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Least squares analyses of experimental differential cross-section data for the U-235 nucleus have yielded single level Breit-Wigner resonance parameters that fit, simultaneously, three nuclear cross sections of capture, fission, and total.
Value and limitations of transition-state theory (TST) are reviewed. TST analyses of the temperature dependence of the 'direct' reactions CH3 + CH3CHO yields CH4 + CH3CO(1) and O + CH4 yields OH + CH3(2) are presented in detail, and other examples of TST usefulness are recalled. Limitations are discussed for bimolecular processes in terms of 'complex' vs. 'direct' mechanisms. The reaction OH + CO yields CO2 + H is discussed in this context. Limitations for unimolecular processes seem to arise only for simple bond fission processes, and recent advances are noted.
In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.
Analyses of noble gases in whole rock samples of the howardites Bholghati and Kapoeta and grain-size separates of Kapoeta yield evidence for excesses of the Xe isotopes Xe-129 , Xe-131, Xe-132, Xe-134, and Xe-136 in a low-temperture component, similar to lunar excess fission Xe. Such a component may be able to provide chronometric information if the relative abundances of radioactive progenitors (I-129, Pu-244, and U-238) can be determined, but the isotopic spectra obtained are not sufficiently precise to do so. Eucritic clast BH-5 in Bholghati contains Xe produced in situ by the decay of Pu-244. Calculated fission Xe retention ages are 30-70 Ma after the formation of the solar system, consistent with the apparent presence of Sm-146 decay products. Both the clast and the matrix of Bholghati have K-Ar ages of about 2 Ga, suggesting a common thermal event at least that recently.
Recent public concern over the safety, cost, and environmental impact of the worldwide fission reactors has focused the attention of scientists and engineers towards perfecting fusion technology because it promises a much more environmentally acceptable 'clean' energy supply. The fusion reaction D-2 + He-3 yields p(14.7 MeV) + He-4(3.6 MeV) has long been recognized as an ideal candidate for producing commercially 'safer and cleaner' fusion power. Naturally occurring He-3 is scarce on earth; however, lunar regolith is a potential ore for He-3 because the high He-3 in solar wind has been implanted in the lunar regolith for more than 4 x 10(exp 9) years, along with other volatile species. The helium abundance in lunar soils is dependent not only on the maturity of soils (I(sub S)/FeO) but also on their mineralogy. The titanium-rich (ilmenite) lunar soils are important repositories for volatiles, which may be released by heating these soils up to approximately 700 C.
The separation of U from Tc and other problematic fission product elements like Mo and Ru, along with Sr, Zr, Cs, and Nd, has been achieved via the crystallization of uranyl nitrate hexahydrate (UNH). Rejection of technetium as pertechnetate anion ( 99 TcO 4 – ) is an especially important feature of this system, as it otherwise tends to follow U (VI) in extractive separations. It also raises the salient question regarding why this oxoanion cannot replace nitrate within the crystalline lattice of UNH. Results showed high-yield (>90 %), high-purity (>99 %) recovery of U as UNH from solutions containing 99 TcO 4 – by simple reduction of temperature from 60°C to 20°C. There was no observable interaction of 99 TcO 4 – with UO 2 2+ . The addition of other cations like, Sr 2+ , Zr 4+ , Cs + , and Nd 3+ , also did not form secondary, contaminant solid phases, leaving the > 99 % of the fission product elements in the mother liquor, while the U was recovered at > 90 %. Similarly, Mo and Ru, when added to the mixture, were shown to behave as the other fission-product elements, remaining in the mother liquor during crystallization. As a result, DFT calculations showed that, despite the higher binding strength of TcO 4 – , HMoO 4 – , and BiO 3 – with the UO 2 2+ cation compared to NO 3 – , the hydrogen-bonding network of the two coordinated ions and four waters of hydration in the UNH crystal structure is the driving force for the high specificity of this separation.
Here, a time-resolved x-ray scattering (TRXS) investigation of the photodissociation dynamics of gas-phase 2-iodothiophene molecules following 252 nm excitation is presented. Structural evolution of the molecule and dynamical information on the resulting photofragments were captured using femtosecond x-ray free-electron laser pulses. Two dissociation pathways were identified, arising via excitation to ππ* and (n/π)σ* states, respectively, yielding distinct interfragment recoil velocities of ∼6.4 Åps −1 and 17.0 Åps −1 . A comparison of asymptotic scattering data with simulated patterns indicates that the thiophene ring remains closed following dissociation at this wavelength. Modeling the experimental data yields a branching ratio of ∼3:1 in favor of the high velocity channel. These findings demonstrate the capability of TRXS to disentangle concurrent ultrafast reaction pathways and provide detailed structural insight into energy redistribution during photoinduced bond fission in complex molecular systems.
This work explores the parameter space of a cylindrical hybrid nuclear fuel in support of a z-pinch driven pulsed fission-fusion (PuFF) engine. 0D power balance and 1D burn wave calculations have been performed to explore the parameter space of hybrid cylindrical nuclear fuels. The boundary of minimal initial conditions needed to reach breakeven are found within the context of the model. The effects of initial conditions upon the yield at the end of burn wave expansion are also determined. The model is used to examine the minimum initial energies predicted to result in yields of a few MJ. The goal of this work is to guide fuel design, inform future models and experiments as well as look for the most efficient parameter space for ignition in a z-pinch driven hybrid nuclear reaction. The impact of initial parameters upon ignition, burn, and gain are discussed. It is found that a hybrid cylindrical target may breakeven with initial energy near the axis of approximately 4 to 6 MJ in lithium deuteride/uranium 235 fuel. It is also found that magnetic field can lower the threshold further of which the magnitude changes across the parameter space. The dual fusion/fission reactions are found to boost each other leading to lower initial driving energies needed to reach breakeven in the hybrid cylindrical nuclear fuel.
The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.
Abstract Separation of an individual heavy actinide from other actinides, lanthanides, and coproduced fission products is challenging not only because of their similarity in chemistry but also because the chemistry of heavy actinides is largely unknown. At present, the cation-exchange chromatography with α-hydroxyisobutyric acid (CX-AHIB) method is used to isolate milli- to picogram quantities of heavy actinides (i.e., 249Bk, 252Cf, 254Es, and 257Fm). This method allows simultaneous separation of these actinides; however, isolating a clean individual product with a high yield has proven challenging. The process is also very slow and labor-intensive and requires precise control of various chemical conditions, such as pH, temperature, and AHIB concentration. Developing a separation scheme for heavy actinides requires identifying their unique feature and then harnessing this feature in the separation process design. The unique characteristic of Bk4+ is that it does not adsorb onto anion exchange resin columns, unlike other tetravalent actinides. This article discusses what makes Bk unique and how this discovery led to a new method for separating Bk from adjacent actinides, lanthanides, and coproduced fission products. The method employed two different resin columns in tandem to separate unwanted actinides from 249Bk, followed by fine cleanup of 249Bk. The advantages of the new Bk method over the CX-AHIB method in Bk production are discussed, and the performance and robustness of the proposed method were assessed in two recent production campaigns.
The separation of 241 Am, 244 Cm and 249 Cf from rare earth elements using LN resin and lactate buffer solutions with diethylenetriaminepentaacetic acid is described with numerous columns studies. The elutions are based on TALSPEAK liquid–liquid extraction chemistry, and provide high yield (> 90%), rapid, simple separations of the trivalent actinides from the lanthanides, which is often challenging due to similarities in chemical behavior among the trivalent f-block elements. Further, all three actinides can be separated from rare earth elements, including from massless fission product samples and samples with small amounts (~ 1–5 mg) of stable lanthanides. Separations with no detectable overlap between the actinide and lanthanide elutions is possible with massless samples, and in samples with mass high separation factors can be achieved (Am/Eu: 171; Am/La: 10 6 ).
This work introduces the pin power reconstruction capability available in the Griffin reactor physics code. This capability is implemented in an unstructured mesh framework, and the methods introduced are applied to the 2D SIMBA reactor core, which has assemblies and pins arranged in a hexagonal lattice. Since this reactor has a non-Cartesian geometry and also operates in the thermal spectrum, a general approach to pin power reconstruction is adopted, where SPH-based equivalence is leveraged to preserve assembly-wise reaction rates, while computing full-core form functions to preserve pin-wise fission production rates within the fuel pins of the reactor core. In a 2D microreactor benchmark problem, this pin power reconstruction approach was shown to reproduce pin powers compared to the Serpent2 Monte Carlo code for fixed temperature conditions and control drum rotation angles, yielding a core-wide RMS error level of 0.6\% and a maximum absolute pin error of 2.3\%. In addition, a tabulated library of multigroup cross sections, SPH factors, and form functions was generated to demonstrate the applicability of pin power reconstruction to a thermal feedback problem. Finally, a control drum transient was successfully simulated, showcasing the application of pin power reconstruction in a transient multiphysics feedback problem.
The energy efficiency of heat engines (gas and steam turbines) for electricity production and propulsion is determined by the Carnot cycle and scales with operating temperature. Commercial nickel- and cobalt-based superalloys melt near 1,500 °C and rapidly lose mechanical strength beyond 1,000 °C. Refractory metals melt well above 2,000 °C but have inherent manufacturability challenges that are barriers to adoption, such as high ductile-to-brittle transition temperatures. Using density functional theory-guided design, we demonstrate tailored local lattice distortions that promote phase-stable, non-equiatomic refractory concentrated solid solutions with both high ductility and strength. Here, we exemplify this for single-phase, body-centred cubic Nb 4 Ta 4 V 3 Ti that exhibits castability, excellent room-temperature tensile yield strength (∼1 GPa) and ductility (approaching 20% uniform strain), and exceptional high-temperature tensile strength (500 MPa at 1,000 °C). These findings illustrate a path for designing materials that hold great potential for advancing next-generation technologies such as Generation IV fission reactors, first-generation fusion-plasma reactors, and more efficient gas turbines for electricity generation and propulsion.
Understand and Predict Radiation-Induced Iodine Speciation, Chemistry, and Transport in High-Temperature Molten Salts Award Number: DE-AC07-05ID1451 Gregory P. Holmbeck (Gregory.Holmbeck@inl.gov), Center for Radiation Chemistry Research, Idaho National Laboratory, 1955 N. Fremont Avenue Idaho Falls, ID, 83415, USA. Project Scope The goal of this Chemical and Materials Sciences to Advance Clean Energy Technologies and Low-Carbon Manufacturing project is to understand and predict the radiation-induced speciation, chemistry, and transport of fission product iodine in the triumvirate extremes of high-temperature, ionizing radiation, and corrosive molten salts. This missing fundamental information is critical for the accelerated development and deployment of safe, clean nuclear energy based on molten salt reactor (MSR) and pyrochemical reprocessing technologies. The central hypothesis driving this research is, the radiation-induced conversion of iodide will yield an extensive suite of transient and steady-state iodine radiolysis products that will alter the bulk chemical and physical properties of the irradiated molten salt system—the speciation, distribution, and chemical transport of which will be dictated by the composition and the availability of multivalent metal cations and metal alloy interfaces. To test this hypothesis, this project initiated three synergistic Research Objectives: (1) determine how the inclusion of iodine/iodide influences the chemical and physical properties of complex molten salt mixtures; (2) elucidate the speciation and fundamental chemical behavior of transient and steady-state iodine/iodide species formed by the irradiation of molten and solid salt mixtures; and (3) understand the influence of interfacial processes on determining the final disposition of iodine in high temperature molten salts, notably the structure and chemical speciation of iodine at metal-salt interfaces.
Proposed thorium-based nuclear fuel cycles are likely to require quantification and verification of 233 U within nuclear material. Because of their similar fission cross sections, active neutron nondestructive assay (NDA) systems may respond similarly to 233 U and 235 U. Traditional safeguards equipment has been optimized for 235 U and 238 U quantification associated with conventional uranium/plutonium fuel cycles and may not be directly applicable to 233 U quantification when mixed with other actinides. This work used models of the large volume active well coincidence counter (LV-AWCC) at Oak Ridge National Laboratory to evaluate the performance of this neutron NDA system to differentiate fissile uranium isotopes. The models were developed to simulate NDA system performance in response to a number of triangular radiation signature training device sources within the central cavity or well. This work predicted that the LV-AWCC can effectively differentiate 233 U from 235 U in certain modes of operation. In active mode, the LV-AWCC with the cadmium liner results in different doubles count rates between the fissile isotopes for a given fissile uranium mass. Without the cadmium liner, the uranium isotopes provide a statistically indistinguishable doubles count rate response for the fissile masses considered in this work (up to approximately 150 g). The cadmium liner serves to harden the neutron interrogation spectrum, which better exploits the notable difference in the 233 U and 235 U fission cross sections at approximately 1 eV. In passive mode, the two fissile isotopes exhibit different doubles and singles count rates regardless of liner presence because the passive source strength of 233 U is approximately 2 orders of magnitude stronger than that of 235 U due to the shorter half-life and correspondingly higher (α, n) yield. We conclude that using neutron interrogation in the LV-AWCC, two measurements are needed to quantify 233 U content in mixed uranium items. The first measurement is used to determine the total fissile uranium mass using a mode that cannot distinguish fissile isotopes (i.e., where a similar response is observed for both fissile uranium isotopes such as active doubles without cadmium or using a thermal neutron interrogation source). In conclusion, the second measurement is used to determine the 233 U content by using a differentiating technique (e.g., passive doubles, passive doubles to singles ratio, active doubles with cadmium).
An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.
Reliable and rapid inspection techniques play a vital role in preventing illicit trafficking of special nuclear materials. Active interrogation systems using neutrons produced by portable, high-flux deuterium-deuterium or deuterium-tritium neutron generators are being actively developed as a secondary scanning tool for this purpose. In this study, a neutron noise analysis-based approach for detecting unshielded and shielded special nuclear materials by using a pulsed deuterium-tritium neutron generator was evaluated. Here, this approach analyzes the fluctuation of neutron counts. Its performance was quantified with regard to time-to-detection to achieve a minimum probability of detection of 99% and a probability of false alarm of less than 1% considering various amounts of special nuclear materials and different shielding configurations. It was demonstrated that this approach could detect 17 uranium slugs in 5 s given a neutron generator yield of 8.1 × 10 7 n/s. These slugs could be detected within a reasonable time frame (200 s) when they were shielded by 10.16 cm of high-density polyethylene. The results obtained using the neutron noise analysis approach were compared with those obtained using the commonly used differential die-away analysis technique, a sensitive technique for detecting the presence of fissile materials by utilizing the prompt fission neutrons produced when the source neutrons from a neutron generator are completely diminished. For example, the time to detect 2 unshielded uranium slugs was 2.1 s when using the differential die-away analysis technique; it increased to 93 s for the neutron noise analysis approach. Although the noise analysis-based approach exhibits an overall performance which is not as good as that of differential die-away, neutron noise provides an alternative method for effective detection of special nuclear materials.
Zircaloy-4 is an essential material for cladding structures within fission-based reactors. To explore the changes in properties measured on differing length scales, FIB-machined micro-scale tensile tests were performed on both irradiated and control groups of Zircaloy-4. This was correlated with tensile testing on femtosecond laser-machined meso-scale specimens. Pronounced size effects were found when varying specimen geometry. Increases in tensile geometry size were associated with a reduction in measured yield stress for both irradiated and unirradiated samples. Here, meso-scale testing found strength and strain values similar to that of bulk-scale testing.