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.     

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 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.     

Fission cross-sections of U and Pu averaged over Cf neutron spectrum

A series of measurements have been carried out to derive values for the spectrum-averaged fission cross-section of ²³⁵U and ²³⁹Pu for ²⁵²Cf fission neutrons. Two nearly identical target foils were mounted on either side of a Cf source (10⁷ neutron/sec) in a compensated beam geometry. Fission fragments passing through limited solid angle apertures were recorded from each foil by solid-state track-etch techniques. The Cf neutron source strength was calibrated in manganese bath relative to the standard source NBS-II. Values of 1.215 ± 0.022 barn for ²³⁵U and 1.790 ± 0.041 barn for ²³⁹Pu were obtained for the fission cross-sections, corresponding to a ratio value of 1.473 ± 0.041.     

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.     

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.     

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.     

Relative Yield and Period of Individual Groups of Delayed Neutrons in 233U, 235U, and 239Pu Fission by Epithermal Neutrons

The relative yield of delayed neutrons and the half-life of their precursor nuclei in fissioning of ²³³U, ²³⁵U, and ²³⁹Pu by epithermal neutrons are measured on a setup which is based on a KG-2.5 electrostatic accelerator. The experimental samples are irradiated in a polyethylene cube, one face of which is irradiated by a neutron beam consisting of T(p, n)³He neutrons. A procedure is developed for averaging the group parameters of delayed neutrons from different series of measurements, taking account of their correlation. A comparative analysis of the data of this work and in ENDF/BVI in terms of the average half-life of the precursor nuclei of the delayed neutrons and using the dependence of the reactivity on the asymptotic period of the reactor is performed.It is shown that the data of this work agree, to within the uncertainty limits, with the recommended experimental data and differ substantially from the ENDF/BVI data.     

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.     

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.     

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.     

Thinking in Different Ways to Combine Fusion with Fission

The common goal of CTR, but in particular of ICF, is low yield-high gain. Fission triggered large TN explosive devices meet the second but not the first of these conditions. These devices depend on the rare isotopes U235, Pu239, or U233, but for them the fusion energy output greatly exceeds the output from fission, limiting the fallout. In thinking about different ways to combine fusion with fission, there are three questions: (1) Are there ways where both conditions can be met, and where the fallout from fission is small? (2) Can the conditions be met without the use of U235, Pu239, or U233, but with U238, Th232, and perhaps with the fission of light nuclei like B10 or Li6, the latter having no fallout? (3) Are there concepts for MF, combining fusion with fission, without U235, Pu239 or U233? In my talk I will present reasons why under the above stated conditions two things seem to be possible: (1) The greatly facilitated fast ignition of thermonuclear microexplosions with a small amount of U238 or Th232. (2) The greatly enhanced pulsed MF burn aided by the fission of light nuclei such as B10, but also of the U238 and Th232 and with a neutron moderator. In either one of these cases the burn is “autocatalytic” in the sense that neutron-induced nuclear reactions in a halo surrounding the fusion plasma drive thermomagnetic currents compressing and increasing its neutron production rate.     

<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.     

Benchmark Tests of Principal Fissile Nuclide Data in JENDL-3 with Simple Fast Critical Assemblies

Abstract

A series of benchmark tests was made to check the neutron nuclear data of main fissile nuclides (²³⁹Pu, ²³⁵U and ²³³U) of JENDL-3 for fast reactors. A total of nine critical assemblies were analyzed. They are assemblies of single material, high enrichment and simple geometry with small volume and therefore suitable for nuclear data testing. Criticality calculation was made by ANISN with S16P5 using the VITAMIN-J 175-energy-group. Discussions are made on keft, spectral indices at core center and leakage spectra.

From the study, a problem was pointed out relating to the interpolation of secondary-neutron energy distributions for threshold reactions near the threshold energy point adopted in the original JENDL-3 and its remedy was proposed. By the benchmark tests of thus JENDL-3 (JENDL-3.1), it was shown that integral experiments for ²³⁹Pu and ²³⁵U cores were reproduced quite satisfactorily. On the contrary, it was revealed that large deviations for ²³³U cores from the experiment were due to uncertainties of the fission spectrum and the inelastic scattering cross sections. In the present work, sensitivity of “a” parameter (level density parameter) of Madland-Nix's fission spectrum formula to the integral data was extensively studied. Some recommendations are made to improve JENDL-3.1.     

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.     

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.     

Use of γ Radiation from Short-Lived Fission Products for Detecting <Superscript>235</Superscript>U and <Superscript>239</Superscript>Pu

It is suggested that γ radiation with E _γ > 4900 keV from short-lived fission products produced by thermal neutrons be used to detect ²³⁵U and ²³⁹Pu in samples. A time regime is substantiated: 120 sec irradiation, 60 sec holding time, and 120 sec measurement time. The contribution of the reaction (n, p) on fast neutrons is studied.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 365–370, May 2005.     

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.     

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.