Simple Estimation of Fission Yields with Selective Channel Scission Model

The mass distributions of fission yields for neutron-induced fissions of ²³³U, ²³⁶U, ²³⁸U, ²³⁷Np, ²⁴⁰Pu, and ²⁴²Pu, spontaneous fissions of²⁴¹Am and ²⁴³Am, and fissions of ¹⁰⁶Pd and ¹⁹⁷Au were calculated by the selective channel scission (SCS) model with simple assumptions. The channel-dependent fission barriers were deduced by using shapes of fission fragments in the ground states. The present method makes it possible to estimate fission yields for a wide range of fissionable nuclei without adjustable parameters, although there exist discrepancies between the fission yields calculated by the SCS model and the data of JENDL-3.3 in the mass regions of A = 80–95 and A = 135–150.     

Summation calculations of fission-product decay heat, their uncertainties and their applications to a fast breeder reactor

Prediction of unmeasured fission parameters is described concerning mass yields of fission products, average total kinetic energies of fission fragments and average numbers of prompt neutrons in neutron-induced fission of several nuclides related to nuclear reactors. Reliability of predicting unmeasured mass yields of the products from measured mass yields of the primary fragments is discussed. The predicted values of the mass yields of the products are presented for the fissions of ^233,235U, ^239,241Pu by thermal neutrons, of ²³²Th by 1.38-MeV neutrons and of ²³¹Pa, ²³⁸U and ²³⁷Np by fission-spectrum neutrons.A semi-empirical method based on the statistical theory of nuclear fission is extended to unmeasured distributions of the fragments. The mass yields of the products, the average total kinetic energies of the fragments and the average numbers of prompt neutrons are predicted for the fissions of ²³²U and ²³⁸Pu by thermal neutrons, and of ^234,236U, ²⁴⁰Pu and ²⁴²Pu by 2-MeV neutrons. Errors in the predicted values are also discussed.     

Measurement of Fast Neutron Induced Fission Cross Section Ratios of Pu-240 and Pu-242 Relative to U-235

Fission cross section ratios of ²⁴⁰Pu and ²⁴²Pu relative to ²³⁵U were measured by using the 4.5 MV Dynamitron accelerator of Tohoku University. The measurement using mono-energetic neutrons was performed in the neutron energy range of 0.6–7 MeV with the time-of-flight method. Prior to the measurement, a fast timing back-to-back fission chamber was developed with good time resolution to reduce the backgrounds due to α-particles and spontaneous fissions. Furthermore, we took account of the effect of the nonuniformity of fission sample thickness for accurate determination of fission cross section ratio. The uncertainty was estimated by analyzing the correlation between the error sources. The correlation matrix between the measured data was given. The overall uncertainty of the present results is about 2%. For both nuclides, the present results agree well with those by Meadows and by Kuprijanov et al. The JENDL-3 evaluation generally has good agreement with the present results. However, the evaluated data are slightly higher around 1 MeV and lower above 6 MeV than the present results.     

Role of Effective Distance in the Fission Mechanism Study by the Double-energy Measurement for Uranium Isotopes

Fission product kinetic energies were measured by the double-energy method for thermal-neutron fission of ^235,233U and proton-induced fission of ²³⁸U at the 15.8-MeV excitation. From the obtained energy-mass correlation data, the kinetic-energy distribution was constructed from each mass bin to evaluate the first moment of the kinetic energy for a given fragment mass. The resulting kinetic energy was then converted to the effective distance between the charge centers at the moment of scission. The effective distances deduced for the proton-induced fission was concluded to be classified into two constant values, one for asymmetric and the other for symmetric mode, irrespective of the mass though an additional component was further extracted in the asymmetric mass region. This indicates that the fission takes place via two well-defined saddles, followed by the random neck rupture. On the contrary, the effective distances obtained for thermal-neutron induced fission turned out to lie along the contour line at the same level as the equilibrium deformation in the two-dimensional potential map. This strongly suggests that it is essentially a barrier-penetrating type of fission rather than the over-barrier fission.     

Inherent Protection of Plutonium by Doping Minor Actinide in Thermal Neutron Spectra

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% ²³⁵U and 20% ²³⁵U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, ²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% ²³⁵U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties.     

Multi-parametric Measurement of Prompt Neutrons and Fission Fragments for 233U(n th,f)

The multiplicity and the energy of prompt neutrons from the fragments for ²³³U(n _th, f)were measured as functions of fragment mass and total kinetic energy. Average neutron energy against the fragment mass showed a nearly symmetric distribution about the half mass division with two valleys at 98 and 145 U. This shape formed a contrast with a saw-tooth distribution of the average neutron multiplicity. It indicates that the shell-effects, which are pronounced for the fragments having the proton number or neutron number close to the magic-number of 50 or 82, affected the neutron emission process. The slope of the neutron multiplicity with total kinetic energy depended on the fragment mass and showed the minimum at about 130 U. The obtained neutron data were applied to determine the total excitation energy of the system, and the resulting value in the typical asymmetric fission lied between 22 and 25 MeV. The excitation energy agreed with that determined by subtracting the total kinetic energy from the Q-value within 1MeV, thus satisfied the energy conservation. In the symmetric fission, where the mass yield was drastically suppressed, the total excitation energy is significantly large and reaches to about 40MeV: suggesting that fragment pairs are preferentially formed in a compact configuration at the scission point.     

Neutron Emission,Mass Distribution and Energetics in 14.5 MeV-Neutron Induced Fission of Uranium-238

A study is performed on 14.5 MeV-neutron induced fission of ²³⁸U by means of three-parameter experiment in which the energies of both fragments and the time-of-flight of one fragment are measured. A mosaic-arrayed surface barrier detector of large sensitive area is used at the remote end of a flight tube. The pre- and post-neutron-emission fragment mass distributions are obtained, together with the average total kinetic energy of fragment as a function of its mass. The average number of neutrons emitted from an individual fragment and the average total number of emitted neutrons are also derived as a function of fragment mass. The results agree well with those calculated by the method developed in our laboratory for medium-excitation fission. The average number of emitted neutrons and the mass distribution of fission fragment are derived for the respective reactions of first-, second- and third-chance fission.     

Estimation of Neutron Yield from Individual Fragment in Medium-Excitation Fission

A method of calculation is described to estimate the average number of neutrons emitted per fragment in medium-excitation fission from published experimental data on neutron emission in thermal-neutron induced fission, average total kinetic energy as a function of fragment mass and mass yield in low- and medium-excitation fission reactions. Use is made of a relation of fragment excitation energy with internal excitation and deformation energies, and the difference in kinetic energy between the fission reactions at two-excitation energies. A tentative calculation is made for the fission of ²³⁸U induced by 12 MeV protons. The results are in good agreement with experimental data.

The method developed in the present work may make it possible to predict the average number of neutrons emitted from individual fragment in medium-excitation fission which has not yet been measured experimentally.     

Measurement of homogeneity of U and Pu isotopes in PWR spent fuel powder mixture

We present a feasibility study of the homogenization of pressurized water reactor spent nuclear fuel (SNF) powder through a mechanical mixing process. Because burn-up of the SNF depends on the position in the SNF assembly, concentrations of uranium, plutonium, and fission products are distributed differently according to the burn-up profile. The heterogeneity of the material elements affects the selection of a representative sample for quantitative analysis. Homogenization process improvement to reduce the sampling error is thus required to precisely determine the amount of uranium, plutonium, and fission products in the SNF. In this study, fine powders (<70 μm) extracted from one SNF rod were mixed, and the degree of homogenization was determined as a function of the mixing time indicating the relative standard deviations of the ¹³⁴Cs/¹³⁷Cs, Pu, and U isotope ratio measurements.     

Estimation of 239Pu independent and cumulative fission product yields from the chain yield data using a Bayesian technique

The independent and cumulative fission product yields (FPYs) are obtained by using the Bayesian technique based on the evaluated mass chain yield, where required constraints such as the normalization can be straightforwardly included. We apply this technique to the ²³⁹Pu FPY data at neutron incident energies of 0.5, 2.0, and 14 MeV, where the most updated mass chain yield ENDF/B-VII.1 data are available. The obtained yield data are compared with the evaluated values by England and Rider in ENDF/B-VI, and differences from their values are investigated. We show that the modern decay data used, such as branching ratios to ground and metastable states, cause differences in the evaluated individual and cumulative fission yields.     

Approach for the fission fragment total kinetic energy TKE(A) calculation: Application to prompt neutron emission models

A simple approach for the calculation of the fission fragment total kinetic energy, TKE(A), based on the electrostatic repulsion between the fragments connected by a neck in the pre-scission configuration is described. The calculated TKE(A) is obtained in good agreement with the experimental data for many fissioning systems, such as ^233,235U(n_th, f), ²³⁹Pu(n_th, f), ²³⁷Np(n, f), ²⁴²Pu(SF), with minor adjustment of only one parameter. Due to the fact that the present approach can provide with enough trust TKE(A) distributions for fissioning systems for which experimental TKE(A) data do not exist, the possibilities to use the refined Point by Point model of prompt neutron emission can be considerably extended.     

Calculation of Prompt Fission Product Average Cross Sections for Neutron-Induced Fission of 235U and 239Pu

Yield-weighted average cross sections of neutron radiative capture, (n,2n), and (n,3n) reactions over prompt fission products (FPs) from ²³⁵U and ²³⁹Pu are calculated. The prompt fission production yields are taken from the ENDF/B-VII.0 library. The FPs for each fissile material exist over a range of approximately 1000 neutron-rich nuclides. Several nuclear reaction codes are utilized for calculating the cross sections on each individual fission product—EMPIRE-2.19, TALYS-1.0, GNASH, and CoH. The influence of the FP isomers on the average cross sections is examined with TALYS. We investigate the dependence of the average cross sections on the number of FPs taken for averaging. It is shown that the average capture cross section is much more sensitive to the number of FPs included, compared with the (n,2n) and (n,3n) reactions. An intercomparison of the calculated cross sections with the different reaction codes is carried out. In the capture reaction, EMPIRE predicted lower cross section than TALYS and CoH owing to different default assumptions used in the γ-ray strength function modeling. Moreover, the preequilibrium models implemented in each code give different predictions for the neutron-emission reactions, although the differences are relatively small. We also discuss a difference between the macroscopic and microscopic calculation options in TALYS for the pre-equilibrium model, optical potential model, and γ-ray strength function. The predictive capability of the reaction codes for the capture reaction is examined by comparing their calculations with the ENDF data, which are based on measurements. Compared with the historic Foster and Arthur's evaluation, our new (n,2n) predictions are similar, although our capture predictions are almost an order of magnitude higher. Recommended cross sections for use in applications have been tabulated in ENDF-formatted files.     

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.     

Product yields for the photofission of ~(232)Th,~(234,238)U,~(237)Np,and ~(239,240,242)Pu actinides at various incident photon energies

Photofission fragments mass yield for~(232)Th,~(234;238) U,~(237) Np, and~(239;240;242) Pu isotopes are investigated.The calculations are done using a developed approach based on Gorodisskiy's phenomenological formalism. The Gorodisskiy's method is developed to be applied for the neutron-induced fission. Here we revised it for application to photofission. The effect of emitted neutron prior to fission on the fission fragment mass yields has also been studied. The peak-to-valley ratio is extracted for the240 Pu isotope as a function of energy. Obtained results of the present formalism are compared with the available experimental data. Satisfactory agreement is achieved between the results of present approach and the experimental data.     

<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.     

Fission Product Model for BWR Lattice Calculation Code

A benchmark calculation of full fission product was performed for thermal reactor application using an isotope transmutation code DCHAIN based on 185 nuclides with revised nuclear data library. The fission product model for BWR lattice calculation was studied and tested with the benchmark results, and a model containing 45 explicit nuclides and one pseudo nuclide was selected as a reasonably best model to predict the burn up reactivity with high precision for practically all types of fuel and reactor operating conditions. The evaluated thermal cross section and resonance integral for the pseudo nuclide are σ_2,200 = 2.6b and.RI = 10.6b, combined with the pseudo fission yield values of 1.3898, 1.3233, 1.3675 and 1.2773 for fissions from ²³⁵U, ²³⁸U, ²³⁹Pu and ²⁴¹Pu, respectively. The present results are believed as equally applicable to PWR lattice calculation.     

Phenomenological Model for Estimating the Activity of Fission Products in Nuclear Fuel and Accidental Reactor Emissions

A phenomenological model is proposed for estimating the production of fission products with half-life ∼1 day and longer in the core of a thermal reactor. The model takes account of the fissioning of ²³⁵U, ²³⁹Pu, and ²⁴¹Pu by thermal neutrons and ²³⁸U by fission-spectrum neutrons. The fission rate of individual nuclides can contain exponential time dependences and power-law dependences of the form t, t ², and t ³. The model makes it possible to obtain in an analytic form the dependence of the specific mass and specific activity on the operating time of the reactor for most fission products which are of practical interest in radioecology. The calculations were performed for a RBMK-1000 reactor operating in a constant-power regime. In all cases, the analytic calculations agree with the numerical calculations.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 380–386, May 2005.     

Uranium Crystallization Test with Dissolver Solution of Irradiated Fuel

The crystallization process has been developed as a part of the advanced aqueous process, NEXT (New Extraction System for TRU recovery) for fast reactor (FR) cycle. In this process, a large part of U is separated from dissolver solution by crystallization as UO₂(NO₃)₂.6H₂O. The U crystallization test was carried out with real dissolver solution of irradiated FR fuel to investigate the influence of cooling rate on the crystal size and the behavior of fission product (FP) compared with that of Pu(IV). In regard to the influence of the cooling rate, it was confirmed that the crystal size was smaller as the cooling rate is faster. Although it was expectable that the decontamination performance was improved by diminishing the specific surface of the crystals, it was suggested that a large crystal produced by crystallization was not always high purity. Concerning the behavior of FPs, Eu behaved similarly to Pu(IV). Cs accompanied with U into the crystals under the condition in this test.     

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.