Thermal Fission Research Articles

Impact of material thicknesses on fission observables obtained with the FALSTAFF experimental setup

In the past years, the fission studies have been mainly focused on thermal fission because most of the current nuclear reactors work in this energy domain. With the development of GEN-IV reactor concepts, mainly working in the fast energy domain, new nuclear data are needed. The FALSTAFF spectrometer under development at CEA-Saclay, France, is a two-arm spectrometer which will provide mass yields before (2V method) and after (EV method) neutron evaporation and consequently will have access to the neutron multiplicity as a function of mass. The axial ionization chamber, in addition to the kinetic energy value, will measure the energy loss profile of the fragment along its track. This energy loss profile will give information about the fragment nuclear charge. This paper will focus on recent developments on the FALSTAFF design. A special attention will be paid to the impact of the detector material thickness on the uncertainty of different observables.

Eutectic Carbides in Damascus steel Ledeburite Class (Wootz)

Considered the nature of the change of the morphology of excess carbides in Damascus steel (Wootz), depending on the degree of supercooling of the melt, heat treatment and plastic deformation. Discovered that some of blades Damascus steel has an unusual nature of origin of the excess cementite, which different from the redundant phases of secondary cementite, cementite of ledeburite and primary cementite in iron-carbon alloys. It is revealed that the morphological features of separate particles of cementite in Damascus steels lies in the abnormal size of excess carbides having the shape of irregular prisms. Considered three hypotheses for the formation of excess cementite in the form of faceted prismatic of excess carbides. The first hypothesis is based on thermal fission of cementite of a few isolated grains. The second hypothesis is based on the process of fragmentation cementite during deformation to the separate the pieces. The third hypothesis is based on the transformation of metastable cementite in the stable of angular eutectic carbide. It is shown that the angular carbides are formed within the original metastable colony ledeburite, so they are called “eutectic carbide”. It is established that high-purity white cast iron is converted into of Damascus steel during isothermal soaking at the annealing. It was revealed that some of blades Damascus steel ledeburite class do not contain in its microstructure of crushed ledeburite. It is shown that the pattern of carbide heterogeneity of Damascus steel consists entirely of angular eutectic carbides. Believe that Damascus steel refers to non-heat-resistant steel of ledeburite class, which have similar structural characteristics with semi-heat-resistant die steel or heat-resistant high speed steel, differing from them only in the nature of excess carbide phase.

Covariance generation and uncertainty propagation for thermal and fast neutron induced fission yields

Fission product yields (FY) are fundamental nuclear data for several applications, including decay heat, shielding, dosimetry, burn-up calculations. To be safe and sustainable, modern and future nuclear systems require accurate knowledge on reactor parameters, with reduced margins of uncertainty. Present nuclear data libraries for FY do not provide consistent and complete uncertainty information which are limited, in many cases, to only variances. In the present work we propose a methodology to evaluate covariance matrices for thermal and fast neutron induced fission yields. The semi-empirical models adopted to evaluate the JEFF-3.1.1 FY library have been used in the Generalized Least Square Method available in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation) to generate covariance matrices for several fissioning systems such as the thermal fission of U235, Pu239 and Pu241 and the fast fission of U238, Pu239 and Pu240. The impact of such covariances on nuclear applications has been estimated using deterministic and Monte Carlo uncertainty propagation techniques. We studied the effects on decay heat and reactivity loss uncertainty estimation for simplified test case geometries, such as PWR and SFR pin-cells. The impact on existing nuclear reactors, such as the Jules Horowitz Reactor under construction at CEA-Cadarache, has also been considered.

Evaluation of the235U resonance parameters to fit the standard recommended values

A great deal of effort has been dedicated to the revision of the standard values in connection with the neutron interaction for some actinides. While standard data compilation are available for decades nuclear data evaluations included in existing nuclear data libraries (ENDF, JEFF, JENDL, etc.) do not follow the standard recommended values. Indeed, the majority of evaluations for major actinides do not conform to the standards whatsoever. In particular, for the n + 235 U interaction the only value in agreement with the standard is the thermal fission cross section. A resonance re-evaluation of the n + 235 U interaction has been performed to address the issues regarding standard values in the energy range from 10−5 eV to 2250 eV. Recently, 235 U fission cross-section measurements have been performed at the CERN Neutron Time-of-Flight facility (TOF), known as n_TOF, in the energy range from 0.7 eV to 10 keV. The data were normalized according to the recommended standard of the fission integral in the energy range 7.8 eV to 11 eV. As a result, the n_TOF averaged fission cross sections above 100 eV are in good agreement with the standard recommended values. The n_TOF data were included in the 235 U resonance analysis that was performed with the code SAMMY. In addition to the average standard values related to the fission cross section, standard thermal values for fission, capture, and elastic cross sections were also included in the evaluation. This paper presents the procedure used for re-evaluating the 235 U resonance parameters including the recommended standard values as well as new cross section measurements.

Influence of void fraction in the power distribution for a GE-12 fuel assembly

Analysis of the influence of void fraction distribution is very important to learn about the fission process and heat produced by the fuel assembly, here in this study several void fraction (VF) values along different burnup values have been considered in order to observe their influence in power distribution, uranium consumption, neutron flux andbehaviour for a GE-12 fuel assembly. For this study, burnups up to 60 MWd/kg and VF values up to 0.8 were considered setting the uranium enrichment at 3.5 weight percent at the start of every VF scenario, results show that higher void fractions reduce the thermal flux decreasing thermal fission and limiting heat production.

Consistent adjustment of radioactive decay and fission yields data with measurement data of decay heat and [formula omitted]-delayed neutron activities

Decay heat and delayed neutron yields, which are important physical quantities in the field of the nuclear engineering, are dependent on common nuclear data such as radioactive decay data and fission yields data of fission product nuclides. In the present study, correlations between uncertainties of these two quantities are investigated.Nuclear data relevant to uranium-235 and plutonium-239 fissions with thermal neutron are adjusted consistently with a procedure based on Bayes’ theorem using the measurement data of decay heat and delayed neutron activities. Numerical results suggest that the correlation between decay heat and delayed neutron activities uncertainties is not significant, and that independent treatments of decay heat or delayed neutron activities are possible. The effect of the consistent treatment of decay heat and delayed neutron activities is, however, observed in the adjustment results in some nuclear data such as uranium-235 thermal fission yields of yttrium-100m and zirconium-100.

The experimental determination of the 238U(n,γ) and total fission reaction rates along the pellet radius of the IPEN/MB-01 reactor

This work presents a new experimental approach to determine experimentally the reaction rate along the pellet radius of a research reactor facility. The methodology employs concentric hollow cylindrical lead collimators together with gamma-ray spectrometry in a special kind of HPGe detector for the discrimination of very low gamma energy emitted by 239Np and 99Mo. The 238U neutron capture and total fission rates were inferred, respectively, from the 106.2keV gamma emitted by 239Np and from the 140.51keV gamma-ray emitted by 99Mo. These lower gamma-ray energies allow the thickness of the lead collimator to be small enough so that the correction factors applied to the procedure were minimized. This kind somewhat challenging experiment was successfully performed at the IPEN/MB-01 research reactor facility. The experiments are claimed to be well-defined, and they are suitable for a benchmark. The measured values of the total fission rates are mainly due to the thermal fissions in 235U since in the IPEN/MB-01 reactor nearly 85% of the total fissions occur in the thermal neutron energy region. The theoretical analyses were performed using MCNP-5 together with the ENDF/B-VII.0 library. The analyses reveal a very good agreement between the calculated and experimental results for the 238U epithermal neutron capture reaction rates. However, the same can not be said for the thermal reaction rates which show discrepancies both in magnitude as well as in the shape of the attenuation of the reaction rates inside of the fuel pellet. The suspected reason for these discrepancies is the shape of the 235U cross sections below 0.3eV which might be different from that adopted in the ENDF/B-VII.0 library.

Thermal fission rates with temperature dependent fission barriers

The fission processes of thermal excited nuclei are conventionally studied by statistical models which rely on inputs of phenomenological level densities and potential barriers. Therefore the microscopic descriptions of spontaneous fission and induced fission are very desirable for a unified understanding of various fission processes. We propose to study the fission rates, at both low and high temperatures, with microscopically calculated temperature-dependent fission barriers and collective mass parameters. The fission barriers are calculated by the finite-temperature Skyrme-Hartree-Fock+BCS method. The mass parameters are calculated by the temperature-dependent cranking approximation. The thermal fission rates can be obtained by the imaginary free energy approach at all temperatures, in which fission barriers are naturally temperature dependent. The fission at low temperatures can be described mainly as a barrier-tunneling process. While the fission at high temperatures has to incorporate the reflection above barriers. Our results of spontaneous fission rates reasonably agree with other studies and experiments. The temperature dependencies of fission barrier heights and curvatures have been discussed. The temperature dependent behaviors of mass parameters have also been discussed. The thermal fission rates from low to high temperatures with a smooth connection have been given by different approaches. \item[Conclusions] Since the temperature dependencies of fission barrier heights and curvatures, and the mass parameters can vary rapidly for different nuclei, the microscopic descriptions of thermal fission rates are very valuable. Our studies without free parameters provide a consistent picture to study various fissions such as that in fast-neutron reactors, astrophysical environments and fusion reactions for superheavy nuclei.

Multiphysics modeling of UO2-SiC composite fuel performance with enhanced thermal and mechanical properties

The relatively poor thermal conductivity of the UO2 fuel is a major challenge for optimizing reactor operation and safety performance, despite its widespread use in the majority of power reactors. The already low thermal conductivity further degrades with burnup due to accumulation of defects, fission product precipitates and fission gas bubbles. The high thermal stresses cause significant pellet cracking, leading to more pellet expansion and causing pellet cladding interaction and the release of fission product gases. Mitigation of these phenomena is accomplished by limiting operating power and ramp rates at high burnup. This study presents the development of a model for UO2-SiC composite light water reactor fuel with enhanced thermal and mechanical properties to mitigate some of the material performance limitations. The model is constructed using self-defined physics modules which are fully coupled and solved using the COMSOL Multiphysics platform. Key fuel performance phenomena being considered include heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the equations are solved via finite-element method on a 2D axisymmetric geometry of a fuel pellet and cladding. The use of UO2-SiC will enhance thermal conductivity, which in turn will decrease nuclear fuel temperatures. This subsequently will improve fission gas retention and reduce pellet cladding interaction resulting from fuel cracking, relocation, and swelling.

Dynamical simulation of neutron-induced fission of uranium isotopes using four-dimensional Langevin equations

Four-dimensional Langevin equations have been suggested for the dynamical simulation of neutron-induced fission at low and medium excitation energies. The mass distribution of the fission fragments, the neutron multiplicity, and the fission cross section for the thermal and fast neutron-induced fission of $^{233}\mathrm{U},\phantom{\rule{0.16em}{0ex}}^{235}\mathrm{U}$, and $^{238}\mathrm{U}$ is studied by considering energy dissipation of the compound nucleus through the fission using four-dimensional Langevin equations combined with a Monte Carlo simulation approach. The calculated results using this approach indicate reasonable agreement with available experimental data.

Probing energy dissipation, $\gamma$ -ray and neutron multiplicity in the thermal neutron-induced fission of 239Pu

The incorporation of the four-dimensional Langevin equations led to an integrative description of fission cross-section, fragment mass distribution and the multiplicity and energy distribution of prompt neutrons and \( \gamma\)-rays in the thermal neutron-induced fission of 239Pu. The dynamical approach presented in this paper thoroughly reproduces several experimental observables of the fission process at low excitation energy.

Effects of Fission Yield Data in the Calculation of Antineutrino Spectra for ^{235}U(n,fission) at Thermal and Fast Neutron Energies.

Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 ^{235}U fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of ^{86}Ge generates an excess of antineutrinos at 5-7MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0-7MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Finally, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.

Mechanism study and theoretical simulation on heat split phenomenon in dual-cooled annular fuel element

In this study, the mechanisms that cause heat split, a specific phenomenon in dual-cooled annular fuel elements, were investigated. On the basis of thermal resistance analysis, the developing process of heat split phenomenon was summarized. It was found that thermal expansion, fuel densification, swelling, creep, relocation and fission gas release are the original parameters driving the development of heat split. Due to these factors, thermal resistances for heat flux to internal and external channels change with the operating time, which results in heat split. Furthermore, the theoretical simulation on heat split was performed by FROBA-ANNULAR, which is a coupled thermal–mechanical analysis code for dual-cooled annular fuel elements. Key parameters at different burnup stages, including gap size, gap conductance, temperature profile, coolant flux and heat flux were obtained. The calculation result shows that the fraction of heat flux to the internal cooling channel experiences complicated variation at the range of 32.9–39.7%. In addition, it was found that heat split does not match with the coolant split, especially at the lower burnup stage. Asymmetric heat transfer occurred because of the imbalance between heat split and coolant split, resulting in a substantial asymmetric temperature profile. Furthermore, heat split exerts an influence on DBNRs of inner and outer channels, although the computational values were still within the allowable limits.

Determination of Nuclear Charge Distributions of Fission Fragments from $$^{235}$$ 235 U ( $$n_\mathrm{th}$$ n th , f) with Calorimetric Low Temperature Detectors

Calorimetric low temperature detectors (CLTD’s) for heavy-ion detection have been combined with the LOHENGRIN recoil separator at the ILL Grenoble for the determination of nuclear charge distributions of fission fragments produced by thermal neutron-induced fission of \(^{235}\)U. The LOHENGRIN spectrometer separates fission fragments according to their mass-to-ionic-charge ratio and their kinetic energy, but has no selectivity with respect to nuclear charges Z. For the separation of the nuclear charges, one can exploit the nuclear charge-dependent energy loss of the fragments passing through an energy degrader foil (absorber method). This separation requires detector systems with high energy resolution and negligible pulse height defect, as well as degrader foils which are optimized with respect to thickness, homogeneity, and energy loss straggling. In the present, contribution results of test measurements at the Maier Leibnitz tandem accelerator facility in Munich with \(^{109}\)Ag and \(^{127}\)I beams with the aim to determine the most suitable degrader material, as well as measurements at the Institut Laue–Langevin will be presented. These include a systematic study of the quality of Z-separation of fission fragments in the mass range \(82\le A \le 132\) and a systematic measurement of \(^{92}\)Rb fission yields, as well as investigations of fission yields toward the symmetry region.

Analysis of prompt fission neutrons in235U(nth,f) and fission fragment distributions for the thermal neutron induced fission of234U

This paper presents the ongoing analysis of two fission experiments. Both projects are part of the collaboration between the nuclear reactions group at Uppsala and the JRC-IRMM. The first experiment deals with the prompt fission neutron multiplicity in the thermal neutron induced fission of 235 U(n,f). The second, on the fission fragment properties in the thermal fission of 234 U(n,f).The prompt fission neutron multiplicity has been measured at the JRC-IRMM using two liquid scintillators in coincidence with an ionization chamber. The first experimental campaign focused on 235 U(n th ,f) whereas a second experimental campaign is foreseen later for the same reaction at 5.5 MeV. The goal is to investigate how the so-called sawtooth shape changes as a function of fragment mass and excitation energy. Some harsh experimental conditions were experienced due to the large radiation background. The solution to this will be discussed along with preliminary results.In addition, the analysis of thermal neutron induced fission of 234 U(n,f) will be discussed. Currently analysis of data is ongoing, originally taken at the ILL reactor. The experiment is of particular interest since no measurement exist of the mass and energy distributions for this system at thermal energies. One main problem encountered during analysis was the huge background of 235 U(n th ,f). Despite the negligible isotopic traces in the sample, the cross section difference is enormous. Solution to this parasitic background will be highlighted.

A Covariance Generation Methodology for Fission Product Yields

Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1) no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM) implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation), developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

Medical Radioisotope Production in a Power-Flattened ADS Fuelled with Uranium and Plutonium Dioxides

This study presents the medical radioisotope production performance of a conceptual accelerator driven system (ADS). Lead-bismuth eutectic (LBE) is selected as target material. The subcritical fuel core is conceptually divided into ten equidistant subzones. The ceramic (natural U, Pu)O2fuel mixture and the materials used for radioisotope production (copper, gold, cobalt, holmium, rhenium, thulium, mercury, palladium, thallium, molybdenum, and yttrium) are separately prepared as cylindrical rods cladded with carbon/carbon composite (C/C) and these rods are located in the subzones. In order to obtain the flattened power density, percentages of PuO2in the mixture of UO2and PuO2in the subzones are adjusted in radial direction of the fuel zone. Time-dependent calculations are performed at 1000 MW thermal fission power (Pth) for one hour using the BURN card. The neutronic results show that the investigated ADS has a high neutronic capability, in terms of medical radioisotope productions, spent fuel transmutation and energy multiplication. Moreover, a good quasiuniform power density is achieved in each material case. The peak-to-average fission power density ratio is in the range of 1.02–1.28.

Time-Dependent Neutronic Analysis of a Power-Flattened Gas Cooled Accelerator Driven System Fuelled with Thorium, Uranium, Plutonium, and Curium Dioxides TRISO Particles

This study presents the power flattening and time-dependent neutronic analysis of a conceptual helium gas cooled Accelerator Driven System (ADS) loaded with TRISO (tristructural-isotropic) fuel particles. Target material is lead-bismuth eutectic (LBE). ThO2, UO2, PuO2, and CmO2TRISO particles are used as fuel. PuO2and CmO2fuels are extracted from PWR-MOX spent fuel. Subcritical core is radially divided into 10 equidistant subzones in order to flatten the power produced in the core. Tens of thousands of these TRISO fuel particles are embedded in the carbon matrix fuel pebbles as five different cases. The high-energy Monte Carlo code MCNPX 2.7 with the LA150 library is used for the neutronic calculations. Time-dependent burnup calculations are carried out for thermal fission power (Pth) of 1000 MW using the BURN card. The energy gain of the ADS is in the range of 99.98–148.64 at the beginning of a cycle. Furthermore, the peak-to-average fission power density ratio is obtained between 1.021 and 1.029 at the beginning of the cycle. These ratios show a good quasi-uniform power density for each case. Furthermore, up to 155.1 g233U and 103.6 g239Pu per day can be produced. The considered system has a high neutronic capability in terms of energy multiplication, fissile breeding, and spent fuel transmutation with thorium utilization.

Estimation of Covariances on Prompt Fission Neutron Spectra and Impact of the PFNS Model on the Vessel Fluence

As the need for precise handling of nuclear data covariances grows ever stronger, no information about covariances of prompt fission neutron spectra (PFNS) are available in the evaluated library JEFF-3.2, although present in ENDF/B-VII.1 and JENDL-4.0 libraries for the main fissile isotopes. The aim of this work is to provide an estimation of covariance matrices related to PFNS, in the frame of some commonly used models for the evaluated files, such as the Maxwellian spectrum, the Watt spectrum, or the Madland-Nix spectrum. The evaluation of PFNS through these models involves an adjustment of model parameters to available experimental data, and the calculation of the spectrum variance-covariance matrix arising from experimental uncertainties. We present the results for thermal neutron induced fission of 235 U. The systematic experimental uncertainties are propagated via the marginalization technique available in the CONRAD code. They are of great influence on the final covariance matrix, and therefore, on the spectrum uncertainty band width. In addition to this covariance estimation work, we have also investigated the importance on a reactor calculation of the fission spectrum model choice. A study of the vessel fluence depending on the PFNS model is presented. This is done through the propagation of neutrons emitted from a fission source in a simplified PWR using the TRIPOLI-4 ® code. This last study includes thermal fission spectra from the FIFRELIN Monte-Carlo code dedicated to the simulation of prompt particles emission during fission.

Comparison of deterministic and stochastic approaches for isotopic concentration and decay heat uncertainty quantification on elementary fission pulse

Uncertainty quantification of interest outputs in nuclear fuel cycle is an important issue for nuclear safety, from nuclear facilities to long term deposits. Most of those outputs are functions of the isotopic vector density which is estimated by fuel cycle codes, such as DARWIN/PEPIN2, MENDEL, ORIGEN or FISPACT. CEA code systems DARWIN/PEPIN2 and MENDEL propagate by two different methods the uncertainty from nuclear data inputs to isotopic concentrations and decay heat. This paper shows comparisons between those two codes on a Uranium-235 thermal fission pulse. Effects of nuclear data evaluation’s choice (ENDF/B-VII.1, JEFF-3.1.1 and JENDL-2011) is inspected in this paper. All results show good agreement between both codes and methods, ensuring the reliability of both approaches for a given evaluation.