCs (33.41 min) and ^138mCs (2.91 min). The production probability of ^138mCs was obtained as 0.75plusmn;0.18 and this value revised the effective cross section upwards by 9plusmn;2percnt;. The effective cross section ô and the thermal neutron capture cross section σo were obtained as ô=0.29±0.02 b and σ=0.27±0.03b with taking into account the production of ^138mCs.     

CdZnTe quasi-hemispherical detector for gamma–neutron detection

Quasi-hemispherical CdZnTe detector was manufactured successfully to fully understand the performance in the mixed gamma–neutron detection field. Together with the software of COMSOL, Geant4, and Matlab, the detector structure has been optimized. The CdZnTe detector performs good energy resolutions for ²⁴¹Am, ⁵⁷Co, and ¹³⁷Cs radiation sources, especially for ¹³⁷Cs (10.91 keV full width at half maximum [FWHM] at 662 keV). A linear relationship between the energy positions and spectrum channels indicates that the detector is effective for the precise energy detection from 59.5 to 662 keV. Finally, neutron and gamma events were detected simultaneously at room temperature using ²⁴¹AmBe neutron source. The spectrum shows good energy resolution for neutron capture gamma ray (14.28 keV FWHM at 558 keV). Our work demonstrates that the quasi-hemispherical CdZnTe detector is promising for simultaneous detection of neutrons and gamma radiation.     

Evaluation of differential nuclear data for americium isotopes part I: Am

An evaluation of differential neutron data has been carried out for ²⁴¹Am, the result of which has been incorporated into the U.K. Nuclear Data Library as DFN 1009B. The evaluation covers the energy range from 10^?5 eV to 15 MeV and includes the total, capture, fission, elastic and inelastic scattering, (n, 2n) and (n, 3n) cross-sections ?n and the fission neutron spectrum. Also discussed are the half-life and decay of ²⁴¹Am, the branching of the radiative capture cross-section to form the ground and isomeric states of ²⁴²Am and resonance integrals. Where possible the evaluation is based on measured data, but much use has been made of nuclear reaction theory and systematic behaviour of the nuclear parameters of the actinides to calculate the required data where no measurements exist. Uncertainties have been assigned at representative energies throughout the range. Some of the results of the evaluation are compared with data from other evaluations.     

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.     

Comparisons of HELIOS,ORIGEN2, and Monteburns Calculated 241 Am and 243 Am Concentrations to Measured Values for PWR,BWR, and VVER Spent Fuel

A study was performed at Los Alamos National Laboratory to explore the accuracy of several reactor analysis codes in calculating ²⁴¹ Am and ²⁴³Am concentrations in light water reactor spent fuel. Calculated higher-actinide concentrations were compared to measured values from the literature for three reactor fuels. The fuel samples were taken from the Mihama Unit 3 pressurized water reactor, the Garigliano boiling water reactor, and a VVER-440. The ²⁴¹Am and ²⁴³Am concentrations were calculated using the HELIOS-1.4 lattice-physics code, the ORIGEN2 burnup code, and a linked MCNP/ORIGEN2 code named Monteburns 3.01. Comparisons were made between the calculated and measured values. It was determined that all codes performed consistently well for the Mihama Unit 3 measurements (within ±5% for ²⁴¹Am and ±20% for ²⁴³Am) and the Garigliano measurements (within ±12% for ²⁴¹ Am and ±20% for ²⁴³Am). It was determined that the ORIGEN2 pressurized water reactor libraries are insufficient for the VVER-440 measurements. The HELIOS and MONTEBURNS codes both demonstrated good ability to calculate these isotopes for VVER-440 fuel (±10% for ²⁴¹Am and ±12% for ²⁴³Am). The accuracies of these codes and the associated radiochemical measurements of these higher-actinide isotopes may be insufficient for safeguards and fuel management purposes; thus, development of new methods and modification to existing data libraries may be necessary in order to enable cost-effective safeguarding of these higher-actinide materials.     

Correlation of Amounts of Transplutonium Nuclides with Burnup in the JPDR-I Spent Fuel

The amounts of ²⁴¹Am, ²⁴²Cm and ²⁴⁴Cm were determined by means of a radiochemical technique in several specimens taken from the spent fuel of the JPDR-I. The yields of the transplutonium nuclides were examined in connection with the burnup determined by a conventional method. It was found that the burnup correlates well with the yield ratios of various transplutonium nuclides.     

Measurements of Thermal Neutron Cross Section and Resonance Integral for 237Np(n, γ)238Np Reaction

Making use of a standard neutron spectrum field with a pure Maxwellian distribution, the thermal neutron cross section for the ²³⁷Np(n, γ)²³⁸Np reaction was measured at a neutron energy of 0.0253 eV by the activation method. The result is 158±3 b, which is obtained relative to the reference value of 98.65±0.09 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction. Although the data in JENDL-3 is larger by about 15% than the present value, the recently revised data in JENDL-3.2 is close to the present. The ENDF/B-V, ENDF/B-VI, JEF-2 and Mughabghab's data are also larger by 7–15%. Old measurements are larger by 7–18% than the present data.

The resonance integral for the ²³⁷Np(n, γ)²³⁸Np reaction was also measured relative to the reference value of 1,550±28 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction with a 1/E standard neutron spectrum field. By defining the Cd cut-off energy as 0.5 eV for the ²³⁷Np(n, γ)²³⁸Np reaction, the present resonance integral is 652 ± 24 b, which is in good agreement with the JENDL-3, -3.2, ENDF/B-V, -VI, JEF-2 and Mughabghab's data. However, most of the old experimental data are, in general, larger by 24–38% than the present measurement.     

Energy Response of Neutron Area Monitor with Silicon Semiconductor Detector

Abstract

A prototype neutron area monitor with a silicon semiconductor detector has been developed which has the energy response of 1 cm dose equivalent recommended by the 1CRP-26. Boron and proton radiators are coated on the surface of the silicon semiconductor detector. The detector is set at the center of a cylindrical polyethylene moderator. This moderator is covered by a porous cadmium board which serves as the thermal neutron absorber. Neutrons are detected as α-particles generated by the nuclear reaction ¹⁰B(n, α)⁷Li and as recoil protons generated by the interaction of fast neutrons with hydrogen.

The neutron energy response of the monitor was measured using thermal neutrons and monoenergetic fast neutrons generated by an accelerator. The response was consistent with the 1cm dose equivalent response required for the monitor within ±34% in the range of 0.025–15 MeV.