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.     

贫铀/聚乙烯交替球壳中裂变反应率的测量与分析

为校验次临界能源堆的概念设计,采用活化法在贫铀/聚乙烯球壳交替装置上开展14 MeV中子学积分实验。用HPGe探测器测量²³⁸U(n,f)及²³⁵U(n,f)反应的裂变碎片¹⁴³Ce衰变产生的293.3 keV特征γ射线,得到装置中与入射D粒子束成90°方向上的²³⁸U(n,f)及²³⁵U(n,f)反应率分布,相对不确定度为5.1%~6.9%。采用MCNP5程序在ENDF/B-Ⅵ库下进行模拟计算,计算结果较实验结果高约5%。     

EMPIRE计算30MeV以下~(238)U中子核反应数据

为进一步完善核数据评价手段,本文将EMPIRE应用到中子引起锕系核素的核反应模型分析中,根据中子核反应机制的特点,选取恰当的核反应模型及模型参数,以实验数据为基础对模型参数进行调整,由EMPIRE计算获得30 MeV以下能区n+238 U的核反应数据。从计算结果与实验数据以及各评价库数据对比来看,EMPIRE可得到较合理的结果。     

U measurement with accelerator mass spectrometry at CIAE

²³⁶U is a long-lived radioactive isotope which is produced principally by thermal neutron capture on ²³⁵U. ²³⁶U may be potentially applied in geological research and nuclear safeguards. Accelerator mass spectrometry is presently the most sensitive technique for the measurement of ²³⁶U and a measurement method for long-lived heavy ion ²³⁶U has been developed. The set-up uses a dedicated injector and the newly proposed ²⁰⁸Pb¹⁶ molecular ions for the simulation of ²³⁶U ion transport. A sensitivity of lower than 10⁻¹⁰ has been achieved for the isotopic ratio ²³⁶U/²³⁸U in present work.     

The investigation of the neutron emission spectra of Th and U for neutron incident energy from 2 to 18 MeV

In this study, experimental neutron-emission spectra produced by (n, xn) reactions on nuclei ²³²Th have been compared with experimental ²³⁸U(n, xn) neutron-emission spectra from 2 to 18 MeV. Angle-integrated cross-sections in neutron induced reactions on targets ²³⁸U have been calculated at the bombarding energies from 2 to 18 MeV. In the calculations, the geometry dependent hybrid model and the cascade exciton model including the effects of pre-equilibrium have been used. In addition, we have described how multiple pre-equilibrium emissions can be included in the Feshbach–Kerman-Koonin (FKK) fully quantum-mechanical theory. By analyzing (n, xn) reaction on ²³⁸U with the incident energy from 2 to 18 MeV, the importance of multiple pre-equilibrium emission can be seen clearly. All calculated results have been compared with experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other.     

Fission Product Yields of U, U, U and Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for ²³⁵U(n,f), ²³⁹Pu(n,f) in a thermal spectrum, for ²³³U(n,f), ²³⁵U(n,f), and ²³⁹Pu(n,f) reactions in a fission neutron spectrum, and for ²³³U(n,f), ²³⁵U(n,f), ²³⁸U(n,f), and ²³⁹Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.     

Accelerator mass spectrometry of U at low energies

The performance of the compact ETH-TANDY system for accelerator mass spectrometry measurements of ²³⁶U is presented. Despite the low ion energies of around 1.2 MeV we can demonstrate a background level that is comparable to larger facilities. The careful ion-optical design of the high-energy spectrometer leads to a high suppression of neighboring isotopes sufficient to measure samples with isotopic ratios of ²³⁶U/²³⁸U > 10^-11 the ion chamber only, as demonstrated by systematic investigations with different slit settings and time-of-flight measurements. Additionally, a high overall efficiency is achieved due to a high transmission through the accelerator.     

Reaction-yield dependence of the (γ, γ′) reaction of 238U on the target thickness

The dependence of the nuclear resonance fluorescence (NRF) yield on the target thickness was studied. To this end, an NRF experiment was performed on ²³⁸U using a laser Compton back-scattering (LCS) γ-ray beam at the High Intensity γ-ray Source facility at Duke University. Various thicknesses of depleted uranium targets were irradiated by an LCS γ-ray beam with an incident beam energy of ～2.475 MeV. The scattering NRF γ-rays were measured using an High-purity Germanium (HPGe) detector array positioned at scattering angles of 90° relative to the incident γ-beam. An analytical model for the NRF reaction yield (NRF RY model) is introduced to interpret the experimental data. Additionally, a Monte Carlo simulation using GEANT4 was performed to simulate the NRF interaction for a wide range of target thicknesses of the ²³⁸U. The measured NRF yield shows the saturation behavior. The results of both of the simulation and the analytical model can reproduce the saturation curve of the scattering NRF yield of ²³⁸U against the target thickness. In addition, we propose a method to deduce the precise integral cross section of the NRF reaction by fitting the NRF yield dependency on the target thickness without any absolute measurements.     

Fast Neutron Dosimetry by 238U(n,2n) Reaction together with Thermal Neutron Dosimetry by 238U(n, γ) Reaction

The fast neutron flux in the central thimble of a TRIGA-II reactor was estimated by using the ²³⁸U(n,2n)²³⁷U reaction as flux monitor. The thermal neutron flux also was estimated by ²³⁸U(n,γ)→ ²³⁹Np and ²³²Th(n,γ)→ ²³³Pa reactions. The latter flux was further measured at the same time by γ-ray spectrometry of the product nuclides and by chemical separation and β counting with αemitting isotopes as tracer. The values of flux thus obtained were compared with those estimated from the amounts of ²⁸Al, ²⁷Mg and ²⁴Na respectively produced by the (n,γ), (n,p) and (n,α) reactions on ²⁷Al. Furthermore, for the first time, the effective cross section for the ²³⁸U(n,2n)²³⁷U reaction was estimated to be 2.8 b from a comparison with that for the ²⁷Al(n,α)²⁴Na reaction.     

Prompt fission neutron multiplicity and spectrum evaluations for n+238U reaction

The prompt fission neutron multiplicity and spectra for n+²³⁸U reaction are calculated using an improved Los Alamos model which includes the linear relation between the average prompt gamma ray energy and the prompt neutron multiplicity and also the average fission fragment kinetic energy dependence on the incident neutron energy. The coefficients describing the quadratic variation of the fission fragment kinetic energy versus the incident energy are obtained by extrapolation of the data and procedure used for n+²³⁵U reaction. The inverse process compound nucleus cross-section of the fissioning nucleus is calculated using the coupled channel method. In the incident energy range where only the first fission chance is involved the comparison of present spectrum evaluation with spectrum calculation using multi-modal model is made too. The calculated prompt neutron multiplicity and spectra of ²³⁸U neutron induced fission are in good agreement with the experimental data for the entire incident energy range required in evaluations, proving the validity of the used procedure.     

<Superscript>239</Superscript>Pu prompt fission neutron spectra

Calculations of ²³⁹Pu(n, F) prompt fission neutron spectra have been performed for neutron energy up to 20 MeV. The exclusive spectra of pre-fission neutron reactions (n, xnf) were calculated on the basis of the Hauser-Feshbach model simultaneously with the cross sections of (n, F) and (n, 2n) reactions. The spectra of neutrons emitted by fission fragments were approximated by a sum of two Watt distributions. The components of the prompt fission neutron spectra due to pre-fission neutrons are manifested in the prompt fission neutron spectra and the average neutron energy. A correlation is established between this effect in the contribution of emissive fission (n, xnf) in the fission cross-section of ²³⁹Pu(n, F) and ²³⁵U(n, F). It is shown that the ²³⁹Pu(n, F) prompt fission neutron spectra used in applied calculations do not correspond to the experimental differential data and the systematic regularities in the spectra and their average energy found for the most carefully studied nuclei ^235,238U and ²³²Th. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 119–124, August, 2007.     

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 the thermal-neutron-induced fission cross-section of 238U

The thermal-neutron-induced fission cross-section of ²³⁸U was measured at an intense and very clean thermal-neutron beam of the Grenoble high-flux reactor. The best ²³⁸U sample material used contained only 12 ppb of ²³⁵U. The fission fragments were detected with surface barrier detectors. Relative to a thermal fission cross-section of 587.6 barn for ²³⁵U, a value of (11 ± 2) μbarn was obtained for the ²³⁸U fission cross-section. By comparing this result with ²³⁸U(n, f) measurements in the resonance region, the existence of a negative resonance or resonances in ²³⁹U is demonstrated.     

Measurement of photo-neutron cross-sections in Pb and Bi with 50-70 MeV bremsstrahlung

The photo-neutron cross-sections of ²⁰⁸Pb and ²⁰⁹Bi induced by 50-70 MeV bremsstrahlung have been measured by using the off-line γ-ray spectrometric technique in the electron linac at the Pohang Accelerator Laboratory. The experimental ²⁰⁸Pb(γ,xn) and the ²⁰⁹Bi(γ,xn) reactions cross-sections at the bremsstrahlung energy region of 50-70 MeV, which are determined for the first time, are in general good agreement with the theoretical values based on the TALYS 1.0 code. We observed that the photo-neutron cross-sections for the (γ,xn) reactions of ²⁰⁹Bi and ²⁰⁸Pb increase with increasing of bremsstrahlung energy from 50 to 70 MeV, which indicates the role of excitation energy. It was also observed that the (γ,xn) reaction cross-sections of the doubly magic shell nuclei ₈₂Pb¹²⁶ are always higher than those of the singly magic shell nuclei ₈₃Bi¹²⁶ in the bremsstrahlung energy of 50-70 MeV. This may be due to the fact that either the shell effect are washed out at the higher excitation energy or due to the lower fission cross-section of ²⁰⁸Pb compared to that of ²⁰⁹Bi. This observation indicates that there is a competition reaction between fission and neutron emission.     

Relative yield of delayed neutrons and half-life of their precursor nuclei from <Superscript>238</Superscript>U fission by 14.2–17.9 MeV neutrons

The energy dependence of the relative yield of delayed neutrons and half-life of their precursor nuclei from ²³⁸U fission by 14.2–17.9 MeV neutrons is measured. In addition, the detector-blocking effect, which is due to the exposure of the detector to intense fluxes of high-energy neutrons, and the effect of the secondary neutron source D(d, n)³He, a reaction which occurs when a T(d, n)⁴He based neutron source is used, are investigated experimentally and taken into account in the data analysis. The data are analyzed in terms of the average half-life < T> of the precursor nuclei of the delayed neutrons. It is shown that the observed increase of <T> with increasing primary-neutron energy in the energy range studied is predominantly due to the opening of emissive nuclear fission channels. The energy dependence of the average half-life of the delayed-neutron precursors can be described on the basis of the model of emissive fission of three nuclei that fission in this energy range − ²³⁷U*, ²³⁸U*, and ²³⁹U*. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 124–132, February, 2007.     

Excitation functions of Ir formed in proton-induced reactions on highly enriched Os up to 66 MeV

Cross sections of proton-induced nuclear reactions on highly enriched ¹⁹²Os were measured up to 66 MeV by using the stacked-foil irradiation technique. Excitation functions are presented for the reactions ¹⁹²Os(p,n)¹⁹²Ir, ¹⁹²Os(p,3n)¹⁹⁰Ir, ¹⁹²Os(p,4n)¹⁸⁹Ir, ¹⁹²Os(p,5n)¹⁸⁸Ir, ¹⁹²Os(p,6n)¹⁸⁷Ir and ¹⁹²Os(p,7n)¹⁸⁶Ir. No earlier experimental cross-section data could be found in the literature except for the ¹⁹²Os(p,n)¹⁹²Ir process. Our new experimental results are compared with theoretical predictions by means of the theoretical model code ALICE/ASH. Integral thick-target yield calculations were also performed for the ¹⁹²Os(p,n)¹⁹²Ir and ¹⁹²Os(p,3n)¹⁹⁰Ir reactions to evaluate the ¹⁹⁰Ir contamination level, as a function of energy, in the case of ¹⁹²Ir productions.     

Nucleogenic Cl, U and Pu in uranium ores

The nucleogenic isotopes ³⁶Cl, ²³⁶U and ²³⁹Pu are produced naturally in subsurface environments via neutron capture of thermal and epithermal neutrons. Concentrations are, however, very low and accelerator mass spectrometry (AMS) is required for quantitative measurements. A particular challenge is presented by the measurement of ²³⁶U/²³⁸U ratios down to the level of 10⁻¹³ that is expected from rocks with low uranium concentration. Here, we present the AMS methodology that has been developed at the ANU for measuring ²³⁶U/²³⁸U ratios at this level. The more established methodologies for ³⁶Cl and ²³⁹Pu measurements are also summarised. These capabilities are then used to characterize the ³⁶Cl, ²³⁶U and ²³⁹Pu concentrations in a range of uranium ores. A simple model of the neutron production and capture processes in subsurface environments has been developed and is presented. It is shown that nucleogenic ³⁶Cl, ²³⁶U and ²³⁹Pu can be used to determine both thermal and epithermal neutron fluxes in subsurface environments. Potential applications include uranium exploration and monitoring of the environmental impact of uranium mining.     

Effects of Mutual Shielding on Resonance Integrals of 235U, 238U, 239Pu and 240pu below 150 eV

The results of some quantitative studies on resonance interference are presented. The calculations were performed on a FORTRAN IV program RICM2, which solves numerically the slowing down of neutrons over many resonance levels in a two region lattice, and gives reaction rates, average cross sections and effective resonance integrals of the nuclides concerned.

Three combinations of resonant nuclides, ²³⁵U-²³⁸U, ²³⁰Pu-²³⁸U and ²³⁹Pu-²¹⁰Pu, were considered, in conjunction with three oxide fuel rod radii, 0.2, 0.5 and 2.0 cm, the moderator (light water) to fuel volume ratio being maintained constant at 2.0. An energy range below 150eV has been covered by the present calculations. The effects of resonance interference have been found to be appreciable in this energy range.     

Development of laboratory standards for AMS measurement of 237Np

A new method for the preparation of ²³⁷Np/²³⁸U laboratory standards for accelerator mass spectrometry (AMS) measurement was developed at China Institute of Atomic Energy (CIAE). ²³⁷Np was generated by β-decay (6.75 days) of ²³⁷U produced via fast neutron induced ²³⁸U(n,2n)²³⁷U reaction. The ⁵⁸Co(n,2n)⁵⁹Co and ⁹³Nb(n,2n)^92mNb reactions were used to monitor the integral incident neutron flux. The cross-section of ²³⁸U(n,2n)²³⁷U was determined by measuring the emission rate of 208.0 keV γ-rays from ²³⁷U decay with a calibrated HPGe γ-ray spectrometer at CIAE. In order to correct for the self-attenuation of 208.0 keV γ-ray in the U sample, a ¹⁷⁷Lu–U method was introduced. The new correction method provided effective and high quality measurements of the cross-section for the ²³⁸U(n,2n)²³⁷U reaction; and the ²³⁷Np/²³⁸U ratio. A series of AMS standards for ²³⁷Np/²³⁸U (10^?9–10^?13) can be obtained by successive dilution of the original standard sample.     

Fission-product yields from neutron-induced fission

Exsting experimental thermal, fast, and 14-MeV neutron-induced fission-product cumulative and independent yieds have been compiled, corrected to common reference values, and listed in tabular form for the following fissile nuclides:Thermal-neutron fission: cumulative yields for ²²⁷Th, ²²⁹Th, ²³³U, ²³⁵U, ²³⁹Pu, ²⁴¹Pu, ²⁴¹Am, ²⁴²Am, ²⁴⁵Cm, ²⁴⁹Cf, ²⁵¹Cf, ²⁵⁴Es, and ²⁵⁵Fm; independent yieds for ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu.Fast-neutron fission: cumulativ yields for ²²⁷Ac, ²³¹Pa, ²³²Th, ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu; independent yields for ²³⁵U and ²³⁸U.14-MeV-neutron fission: cumulative yields for ²³¹Pa, ²³²Th, ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu; independent yields for ²³²Th, ²³³U, ²³⁵U, ²³⁸U, and ²³⁹Pu.11-MeV-neutron fission: cumulative yields for ²³²Th.3-MeV-neutron fission: cumulative yields for ²³¹Pa, ²³²Th, and ²³⁸U.1.1-MeV-neutron fission: cumulative yields for ²³⁷Np.From these experimental values the unknown independent yields are deduced empirically for thermal-neutron fission of ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu; the fast fission of ²³²Th, ²³³U, ²³⁵U, ²³⁸U, ²³⁹Pu, ²⁴⁰Pu, and ²⁴¹Pu (the chain yields for ²⁴⁰Pu and ²⁴¹Pu used at this energy being predictions); and the 14-MeV-neutron fission of ²³²Th, ²³³U, ²³⁵U, and ²³⁸U.Finally, by the fitting of the preceding information to condition equations derived from the conservation laws, adjusted sets of chain and independent yields are calculated for thermal fission of ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu; fast fission of ²³²Th, ²³³U, ²³⁵U, ²³⁸U, ²³⁹Pu, and ²⁴¹Pu; and 14-MeV fission of ²³²Th, ²³³U, ²³⁵U, and ²³⁸U. The literature search is probably complete to the end of 1975; some 1976 results are included.This paper replaces and makes obsolete the following UKAEA reports: AERE-R7209, AERE-R7394, AERE-R7680, and AERE-R8152.