Absolute measurements of U and Pu fission cross-sections with photoneutron sources

the fission cross-sections of ²³⁵U and ²³⁹Pu for Na---Be, La---Be, Na---D and Ga---D photoneutrons have been measured absolutely. The neutron source strength was measured using a manganese bath to compare the photoneutron yield from the sources with the standard source NBS-II. Fission counts were accumulated with the source positioned symmetrically between two identical fission foils in an experiment package suspended in a low-albedo laboratory. Fission fragments passing through limited solid angle apertures were recorded on polyester track-etch films. The masses of the foil deposits were determined by microbalance weighings and confirmed by thermal neutron fission and alpha counting. After making a correction for the calculated energy distribution of the source neutrons, values of 1.471 ± 0.029, 1.274 ± 0.026, 1.162 ± 0.025 and 1.195 ± 0.026 barns were obtained for the ²³⁵U fission cross-section at the source median energies of 140, 265, 770 and 964 keV, respectively. Corresponding values of 1.469 ± 0.045, 1.515 ± 0.038, 1.670 ± 0.039 and 1.643 ± 0.038 barns were determined for ²³⁹Pu.     

Cobalt Deposition and Release on Magnetite Powder in High Temperature Water

By making use of miniature ²³⁹Pu and ²³⁵U fission chambers and an accelerator-based neutron source, the ²³⁹Pu/²³⁵U neutron fission rate ratio profile has been measured in the region of a plane interface between a tank containing ordinary water and a tank containing an aqueous solution of boric acid. The reported preliminary results indicate an approximately 8 mm shift of the central point of the fission rate ratio profile towards the poisoned side of the system and an accompanying larger shift in the ²³⁵U Cd ratio profile towards the unpoisoned side.     

Statistical Estimation of Distributions of Physical Quantities in Nuclear Fission

Making use of a model based on the statistical theory, calculations were performed to obtain the mass yields, the most probable charges, the kinetic energies and the prompt neutron yields of fission fragments, and the mass yields of fission products from thermal-neutron-induced fission of ²³³U, ²³⁵U, ²³⁹Pu, ²⁴¹Pu and from the spontaneous fission of ²⁵²Cf. The calculations are further extended to fast-neutron-induced fission. The scission-point distance is treated as a parameter varying with the mass number of the heavy fragments of fission. This proved successful in approaching the calculated curves closer to the observed values.

It is possible to predict unknown physical quantities in nuclear fission with use made of the method developed in the present work.     

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.     

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.     

Representation of the fission neutron spectrum of235U,239Pu, and252Cf and the reactor spectrum as a superposition of five inelastic scattering functions

The results of investigations of the representation of the fission neutron spectra of²³⁵U,²³⁹Pu, and²⁵²Cf and neutron spectra at the core center of fast reactors as a superposition of five Weisskopf functions are presented. This representation best reflects the physics of nuclear fission and shows that the spectra at the core center of fast reactors are reconstructed with an error close to the error of the best studied spectra in the worldwide practice of neutron measurements. 3 tables, 13 references. All-Russia Scientific-Research Institute of Physical-Technical and Radio Engineering Measurements. Translated from Atomnaya énergiya, Vol. 88, No. 4, pp. 292–299, April, 2000.     

<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.     

Gamma-ray energy release in the decay of fission products

Total disintegration rate, gamma-ray energy release rate and energy spectrum of the fission products of ²³⁵U, ²³⁹Pu, ²⁴¹Pu, and ²³³U by thermal neutrons, and ²³⁵U, ²³⁸U, ²³⁹Pu and ²³²Th by fission spectrum neutrons have been reevaluated as a function of reactor operating times from 10² to 10⁸ sec after shutdown. Decay scheme data were taken mainly from the Table of Isotopes [1] and fission yields from Meek and Rider's 1972 recommendations [2]. Gamma energy releases do not depend strongly on the incident neutron energy, and those for ²³⁹Pu and ²⁴¹Pu are much different from that for ²³⁵U. Soon after fission, the present values for burst fission are lower than those of Perkins [3], but agree after 10⁴ sec and agree better with the experimental results of Sugarman et al. [4] and Borst [5]. The calculated data are tabulated in detail to facilitate interpolation.     

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.     

Integral testing of Np cross-sections in GCFR spectra

Integral measurements of ²³⁷Np capture and fission rates have been carried out relative to the ²³⁹Pu fission rate in two GCFR lattices with different neutron spectra. The results have been compared with calculated values based, respectively, on ENDF/B-IV and FGL5 cross-sections. Both data sets predicted the measured fission ratios correctly within the experimental errors of ±2% but, in the case of the ratio of ²³⁷Np capture to ²³⁹Pu fission, the calculated ratios differed from the measured values by about 10% for both sets.     

Performance of the gas turbine-modular helium reactor fuelled with different types of fertile TRISO particles

Preliminary studies have been performed on operation of the gas turbine-modular helium reactor (GT-MHR) with a thorium based fuel. The major options for a thorium fuel are a mixture with light water reactors spent fuel, mixture with military plutonium or with with fissile isotopes of uranium. Consequently, we assumed three models of the fuel containing a mixture of thorium with ²³⁹Pu, ²³³U or ²³⁵U in TRISO particles with a different kernel radius keeping constant the packing fraction at the level of 37.5%, which corresponds to the current compacting process limit. In order to allow thorium to act as a breeder of fissile uranium and ensure conditions for a self-sustaining fission chain, the fresh fuel must contain a certain quantity of fissile isotope at beginning of life; we refer to the initial fissile nuclide as triggering isotope. The small capture cross-section of ²³²Th in the thermal neutron energy range, compared to the fission one of the common fissile isotopes (²³⁹Pu, ²³³U and ²³⁵U), requires a quantity of thorium 25–30 times greater than that one of the triggering isotope in order to equilibrate the reaction rates. At the same time, the amount of the triggering isotope must be enough to set the criticality condition of the reactor. These two conditions must be simultaneously satisfied. The necessity of a large mass of fuel forces to utilize TRISO particles with a large radius of the kernel, 300 μm. Moreover, in order to improve the neutron economics, a fuel cycle based on thorium requires a low capture to fission ratio of the triggering isotope. Amid the common fissile isotopes, ²³³U, ²³⁵U and ²³⁹Pu, we have found that only the uranium nuclides have shown to have the suitable neutronic features to enable the GT-MHR to work on a fuel based on thorium.     

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.     

Compression and Decompression Waves in Steam-Water System

The ratio of the γactivities per fission from fission products of ²³⁹Pu and ²³⁵U, and its time dependence were measured by double fission chamber technique. The γ-activity from the fission products of ²³⁹Pu fission was lower than the corresponding activity relevant to ²³⁵U fissions. The ratio varied with the cooling time allowed after irradition.

This ratio was applied to power distribution measurements by γ-scanning method in multi-region cores composed of PuO₂-UO₂ and UO₂ fuels. To obtain the relative power, the measured γ-activities from the fission products in the fuel rods were corrected for the difference between the γ-activities per fission from the fission products.     

A least-squares fit of thermal data for fissile nuclei

A consistent set of 2200 m s^?1 neutron cross sections, Westcott g-factors and fission-neutron yields have been developed and evaluated for ²³³U, ²³⁵U, ²³⁹Puand ²⁴¹Pu. Auxiliary parameters such as ²³³U, ²³⁴U and ²³⁹Pu half-lives and ²⁵²Cf $\text{v}\text{?}$ values are also presented.All available relevant experimental data on fission, absorption, capture, total and scattering cross sections were revised using the most up-to-date reported standards. The latest Cf $\text{v}\text{?}$ data as well as the corrected $\text{v}\text{?}{}_{\mathrm{<mtext>p</mtext>}}$ (prompt) ratios of the above four nuclei were included in the study. The half-lives of the U and Pu nuclides have been reevaluated and were used to reassess experimental data which depended on the α-counting technique. A least-squares fitting program (lsf) was used to obtain a best overall fit and provide an estimate of the sensitivity of the parameters to the quoted uncertainties in experimental data.The recommended fissile thermal parameters listed in Table 36 now constitute a self-consistent set. Problems encountered in earlier evaluations have been found not to be significant in casting any doubts on the consistency of certain types of input data (such as $\text{v}\text{?}$ measurements and observations made in pile and Maxwellian neutron spectra) with the main body of input data. The least-squares results are presented and comparison has been made with ENDF/B-V and other evaluations.     

Statistical error in calculating the nuclide composition of low-burnup fuel

The error arising in the change of the ²³⁵U and ²³⁹Pu concentrations as a result of the statistical error in the microscopic cross sections during a computational fuel-run simulation with the MCU and MCNP programs is investigated. The analysis is limited to the thermal neutron spectrum and low fuel burnup. A simplified model simulating a fuel-run calculation using MCU and MCNP type statistical programs is constructed. This model is used to analyze for a commercial uranium-graphite reactor the effect of the rate of recalculation of and the statistical error in the microscopic cross sections over a run on the calculation of the ²³⁵U and ²³⁹Pu concentrations. The results show that the influence of the statistical error on the computed ²³⁵U and ²³⁹Pu concentration is negligible even with 10⁵ neutron histories in the statistical computational sample over a run.__________Translated from Atomnaya Énergiya, Vol. 98, No. 2, pp. 91–97, February, 2005.     

Calculation of Beta-Ray Spectra from Individual and Aggregate Fission Products

The β-ray spectra of individual fission products were calculated by using the β-decay data assuming every β-decay to be allowed transition. For the nuclides without measured decay data the β-feeding function was evaluated with the gross theory of β-decay and the β-ray spectrum was calculated from the function. The measured decay data were also supplemented with the data calculated by the gross theory for the excitation energy range above the highest measured excitation energy level. The β-ray spectra from aggregate fission products after a burst fission were calculated by using the β-ray spectrum and the atom number of each fission product nuclide and they were compared with the ones measured for thermal neutron induced fission of ²³⁵U, ²³⁹Pu and ²⁴¹Pu at Oak Ridge National Laboratory. The spectrum calculations showed excellent agreement with the measured data at shorter cooling times than 10s when many short-lived nuclides without measured decay data contributed considerably to the spectrum.     

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.     

An absolute determination of the U fission cross section at 964 keV

An absolute measurement of the ²³⁵U fission cross section has been carried out using a ²⁴Na---Be photoneutron source with median neutron energy of 964 keV. A symmetric two-foil experiment was set up to measure the fission rate in a low-albedo laboratory, and variations in the source-to-foil spacing used to determine the room background. Fission fragments passing through a limited solid angle aperture were recorded from each foil by solid state tracketch techniques. The photoneutron source was calibrated after each run using the manganese bath method and the secondary national standard source NBS-II. A computed neutron source spectrum with 32 keV FWHM was derived by the Monte Carlo method and used in reducing the data to a cross section at 964 keV. The final value of 1.21 ± 0.025 barns is absolute in that, except for small corrections, its determination was independent of any other cross section data.     

Novel methodology for the quantitative assay of fissile materials using temporal and spectral β-delayed γ-ray signatures

An analytical model for the generation of β-delayed γ-ray spectra following thermal-neutron-induced fission of mixed samples of ²³⁵U and ²³⁹Pu is presented. Using an energy-dependent figure-of-merit to designate the spectral regions employed in the assay, the unique temporal β-delayed γ-ray signatures are utilized to determine the fraction of ²³⁹Pu in a mixed U-Pu sample. By evaluating the β-delayed γ-ray temporal signatures of both ²³⁵U and ²³⁹Pu within a 3 keV energy bin, traditional sources of systematic uncertainty in quantitative assay using β-delayed γ-ray signals, such as self-attenuation of the sample and energy-dependent γ-ray detection efficiency, are significantly reduced. The effects of the time-dependent Compton-continuum and growth of longer-lived nuclides on the quantitative assessment are explored. This methodology represents a promising extension of the conventional means of analysis for quantitative assay of fissile materials using β-delayed γ-ray signatures.