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The photo-atomic sublibrary aims to describe atomic photon cross section data, corresponding to the interaction between photons and the different atoms. For ENDF/B-VIII.1 we adopted the EPICS2023 EPDL evaluations.
The protons sublibrary aims to describe nuclear reactions between incident proton particles and different nuclei. For ENDF/B-VIII.1, compared to ENDF/B-VIII.0, there were only evaluation updates for 4 He.
The neutron standards sublibrary describes specific reaction cross sections, in a limited range, that are so well-known that they are used as ratios or references in other measurements. ENDF/B-VIII.1 remains unchanged from ENDF/B-VIII.0 for the neutron standards sublibrary
The thermal neutron scattering law (TSL) sublibrary aims to describe the interaction of incident neutrons at thermal or sub-thermal energies with different compound materials such fuels, moderators and special-purpose materials. In ENDF/B-VIII.1 there was a large number of new and updated TSL evaluations, including traditional moderators (light water, Beryllium metal, Beryllium Oxide, Calcium Hydride, plastics (Polystyrene and Lucite), graphite (reactor-grade and crystalline), anhydrous Hydrogen Fluoride, and heavy paraffinic oil); exotic moderators (Beryllium Carbide, Zirconium Hydride, Yttrium Hydride, Lithium-7 Hydride and Deuteride), FLiBe molten salt, structural materials and cladding (Silicon Carbide, Silicon Dioxide, Zirconium Carbide), fuels (Plutonium Dioxide, Uranium Carbide, Uranium metal, Uranium Nitride, Uranium Dioxide, Uranium Hydride), and special purpose materials. In ENDF/B-VIII.1 we also distribute alongside the evaluated files, a comma-separated file (CSV), named TSL_MAT_numbers.csv, which lists all evaluated files in the current release and their corresponding unique MAT number.
The tritons sublibrary aims to describe nuclear reactions between incident triton particles and different nuclei. For ENDF/B-VIII.1, compared to ENDF/B-VIII.0, there were only evaluation updates for 4 He.
Here, we analyze correlated-triplet-pair (TT) singlet-fission intermediates toward two-triplet separation (T...T) using spin-state-averaged density matrix renormalization group electronic-structure calculations. Specifically, we compare the triplet–triplet exchange (J) for tetracene dimers, bipentacene, a subunit of the benzodithiophene–thiophene dioxide polymer, and a carotenoid (neurosporene). Exchange-split energy gaps of J and 3J separate a singlet from a triplet and a singlet from a quintet, respectively. We draw two new insights: (a) the canonical tetracene singlet-fission unit cell supports precisely three low-lying TT intermediates with order-of-magnitude differences in J, and (b) the separable TT intermediate in carotenoids emanates from a pair of excitations to the second triplet state. Therefore, unlike with tetracenes, carotenoid fission requires above-gap excitations. In all cases, the distinguishability of the molecular triplets—that is, the extent of orbital overlap—determines the splitting within the spin manifold of TT states. Consequently, J represents a spectroscopic observable that distnguishes the resemblance between TT intermediates and the T...T product.
A search for new neutron-rich isotopes beyond $Z=28$ and $N=50$ was performed using a 345 MeV/nucleon $^{238}$U beam at RI Beam Factory in RIKEN Nishina Center. The BigRIPS separator and state-of-the-art radiation detectors were used to efficiently separate and identify the produced nuclei. For the first time, we have observed and determined the production cross sections for eight neutron-rich isotopes $^{72}$Cr, $^{74,75}$Mn, $^{77}$Fe, $^{79,80}$Co, $^{85}$Cu, and $^{90}$Ga. This study broadens our understanding of nuclear existence and paves the way for investigating the properties of neutron-rich medium-mass nuclei.
There is a very strong need for statistical testing to determine fission source convergence in Monte Carlo criticality calculations. Automation of such tests will greatly streamline and support the work carried out by NCS practitioners. Recent R&D work has shown that no single statistical test for convergence is sufficiently reliable, robust, and “guaranteed.” However, a combination of several standard statistical tests for the similarity of distributions, coupled with a high-fidelity estimate of the fission-matrix source is sufficiently robust, reliable, and repeatable that convergence can be “guaranteed.” During the course of the EG-AMCT studies, a number of statistical metrics and tests were proposed for diagnosing clustering and undersampling. None of these was robust and reliable enough for practical use in production codes. However, the expert group efforts came close. Some recent R&D work stemming from those past efforts has been very successful and promising. This Sub-Group will provide international input and collaboration on the development and implementation of statistical tests for convergence, with the primary goal of having the MC codes automatically detect convergence (or lack thereof). Newly proposed statistical tests to detect undersampling (after convergence) will also be reviewed.
Supported by the DOE-NNSA Office of Material Management & Minimization (M3), a number of domestic entities are pursuing non-highly enriched uranium (non-HEU) production of 99 Mo. As the production technologies of 99Mo pivot toward low-enriched uranium (LEU) or molybdenum targets, new reaction channels and accelerators are being evaluated. Superconducting electron linear accelerators (LINACs) with high-Z converter targets can generate bremsstrahlung photons and neutron fluxes that are capable of inducing photonuclear reactions and LEU fission. A particular advantage of a LINAC is that it does not rely on HEU-fueled reactor cores (which are currently slated for LEU conversion) and can operate on an almost continuous basis. Regarding the chemical purification of 99 Mo from irradiated uranium targets under acidic digestion, there exists a procedure known as Cintichem or modifications thereof with respect to LEU (LEU Modified Cintichem process, LMC). The process relies on a number of selective precipitation steps and column chromatography to purify Mo. It is important to note that LMC prescribes the addition of stable Mo to carry 99 Mo on alpha benzoin oxime, which reduces the specific activity of 99 Mo. This is especially important for processing 99 Mo batches with lower activities (~33 Ci of 99 Mo per batch).
Fission surface power (FSP) systems could be used to provide power anytime, anywhere on the surface of the Moon or Mars. FSP systems could be used at polar locations, at locations away from the poles, or in permanently shaded regions, with excellent performance at all sites. A potential reference 40 kWe option has been devised that is cost-competitive with alternatives while providing more power for less mass anywhere on the lunar surface. The reference FSP system (FSPS) is also readily extensible for use on Mars. At Mars the system would be capable of operating through global dust storms and providing year-round power at any Martian latitude. Under the NASA Exploration Technology Development Program (ETDP), NASA and the Department of Energy (DOE) have begun technology development on Fission Surface Power (FSP). The primary customer for this technology is the NASA Constellation Program which is responsible for the development of surface systems to support human exploration on the moon and Mars. The objectives of the FSP technology project are: 1) Develop FSP concepts that meet expected surface power requirements at reasonable cost with added benefits over other options. 2) Establish a hardware-based technical foundation for FSP design concepts and reduce overall development risk. 3) Reduce the cost uncertainties for FSP and establish greater credibility for flight system cost estimates. 4) Generate the key products to allow Agency decision-makers to consider FSP as a viable option for flight development. To be mass efficient, FSP systems must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial systems. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference FSP system uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance on the surface of the moon or Mars. Recent testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference FSP system, and has helped evaluate methods for system integration. In June, 2009, a representative pumped NaK loop (provided by Marshall Space Flight Center) was coupled to a Stirling power converter (provided by Glenn Research Center) and tested at various conditions representative of those that would be seen during actual FSP system operation. In all areas, performance of the integrated system exceeded project goals. High-temperature NaK pump testing has also been performed at the EFF-TF, as has testing of methods for providing long-duration NaK purity.
Diffusion of fission gas in UO 2 nuclear fuel impacts several important performance metrics, such as fission gas release, swelling, and thermal conductivity. Current empirical models of fission gas release have significant uncertainty, some of which derives from the bulk diffusion rate and its dependence on, for example, fuel chemistry and irradiation. In this work, we have applied the previously-developed Free Energy Cluster Dynamics (FECD) methodology in the code Centipede to calculate xenon cluster concentrations in UO 2 under intrinsic (high temperature) and irradiation-enhanced (intermediate temperature) conditions in order to develop a model of the xenon diffusion coefficient based on the atomic scale mechanisms responsible for transport. While the diffusion mechanism for xenon in UO 2 is adequately described by the Xe + U 2 O vacancy cluster for intrinsic conditions, a similar process is not capable of capturing measured in-pile fission gas diffusivity at intermediate temperatures. Therefore, a different diffusion mechanism must dominate under this regime. Using calculated atomistic data, we have shown that irradiation-enhanced diffusion at intermediate temperatures occurs via the larger Xe + U 4 O y vacancy clusters, which have lower migration barriers and increase in concentration by several orders of magnitude compared to intrinsic conditions. This mechanism is enabled by the increased uranium vacancy concentration under irradiation due to Frenkel pair production. In addition, the fast migration of uranium interstitials with two attached oxygen interstitials lowers the total uranium interstitial concentration through reactions with sinks. This allows the extended defects, such as Xe + U 4 O y vacancy clusters, to maintain high concentrations by limiting annihilation with attached vacancies. Furthermore, predictions using the Xe + U 4 O y diffusion mechanism are in good agreement with experiment, albeit with some differences in the Arrhenius slope, which we believe may be related to either experimental or model parameter uncertainty. Lastly, an analytical expression suitable for application in fuel performance simulations was derived to capture the predictions of the Centipede simulations.
A potential new zirconium recycling protocol has been demonstrated using unirradiated Zircaloy-based claddings and depleted uranium oxide for the chemical removal of zirconium alloy cladding in used nuclear fuel (UNF) rods from light water reactors. This protocol is based on the application of three new scientific findings. First, a new lower temperature chlorination reaction of zirconium with sulfur monochloride is described. Second, the high solubility of zirconium chloride in thionyl chloride is used to separate it from uranium oxide fuel and fission and activation contaminants. Finally, in the third step, the zirconium chloride is purified by simple recrystallization from thionyl chloride. Utilizing sulfur chloride solvents for a lower temperature liquid-based chemical digestion and purification of zirconium alloy claddings reduces technical complications experienced by current high temperature gas phase chlorination strategies, such as contamination of product streams. This novel protocol has been demonstrated on a 3–50 g scale of unirradiated zirconium alloy, with no significant change in the time required for complete chlorination (3–4 h). This zirconium chlorination protocol has also been performed in the presence of depleted uranium oxide pellets. The depleted fuel pellets do not affect zirconium chlorination, and the chlorination protocol does not chemically or physically alter the fuel pellets. Finally, a preliminary description of an industrial protocol to recycle nuclear grade zirconium from UNF rods is presented.
The Department of Energy (DOE)’s NEAMS focuses its efforts on the development of advanced modeling and simulation (M&S) tools for light-water reactors (LWRs) and non–LWRs (i.e., molten salt reactors, high-temperature gas reactors, microreactors, and fast reactors). In the previous fiscal year, the Multiphysics Applications Driver Technical Area funded molten salt reactor (MSR) M&S at Oak Ridge National Laboratory (ORNL) to generate multigroup macroscopic cross sections with Shift for a MSRE 2D lattice model in Griffin. In addition, Shift’s capability to calculate gamma dose rates from activated components in the primary exchangers in a molten salt breeder reactor was also demonstrated. In fiscal year 2023, ORNL generated multigroup macroscopic cross sections using Shift for a 3D MSRE core model. MSRE depletion calculations using Griffin were also demonstrated in this fiscal year. For the depletion calculation, one-group microscopic cross sections for the 3D MSRE core were generated using Shift, and the decay transmutation library from ORIGEN was converted to an ISOXML file, which is required as input in Griffin. Several Monte Carlo codes, such as OpenMC and Serpent, were also used to benchmark and supplement multigroup cross sections generated by Shift. Multigroup libraries were generated with 8 and 20 group structures, and the study found the 8-group structure to be more accurate when comparing Griffin results to continuous energy (CE) Monte Carlo results. The average flux from CE Shift calculations is up to ~6% higher than the CE Serpent calculations because of different values applied for the energy released per fission (κ values). The average flux in the fuel salt calculated by Griffin using cross sections generated with Shift agrees well with the reference CE Shift solution; the same is valid for the corresponding Serpent results. The maximum relative error is ~6% and ~2% compared to the CE Shift and Serpent reference solutions, respectively. Meanwhile, the average flux calculated by Griffin in the graphite moderator shows a higher difference in the thermal range when compared to both reference Monte Carlo solutions; this result suggests a need for improvement in cross section generation for the graphite moderator in the thermal range in both Monte Carlo codes. Griffin depletion calculations using cross sections from Shift and Serpent were performed and compared against ORIGEN calculations, and the nuclide densities calculated by Griffin were found to be generally in agreement with those of ORIGEN. Because a different approach was taken to calculate the energy released per fission ( κ values) in Shift and Serpent, a difference in nuclide densities from differences in the average flux was observed between Griffin using Serpent and Shift cross sections. Griffin calculations with Shift cross sections produced higher average flux in the salt than with Serpent cross sections, leading to higher consumption of 235 U and higher production of 135 Xe. For time-dependent depletion calculations, cross sections were generated with Serpent, and Griffin’s results using these cross sections were compared to CE Serpent depletion results, demonstrating good agreement. The average difference in keff between Serpent and Griffin as a function of burnup is about 155 pcm. Similarly, good agreement with small differences up to ~0.5% was also noticed in the nuclide density of 235 U and 135 Xe. More details regarding the methodologies invoked to generate the cross section to make code-to-code comparisons are discussed further in this report. User feedback on Griffin and Shift capabilities that will enhance these calculations is provided in this report for future consideration. The work performed this fiscal year can be extended further for multiphysics coupling of Griffin-Pronghorn/SAM with Mole to study precursor flow and salt chemistry.
NASA Marshall Space Flight Center (MSFC) is NASA s principle propulsion development center. Research and development is coordinated and carried out on not only the existing transportation systems, but also those that may be flown in the near future. Heat pipe cooled fast fission cores are among several concepts being considered for the Nuclear Systems Initiative. Marshall Space Flight Center has developed a capability to handle high-purity alkali metals for use in heat pipes or liquid metal heat transfer loops. This capability is a low budget prototype of an alkali metal handling system that would allow the production of flight qualified heat pipe modules or alkali metal loops. The processing approach used to introduce pure alkali metal into heat pipe modules and other test articles are described in this paper.
Validated integrated modelling of JET ITER-like wall experiments in which fusion performance is driven by reactions between fast ions and intrinsically present metal wall impurities is presented. A steady-state L-mode plasma with dominant proton-beryllium fusion and neutron yields of up to ≈ 6·10 13 s -1 is developed in He and D, via radiofrequency heating of a H minority. The fusion drive is unambiguously confirmed by the neutral particle analyser, fast ion loss detector, and γ-ray diagnostics. Experiments are analysed via an integrated modelling framework, developed to model the two-stage proton beryllium-fusion chain and produce high-fidelity fusion product source terms. The modelling chain comprises TRANSP and JETTO for plasma core modelling, LOCUST for full orbit product tracking and collisional slowing-down, DRESS to resolve two- and three-body fusion kinematics, and MCNP for neutron transport calculations. Modelling shows that the primary 9 Be(p,n) 9 B reaction is the dominant neutron emitter at naturally present concentrations of beryllium in these experiments. The yield contribution of secondary reactions between fusion products and beryllium, 9 Be(d,n) 10 B and 9 Be(α,n) 12 C, is found to be negligible. The proton-deuteron knock-on effect in D plasmas is modelled, which is calculated to contribute ≈ 25% to the total neutron yield. For both He and D discharges the total computed neutron rates match fission chamber (FC) measurements within the combined experimental and computational uncertainty, with an average discrepancy of ≈ ± 20%. Realistic proton-beryllium neutron sources are propagated through JET’s MCNP neutron transport model which shows that 235 U FCs’ response is sensitive to p–Be source changes, with up to ≈ 10% variation compared to a D–D neutron source. We show that the high-energy tail of the fast proton minority can be studied with multi-foil neutron activation. The framework is also applied to the study of interactions between fast protons and boron impurities, of relevance to ITER. We calculate that in JET conditions a significant alpha source with DT-like energies could be generated through 11 B(p, α)2α fusion, and detected via γ-emission in secondary interactions between fast alphas and boron. The work represents an important step towards validating predictive integrated modelling capabilities for non-standard fusion reactions.
For the advancement and development of nuclear systems used in heavy irradiation environments such as fusion DEMO reactors, fission reactors, fast reactors, and accelerator driven target systems, it is necessary to fully understand the changes of mechanical properties and the other properties of the materials induced by irradiation and to clarify the synergistic effect of displacement damage and helium generation. In this study the mechanical property changes and microstructural development induce by displacement damage and helium production have been mainly examined in austenitic stainless steels, 316FR and type304, and ferritic/martensitic steel, HCM12A, irradiated at around 550oC in JRR-3M reactor and/or JOYO fast reactor. At 550°C, 316FR steel was superior to 304 steel in terms of the amount of ductility and strength with respect to irradiation resistance. It is noteworthy that at 550 °C as well as room temperature, the higher fracture strength of the 316FR steels is a remarkable result. It is found that helium atoms strongly influenced on creep lifetime of the irradiated austenitic stainless steel, 316FR. It was found that the ratio of creep rupture time is slightly lower than the lower limit of previous study’s Miyaji and co-workers in the region from 0.01 appm to 1 appm. It is also found that the lower limit of reduction ratio of creep rupture time (irradiation specimen to unirradiation one) does not decrease linearly with the helium production above 10 appm up to about 33 appm. dpa enhanced the reduction of creep lifetime. Recent R&D of high-energy accelerator driven target systems used under heavy irradiation environment is also introduced and discussed for high radiation resistance materials such as Ti alloys with very high number density of nano size precipitate, which have been studying under RaDIATE collaboration, from points of view of irradiation damage and materials development.
The 9 B neutron capture cross section has significant implications for Be materials in the nuclear industry as well as the α process in stellar nucleosynthesis. While the cross section is well constrained at thermal neutron energies, there is a lack of experimental data at higher neutron energies, and the evaluated nuclear data libraries can differ by up to two orders of magnitude. We calculate the 9 Be(n, γ) 10 Be integral cross section at fission neutron energies in an effort to resolve disagreements amongst the nuclear data libraries. Foil irradiation experiments were performed using the Flattop critical assembly at the National Criticality Experiments Research Center with either the highly enriched U or Pu cores, with target foil stacks placed at multiple locations to exploit different neutron energy profiles. Accelerator mass spectrometry was used to measure the 10 Be/ 9 Be ratio in irradiated Be foils, while all other activation products were quantified through gamma spectrometry. The experiments were simulated using the Monte Carlo N-Particle radiation transport code and combined with experimental results to determine the total neutron fluence, while the staysl-pnnl suite and fispact-ii code were used to validate the model and assess the systematic uncertainty. The new 9 Be(n, γ) 10 Be integral cross sections calculated in this work are 26.5 ± 2.2µb at 0.59 ± 0.07 MeV, 24 ± 3 µb at 0.98 ± 0.14 MeV, 21.7 ± 1.3 µb at 1.26 ± 0.11 MeV, 21.8 ± 1.4 µb at 1.32 ± 0.11 MeV, and 18.6 ± 1.1 µb at 1.46 ± 0.13 MeV. These results do not agree with integral cross sections from any of the nuclear data library evaluations. Discrepancies between the new integral cross sections reported here and the nuclear data libraries suggest a more complex cross-section structure in the MeV range which allows for more resonance contributions, and more work is needed to further constrain the evaluated cross sections.
Relativistic nuclei of krypton, xenon, holmium, and gold, accelerated in a partially stripped state to a maximum rigidity of about 5.6 GeV, interacting with targets of aluminum, carbon, and polyethylene are examined. For each projectile and target combination, determinations are made for the total and partial charge changing cross sections for the production of lighter fragments. From these measurements, a new representation of the dependence of the total charge changing cross sections on beam and target charge is developed. Simple representations of the variation of the partial cross sections were identified with the charge of the produced fragments and shown to be dependent on the charge and energy of the beam. The fission of gold nuclei at high energies in these various targets has also been studied.