Fission Of 239Pu Research Articles

Improved predictions of reactor antineutrino spectra

We report new calculations of reactor antineutrino spectra including the latest information from nuclear databases and a detailed error budget. The first part of this work is the so-called ab initio approach where the total antineutrino spectrum is built from the sum of all beta-branches of all fission products predicted by an evolution code. Systematic effects and missing information in nuclear databases lead to final relative uncertainties in the 10 to 20% range. A prediction of the antineutrino spectrum associated with the fission of 238U is given based on this ab initio method. For the dominant isotopes 235U and 239Pu, we developed a more accurate approach combining information from nuclear databases and reference electron spectra associated with the fission of 235U, 239Pu and 241Pu, measured at ILL in the 80's. We show how the anchor point of the measured total beta-spectra can be used to suppress the uncertainty in nuclear databases while taking advantage of all the information they contain. We provide new reference antineutrino spectra for 235U, 239Pu and 241Pu isotopes in the 2-8 MeV range. While the shapes of the spectra and their uncertainties are comparable to that of the previous analysis of the ILL data, the normalization is shifted by about +3% on average. In the perspective of the re-analysis of past experiments and direct use of these results by upcoming oscillation experiments, we discuss the various sources of errors and their correlations as well as the corrections induced by off equilibrium effects.

Experimental Verification of the Fission Chamber Gamma Signal Suppression by the Campbelling Mode

For the on-line monitoring of high fast neutron fluxes in the presence of a strong thermal neutron component, SCK·CEN and CEA are jointly developing a Fast Neutron Detector System, based on <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">242</sup> Pu fission chambers as sensors and including dedicated electronics and data processing systems. Irradiation tests in the BR2 reactor of <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">242</sup> Pu fission chambers operating in current mode showed that in typical MTR conditions the fission chamber currents are dominated by the gamma contribution. In order to reduce the gamma contribution to the signal, it was proposed to use the fission chambers in Campbelling mode. An irradiation experiment in the BR2 reactor with a <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">242</sup> Pu and a <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">235</sup> U fission chamber, both equipped with a suitable cable for measurements in Campbelling mode, proved the effectiveness of the suppression of the gamma-induced signal component by the Campbelling mode: gamma contribution reduction factors of 26 for the <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">235</sup> U fission chamber and more than 80 for the <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">242</sup> Pu fission chamber were obtained. The experimental data also prove that photofission contributions are negligibly small. Consequently, in typical MTR conditions the gamma contribution to the fission chamber Campbelling signal can be neglected.

Mass distribution in the bremsstrahlung-induced fission of 232Th, 238U and 240Pu

The yields of various fission products in the 10 MeV bremsstrahlung-induced fission of 232Th, 238U and 240Pu were determined using a recoil catcher and off-line γ-ray spectrometric techniques. From the yield data, mass yield distributions were obtained using charge distribution corrections. The higher yields of fission products around mass numbers 133–135, 138–140, 143–145 and their complementary products in the neutron and bremsstrahlung-induced fission of 232Th, 238U and 240Pu were interpreted based on nuclear structure effects. From the mass yield distribution, the peak-to-valley ( P / V ) ratio was also obtained for the above fissioning systems. The present data, along with data from the literature on different bremsstrahlung- and mono-energetic neutron-induced fissions of 232Th and 238U are interpreted to examine the influence of excitation energy on the peak to valley ratio. For the same compound nucleus 240Pu ⁎, the data in the 10–30 MeV bremsstrahlung-induced fission of 240Pu were compared with similar data of thermal to 14 MeV neutron-induced fission of 239Pu and the spontaneous fission of 240Pu to examine the role of excitation energy due to bremsstrahlung radiation and mono-energetic neutrons.

Cross Sections for 239Pu(n, f) and 241Pu(n, f) in the Range En = 0.01 eV to 200 MeV

The 239Pu and 241Pu neutron-induced fission cross sections have been measured from subthermal energies to 200 MeV. These measurements are part of a campaign to measure fission cross sections with high precision in support of advanced fast reactor technology. Plutonium-241 is the most active target measured in this program to date, with a half-life of 14.4 yr. The results for 239Pu are in good agreement with previous experiments and add new information to the limited knowledge on the fission cross section above 30 MeV. Discrepancies of up to 30% between the evaluations and the experimental data for 241Pu are found in the fast region, which is of particular importance for fast spectrum reactor technology, and a reevaluation of the fission cross section for this isotope is recommended.

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.

Reactivity Effects of Differences between JEFF-3.1 and ENDF/B-VI.8 in Analysis of Six MASURCA Cores of the R-Z Program

Reactivity contributions of differences between JEFF-3.1 and ENDF/B-VI.8 were analyzed for six early MASURCA cores of the R-Z program using ERANOS 2.1. These cores were designed such that their neutron spectra would emulate that of an oxide-fueled sodium-cooled fast reactor, some containing enriched uranium and others containing depleted uranium and plutonium. Effects of modeling assumptions and solution methods both in ECCO lattice calculations and in BISTRO Sn flux solutions were first evaluated using JEFF-3.1 cross-section libraries. Comparisons were made between calculated and measured values for reactivity and several spectral indices. Reactivity effects of differences between JEFF-3.1 and ENDF/B-VI.8 were also quantified using perturbation theory analysis. The most important nuclide with respect to reactivity differences between cross-section libraries was 23Na, primarily a result of differences in the angular dependence of elastic scattering, which is more forward peaked in ENDF/B-VI.8 than in JEFF-3.1. Differences in 23Na inelastic scattering cross sections between libraries also generated significant differences in reactivity, more due to the differences in magnitude of the cross sections than to the angular dependence. The nuclide 238U was also found to be important with regard to reactivity differences between the two libraries mostly due to a large effect of inelastic scattering differences and two smaller effects of elastic scattering and fission cross sections. In the cores that contained plutonium, 239Pu fission cross-section differences contributed significantly to the reactivity differences between libraries.

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.

Using low resolution gamma detectors to detect and differentiate 239Pu and 235U fissions

When 239Pu and 235U undergo thermal neutron-induced fission, both produce significant numbers of beta-delayed gamma rays with energies in the several megaelectron volt range. Experiments using high energy-resolution germanium detectors have shown that it is possible to distinguish the fission of 239Pu from that of 235U. It is desirable to detect the presence of 235U or 239Pu using detectors that are less expensive and more rugged than high purity germanium detectors. To this end we demonstrate how differences in the energy spectrum and decay rates of the beta-delayed gamma rays can be used to identify 239Pu and 235U using low resolution plastic and liquid scintillator detectors. Experimental data are used to identify differences in the spectra and also to test the identification algorithms. Results to date are very promising.

Measurement and analysis of fission neutron output spectra from 1-8 MeV neutron-induced fission of 235U and 239Pu

Measurement and analysis of fission neutron output spectra from 1-8 MeV neutron-induced fission of 235U and 239Pu

Fission-product gamma-ray line pairs sensitive to fissile material and neutron energy

The beta-delayed gamma-ray spectra from the fission of 235U, 238U, and 239Pu by thermal and near-14-MeV neutrons have been measured for delay times ranging from 1min to 14h. Spectra at all delay times contain sets of prominent gamma-ray lines with intensity ratios that identify the fissile material and distinguish between fission induced by low-energy or high-energy neutrons.

Recent applications of nuclear tracks in solids

With partial support from an NSF US–China exchange program, five Chinese scientists collaborated with me for one or more years during the 1980s and 1990s. This review begins with a discussion of some of the research they did while at Berkeley. That is followed by brief reviews of recent applications of the nuclear track technique to nuclear fusion; molecular identification with nanopores; ion track filters for imaging X-ray astronomy; magnetic studies with nanowires; polymeric nanowires; microbiology with Nuclepore filters; radon and neutron dosimetry; and thermochronology with 238U and 244Pu fission tracks. My review provides a small sample of the interesting topics to which nuclear track techniques are now being applied.

Prediction Accuracy Improvement of Neutronic Characteristics of a Breeding Light Water Reactor Core by Extended Bias Factor Methods with Use of FCA-XXII-1 Critical Experiments

Two extended bias factor methods, the LC and PE methods, were applied to the prediction accuracy evaluation of neutronic characteristics of a breeding light water reactor, using data of FCA-XXII-1 critical experiments, in order to investigate the features and effectiveness of these methods on the basis of an actual core design and existing experimental results. The present study confirms the following features of these methods. Both the LC and PE methods can improve the prediction accuracy the most when all the experimental results are used. The prediction accuracy improvement is achieved mainly by reducing uncertainty due to errors in cross sections. This is done by realizing a profile of sensitivity coefficients closer to that of the target core and suppressing the influence of errors in experiments and experimental analysis methods. The PE method always improves the prediction accuracy with the use of any combination of experimental results. It is always superior to the LC method in the improvement of the prediction accuracy. Concerning the effectiveness of using the extended bias factor methods with the data of FCA XXII-1 critical experiments, it is concluded that the experimental results regarding multiplication factor are more effective than the other experimental results, namely, reaction rate ratios of 238U capture to 239Pu fission (C28/F49) and void reactivity, in reducing prediction uncertainties of all the neutronic characteristics of the target core investigated: the multiplication factor, the C28/F49, and the void reactivity of the target core. This is due to the fact that the extended bias factor methods cannot fully utilize the potential that these experimental results have for the reduction of the uncertainties due to the errors in cross sections because of their strong correlations to the target core characteristics. This failure is due to large errors in the experiments and/or the experimental analysis methods.

Evaluation of Prediction Error Reduction for Breeding Ratio by Bias Factor Method with Use of Experimental Result on Basic Reaction Rate Ratio

The evaluation technique of prediction error reduction by the bias factor method is extended to the cases where the neutronic property of a target core is related to the properties in critical experiments. This extended procedure can utilize several experimental results, although the conventional procedure utilizes only one experimental result. Calculations were performed where the measured property is the reaction rate ratio of 238U capture to 239Pu fission (C28/F49) in FCA XXII-1 cores and the target core property is C28/F49 or breeding ratio in a breeding light water reactor (BLWR). The prediction error of the breeding ratio in BLWR is reduced from 0.57 to 0.43%, whereas the error of C28/F49 is reduced from 0.72 to 0.19%.

Evaluation of Prediction Error Reduction for Breeding Ratio by Bias Factor Method with Use of Experimental Result on Basic Reaction Rate Ratio

The evaluation technique of prediction error reduction by the bias factor method is extended to the cases where the neutronic property of a target core is related to the properties in critical experiments. This extended procedure can utilize several experimental results, although the conventional procedure utilizes only one experimental result. Calculations were performed where the measured property is the reaction rate ratio of 238U capture to 239Pu fission (C28/F49) in FCA XXII-1 cores and the target core property is C28/F49 or breeding ratio in a breeding light water reactor (BLWR). The prediction error of the breeding ratio in BLWR is reduced from 0.57 to 0.43%, whereas the error of C28/F49 is reduced from 0.72 to 0.19%.

Prediction Accuracy Improvement of Neutronic Characteristics of a Breeding Light Water Reactor Core by Extended Bias Factor Methods with Use of FCA-XXII-1 Critical Experiments

Two extended bias factor methods, the LC and PE methods, were applied to the prediction accuracy evaluation of neutronic characteristics of a breeding light water reactor, using data of FCA-XXII-1 critical experiments, in order to investigate the features and effectiveness of these methods on the basis of an actual core design and existing experimental results. The present study confirms the following features of these methods. Both the LC and PE methods can improve the prediction accuracy the most when all the experimental results are used. The prediction accuracy improvement is achieved mainly by reducing uncertainty due to errors in cross sections. This is done by realizing a profile of sensitivity coefficients closer to that of the target core and suppressing the influence of errors in experiments and experimental analysis methods. The PE method always improves the prediction accuracy with the use of any combination of experimental results. It is always superior to the LC method in the improvement of the prediction accuracy. Concerning the effectiveness of using the extended bias factor methods with the data of FCA XXII-1 critical experiments, it is concluded that the experimental results regarding multiplication factor are more effective than the other experimental results, namely, reaction rate ratios of 238U capture to 239Pu fission (C28/F49) and void reactivity, in reducing prediction uncertainties of all the neutronic characteristics of the target core investigated: the multiplication factor, the C28/F49, and the void reactivity of the target core. This is due to the fact that the extended bias factor methods cannot fully utilize the potential that these experimental results have for the reduction of the uncertainties due to the errors in cross sections because of their strong correlations to the target core characteristics. This failure is due to large errors in the experiments and/or the experimental analysis methods.

Pu–Xe, U–Xe, U–Pb chronology and isotope systematics of ancient zircons from Western Australia

The presence of xenon isotopes from in-situ spontaneous fission of short-lived 244Pu has been confirmed in a suite of 16 Hadean detrital zircons from Western Australia. In order to investigate the effects of xenon loss on estimates of the inferred Pu/U ratio we have irradiated the zircons with thermal neutrons to generate Xe from 235U neutron fission. 131Xe/ 134Xe and 132Xe/ 134Xe ratios have been used to calculate the relative contributions from spontaneous fission of 244Pu and 238U and neutron fission of 235U and hence compare nominal Pu/U ratios and xenon retention ages. U–Xe ages are typically lower than the Pb–Pb ages, indicating that xenon loss is common. We show how the ternary mixing diagram can be used to place constraints on the timing and extent of this loss and to generate a corresponding Pu–U–Xe isochron. Although the zircons investigated in this study were extracted from the same metasedimentary unit, the timing of xenon loss is variable. This suggests that the loss may be the result of variable degrees of metamictization from grain to grain. Inferred (Pu/U) o ratios show a general decrease with the discordancy between Pb–Pb and U–Xe ages. For the least discordant samples we infer (Pu/U) o ∼ 0.008 which is close to the widely adopted chondritic value. While we cannot completely exclude the effects of Pu/U fractionation in magmatic and other processes between formation of the Earth and crystallisation of the zircons we conclude that they have been relatively small (< factor 2?) and reflect the compatible behaviour of Pu and U in the 4+ valence state.

A method and setup for studying the energy dependence of delayed neutron characteristics in nuclear fission induced by neutrons from the T(p, n), D(d, n), and T(d, n) reactions

A method and a setup based on this method are described, with which it has for the first time become possible to measure the energy dependence of the absolute total and relative yields of delayed neutrons and the half-fives of their precursors from neutron-induced fission of heavy nuclei in the course of one experiment. T(p, n)3He, D(d, n)3He, and T(d, n)4He nuclear reactions induced by high-energy charged-particle beams from the KΓ-2.5 electrostatic accelerator at the Institute of Physics and Power Engineering are used as sources of monoenergetic neutrons. The measured total delayed neutron yields from neutron-induced fission of 233U and 239Pu nuclei in the energy range of 0.37–4.7 MeV, as well as the relative yields of delayed neutrons and the half-lives of their precursors from neutron-induced fission of 239Pu in the range of 15.8–17.9 MeV, are presented as an illustration of the method. The uncertainties of the data obtained by means of this method are shown to be significantly lower than the uncertainties of similar data measured using other techniques.

Studies on the Feasibility of the LWRs Waste-Thorium In-Core Fuel Cycle in the Gas Turbine-Modular Helium Reactor

The capability to operate on LWRs waste constitutes one of the major benefits of the Gas Turbine-Modular Helium Reactor; in this paper, it has been evaluated the possibility to incinerate the LWRs waste and to simultaneously breed fissile 233U by fertile thorium. Since a mixture of pure 239Pu-thorium has shown a quite poor neutron economy, the LWRs waste-thorium fuel performance has been also tested when plutonium and thorium are allocated in different TRISO particles. More precisely, when fissile and fertile actinides share the same TRISO kernel, the resonance at 0.29 eV of the fission and capture microscopic cross sections of 239Pu diminishes also the absorption rate of fertile 232Th and thus it degrades the breeding process. Consequently, in the present studies, two different types of fuel have been utilized: the Driver Fuel, made of LWRs waste, and the Transmutation Fuel, made of fertile thorium. Since, in the thermal neutron energy range, the microscopic capture cross section of 232Th is about 80-100 times smaller than the fission one of 239Pu, setting thorium in particles with a large kernel and LWRs waste in particles with a small one makes the volume integrated reaction rates better equilibrated. At the light of the above consideration, which drives to load as much thorium as possible, for the Transmutation Fuel they have been selected the JAERI TRISO particles packed 40%; whereas, for the Driver Fuel they have been tested different packing fractions and kernel radii. Since no configuration allowed the reactor to work, the above procedure has been repeated when fertile particles are packed 20%; the latter choice permits over one year of operation, but the build up of 233U represents only a small fraction of the depleted 239Pu. Finally, the previous configuration has been also investigated when the fertile and fissile fuels share the same kernel or when the fertile fuel axially alternates with the fissile one.

Neutrinos and Non-proliferation in Europe

Triggered by the demand of the IAEA, neutrino physicists in Europe involved with the Double Chooz experiment are studying the potential of neutrino detection to monitor nuclear reactors. In particular a new set of experiments at the ILL is planned to improve the knowledge of the neutrino spectrum emitted in the fission of 235U and 239Pu.

Sensitivity Analysis based on Transport Theory

To estimate the uncertainty of the neutronic parameters it is required to calculate sensitivity coefficients to cross section changes. The sensitivity coefficients are usually calculated based on the diffusion theory. However, the accuracy of the sensitivity coefficients becomes doubtful because of the use of diffusion theory. The purpose of this paper is to derive the generalized perturbation theory using the transport theory, and to evaluate transport effect on the sensitivity coefficients. Sensitivity calculations were performed in 70 energy groups. The calculated sensitivity coefficients were compared with those obtained from direct transport calculations to check the accuracy of the present method, and those obtained from the generalized perturbation calculation based on the diffusion theory to investigate the transport effect on sensitivity coefficients. From the comparison, the difference was remarkably large for the neutronic parameters such as reaction rate distribution, reaction rate ratio, and reactivity worth. For example, the sensitivity coefficient of the 238U capture to 239Pu fission rate ratio in the lower core region of the under-moderated LWR core was large in the energy range 17.6–22.6 eV, and the sensitivity to 238U capture cross section calculated by the transport theory was 20% larger than that by the diffusion theory.