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.     

239Pu(nth,f)短寿命裂变产物产额测量

采用γ能谱相对测量方法,以97Zr为内标参考核素,完成了239Pu(nth,f)短寿命裂变产物88 Rb、95 Y、101 Mo、101 Tc、138 Csg、142 La等核素的累积产额测量。实验测得88 Rb、95 Y、101 Mo、101 Tc、138 Csg、142 La的累积产额数据分别为(1.32±0.05)%、(4.69±0.22)%、(6.13±0.32)%、(6.10±0.22)%、(6.24±0.24)%、(4.74±0.17)%。对高纯锗探测器中高能端效率刻度、样品封装与辐照、γ谱测量几何条件设计、γ谱测量与数据分析进行了研究。实验测量裂变产物核素分别位于非对称裂变质量分布双驼峰曲线轻峰的左侧、中部和右侧,重峰的中部与右侧。     

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.     

Correction to Measurement of Fission Rates of ^235U and ^238u

For the enrichment of fissile nuclide in detecting foils is not so high, the counts of a certain gamma energy （1 596 keV） measured after irradiation would come from the same fission product of the each nuclide （^235U and ^238U）. Consequently the measured value of distribution of fission rate is not precise due to the blending of the nuclide.     

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.     

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.     

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.     

Theoretical estimate of the fast-neutron fission cross sections of235,238U

Institute of Physics and Power Engineering. Translated from Atomnaya énergiya, Vol. 76, No. 3, pp. 201–205, March, 1994.     

Calculation and Evaluation of Neutron-Induced Reactions on ^60Ni Below 150 MeV

Based on the experimental data of total, nonelastic, elastic cross section and elastic scattering angular distributions for n＋^60Ni reactions, a set of neutron optical model potential parameters is obtained in the region of incident neutron energy from 0.456-150 MeV. The reaction cross sections, angular distributions, energy spectra, gamma-ray production cross sections, gamma-ray production energy spectra, are calculated and evaluated by optical model,     

Discrepancies in Estimates of the Number of Secondary Fission Neutrons for 239Pu

The energy dependence of the total number of secondary neutrons v ²³⁹Pu is analyzed. The results of variations of different national libraries are compared: ENDF/B-5, ENDF/B-6.5, JEF-2.2, JENDL-3.2, CENDL-2, and BROND-2. It is concluded that the required accuracy has not been attained for the entire energy range.     

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.     

Calculation of the Spin and Energy Dependence of the Fission Width for U-235 and Pu-239

Direct preparation of Na₂ ³⁵SO₄ from neutron irradiated KC1 has been studied. Irradiated KCl was dissolved in water and passed through a H⁺-form cation exchanger column to remove K⁺ ions. The effluent, after the addition of Na₂SO₄ carrier, was neutralized with NaOH. From this solution was separated solid Na₂ ³⁵SO₄ by the addition of ethanol, leaving NaCl in the solution.

Certain factors such as ethanol concentration and pH of the system, from which labeled sulfate was precipitated, as well as the amount of carrier, were found to affect greatly the recovery of the labeled sulfate. Conditions that bring about sufficiently high recovery were determined, and an effective means was established to prevent contamination with ³²P, a byproduct formed by neutron irradiation of CI.

Similar procedures have been examined to prepare labeled potassium sulfate.     

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.     

Determination of 240Pu and 239Pu Ratio by Spontaneous Fission Counting

The ²⁴⁰Pu content in Pu samples was determined by spontaneous fission measurement with a p-n junction silicon detector. Prolonged continuous measurements were run, extending over several hundred hours, using a transistorized counting system to obtain the required stability during such long periods. The results indicated that this method is applicable to the determination of ²⁴⁰Pu/²²³⁹Pu ratio.     

0.57、1.0和1.5MeV中子诱发~(235)U裂变时~(95)Zr、~(140)Ba和~(147)Nd产额的绝对测量

用HPGeγ能谱法绝对测量0.57、1.0和1.5 MeV中子诱发235U裂变的裂变产物95Zr、140Ba和147Nd的产额。P型同轴HPGe探测器的体积为110 cm3,对60Co 1 332.5 keVγ射线的分辨率为1.8 keV。使用双裂变电离室监测裂变率,应用SPAN程序解析γ谱。对影响产额测量的主要因素进行了分析,并用MCNPIVB程序模拟计算了非主中子的影响。实验测得0.57 MeV中子诱发时,95Zr、140Ba、147Nd的产额分别为(6.64±0.21)%、(6.37±0.22)%、(2.37±0.08)%;1.0 MeV中子诱发时,95Zr、140Ba、147Nd的产额分别为(6.49±0.19)%、(6.35±0.18)%、(2.25±0.07)%;1.5 MeV中子诱发时,95Zr、140Ba、147Nd的产额分别为(6.59±0.19)%、(6.25±0.18)%、(2.37±0.07)%。对95Zr、140Ba、147Nd的产额随入射中子能量的变化趋势作了简要分析与讨论。