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.     

Fission Product Yields for 14 MeV Neutrons on U, U and Pu

We report cumulative fission product yields (FPY) measured at Los Alamos for 14 MeV neutrons on ²³⁵U, ²³⁸U and ²³⁹Pu. The results are from historical measurements made in the 1950s–1970s, not previously available in the peer reviewed literature, although an early version of the data was reported in the Ford and Norris review. The results are compared with other measurements and with the ENDF/B-VI England and Rider evaluation. Compared to the Laurec (CEA) data and to ENDF/B-VI evaluation, good agreement is seen for ²³⁵U and ²³⁸U, but our FPYs are generally higher for ²³⁹Pu. The reason for the higher plutonium FPYs compared to earlier Los Alamos assessments reported by Ford and Norris is that we update the measured values to use modern nuclear data, and in particular the 14 MeV ²³⁹Pu fission cross section is now known to be 15–20% lower than the value assumed in the 1950s, and therefore our assessed number of fissions in the plutonium sample is correspondingly lower. Our results are in excellent agreement with absolute FPY measurements by Nethaway (1971), although Nethaway later renormalized his data down by 9% having hypothesized that he had a normalization error. The new ENDF/B-VII.1 14 MeV FPY evaluation is in good agreement with our data.     

Thermodynamic behavior of (U,Pu) mixed-oxide fuels

Transpiration experiments were performed at temperatures between 1000 and 1700°C to study the thermodynamics and defect structure of hypostoichiometric UO₂-20 wt % PuO₂ solid-solution systems. The oxygen partial pressures were established by using flowing $\text{H}{}_2\text{/H}{}_2\text{O}$ mixtures. After equilibration, the quenched products were analyzed by chemical, X-ray, neutron-diffraction and metallographic techniques. The versus temperature curves for the mixed oxide with different oxygen-to-metal ratios were plotted.Based on X-ray, neutron diffraction and metallographic data, it was concluded that the 20 wt % PuO₂ mixed oxide exists as a single phase under normal conditions, even at an oxygen-to-metal ratio as low as 1.92.The data from density measurements and neutron-diffraction analysis indicated that the predominant defects in the hypostoichiometric UO₂-20 wt % PuO₂ are anion vacancies.     

Mutual Separation of U,Pu and Fission Products with Tri-iso-octylamine

Uranium and Pu were extracted with tri-wo-octylamine (CC1₄) from irradiated U dissolved in sulfuric acid, from which Pu was back-extracted with sulfurous acid, and U with I M NH₄AC-1.5%H₂O₂. The decontamination factors were estimated to be 10² for U fraction (γ and β) activities) and >10³ for Pu fraction.     

1AW溶液中U、Np和Pu的分析

研究了H2O2同时调节镎、钚、铀价态至Np(Ⅳ)、Pu(Ⅳ)以及U(Ⅵ)的条件,在6mol/L HNO3浓度下,使用1.5%H2O2作为氧化还原剂对1AW进行调价,吸附上UTEVA柱并淋洗后,对钚、镎和铀进行洗脱。对模拟放射性样品进行预处理后,其中铀、镎、钚单独顺序洗脱的回收率分别为91.5%、119%、99.8%,137 Cs的去污因子高达7.4×104,单个样品操作时间约为1~1.5h;若钚洗脱后铀、镎同时洗脱并使用ED-XRF测量可以减少操作时间,铀、镎的回收率分别为102.4%、93.9%。均满足样品分析及辐射防护要求。     

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.     

Effective cross sections for fission of U233, U235, Pu239 and Pu240 by neutrons with energies from 30 kev to 5 Mev

Measurements have been made of the absolute values of the effective fission cross sections in U²³³, U²³⁵, Pu²³⁹ and Pu²⁴⁰, using as photoneutron sources Sb + Be ( 30 kev), Na + D₂O ( 250 kev), Na + Be ( 900 kev), a source with a simulated fission-neutron spectrum and Po--Be. The fission cross sections for 30 kev-neutrons are 3.06 ± 0.16, 2.21 ± 0.12 and 1.79 ± 0.11 barns for U²³³, U²³⁵ and Pu²³⁹, respectively. As the neutron energy Is increased from 30 kev to 250 kev the cross sections in U²³³ and U²³⁵ fall off by approximately 35% and then remain almost constant while the cross section in Pu²³⁹ falls off by 12% and then increases. The effective cross sections for fission in Pu²⁴⁰ for 900-kev and 5-Mev neutrons is 1.4 barns and the threshold for fission in Pu²⁴⁰ lies within the limits 250 and 900 kev.In conclusion the authors wish to express their gratitude to P. E. Spivak for help in formulating the problem and a discussion of the results.     

Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on Pu, U, U

We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for ⁹⁹Mo, ⁹⁵Zr, ¹³⁷Cs, ¹⁴⁰Ba, ^141,143Ce, and ¹⁴⁷Nd. Modest incident-energy dependence exists for the ¹⁴⁷Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ∼5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except for ⁹⁹Mo where the present results are about 4%-relative higher for neutrons incident on ²³⁹Pu and ²³⁵U. Additionally, our results illustrate the importance of representing the incident energy dependence of fission product yields over the fast neutron energy range for high-accuracy work, for example the ¹⁴⁷Nd from neutron reactions on plutonium. An upgrade to the ENDF library, for ENDF/B-VII.1, based on these and other data, is described in a companion paper to this work.     

A neutronic evaluation of the (Pu–U) and (Am–Pu–U) insertion in a typical fuel of Angra-I

The goal is to evaluate the neutronic behavior when (Pu–U) and (Am–Pu–U) mixed oxide are inserted in a typical cell of a Pressurized Water Reactor (PWR) such as Angra-I. Four types of fuels were studied: (1) MOX fuel enriched at 3.1% and V_m/V_f = 1.15; (2) MOX fuel enriched at 4.5% and V_m/V_f = 1.15; (3) MOX fuel enriched at 4.5% and V_m/V_f = 2.0 and (4) MOX fuel enriched at 4.5%, with 1% of Americium insertion in its composition (62.8% of Am²⁴¹, 0.1% of Am^242m and 37.1% of Am²⁴³) and with V_m/V_f = 2.0. The first case represents the standard state of Angra I, but with Pu. The second case is similar to the first but the enrichment is increased. To evaluate the Americium insertion, a study of the V_m/V_f was made and better results were obtained with V_m/V_f = 2.0 and to compare, this case was too evaluated to (Pu–U) in the third and fourth cases. The idea is to verify the possibility of using these fuels in Angra-I analyzing neutronic parameters such as infinite multiplication factor, hardening spectrum, Boron worth and reactivity temperature coefficients. The results show that it is possible to use all the studied fuels in Angra-I as well as to burn Am inserted in the MOX fuel by a considerable quantity during PWR operation. The WIMS-D5 code was used to perform a simplified neutronic and burnup simulations to evaluate this possibility.     

A measurement of the U233, U235, Pu239 total cross sections. As well as the fission cross section of U235 for resonance neutrons

The measurement of the interaction cross sections between resonance neutrons and fissionable isotopes are of definite importance both for reactor calculations and for the construction of various nuclear models. In this work we present the results of measurements of the U²³³, U²³⁵, and Pu²³⁹ total cross sections, as well as the fission cross section of U²³⁵. The measurements were performed on a mechanical neutron spectrometer (neutron chopper) with a resolution of 0.1–0.2 sec/m in the neutron energy region from 3–5 to about 500 ev. The resonance parameters are calculated to energies of about 30–50 ev, for which the levels may still be considered resolved. For these resonances the neutron widths g_n are determined; the total widths are determined for sufficiently excited levels when the uncertainty in this determination is no greater than about 50%. For U²³⁵ resonance in the energy region from 2.5 to 20 ev, in addition to the above parameters, the ratio of the fission width _f to the total width is found.     

Analysis of the observed resonance-width distributions of U233 and Pu239

As a result of the strong fluctuations of fission and reduced neutron widths, a significant number of resonances occur with such a small reduced height 2gF_n ⁰/ that they will not be noticed experimentally. If the fraction of transmitted resonances is considerable, then this should lead to the following effects which might be observable in an experiment: 1) change of the distribution function of the reduced neutron and fission widths, which is particularly sharply manifested for _n ⁰/<_n ⁰> 1 and _f/<_f> 1; 2)appearance of correlation between the form of the neutron distribution b of fission channels; 3) appearance in the total cross section and fission cross section of a background which is approximately proportional to . All the effects mentioned are manifested for the U²³³ nucleus. For Pu²³⁹ this effect is smal! and the observed values are welI described by X² distributions.Translated from Atomnaya Énergiya, Vol. 17, No. 1, pp. 22–27, July, 1964     

Uncertainty Analyses in the Resolved Resonance Region of 235U, 238U,and 239Pu with the Reich-Moore R-Matrix Theory for JENDL-3.2

A simple method to estimate covariances for resolved resonance parameters was developed. Although a large number of resolved resonances are observed for major actinides, uncertainties in averaged cross sections are more important than those in resonance parameters in reactor calculations. The method developed here derives a covariance matrix for the resolved resonance parameters which gives an appropriate uncertainty of the averaged cross sections. The method was adopted to evaluate the covariance data for ²³⁵U, ²³⁸U, and ²³⁹Pu resonance parameters in JENDL-3.2, with the Reich-Moore R-matrix formula.     

Prompt-Neutron Spectrum Formation Mechanism in 235U, 239Pu, and 252Cf Fission

The mechanism of the formation of the prompt-neutron spectrum in fissioning of ²³⁵U and ²³⁹Pu by thermal and fast neutrons and spontaneous fission of ²⁵²Cf is investigated. A method is proposed for the formation of the prompt-neutron spectrum in fissioning of nuclei as a superposition of three partial Weisskopf evaporation spectra with average neutron energy 0.4, 2.06, and 2.8 MeV.     

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.