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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2010

FUEL CYCLE ISOTOPE EVOLUTION BY TRANSMUTATION DYNAMICS OVER MULTIPLE RECYCLES

Because all actinides have the ability to fission appreciably in a fast neutron spectrum, these types of reactor systems are usually not associated with the buildup of higher mass actinides: curium, berkelium and californium. These higher actinides have high specific decay heat power, gamma and neutron source strengths, and are usually considered as a complication to the fuel manufacturing and transportation of fresh recycled transuranic fuel. This buildup issue has been studied widely for thermal reactor fuels. However, recent studies have shown that the transmutation physics associated with "gateway isotopes" dictates Cm-Bk-Cf buildup, even in fast burner reactors. Assuming a symbiotic fuel relationship with light water reactors (LWR), Pu-242 and Am-243 are formed in the LWRs and then are externally fed to the fast reactor as part of its overall transuranic fuel supply. These isotopes are created much more readily in a thermal than in fast spectrum systems due to the differences in the fast fission (i.e., above the fission threshold for non-fissile actinides) contribution. In a strictly breeding fast reactor this dependency on LWR transuranics would not exist, and thus avoids the introduction of LWR derived gateway isotopes into the fast reactor system. However in a transuranic burning fast reactor, the external supply of these gateway isotopes behaves as an external driving force towards the creation and build-up of Cm-Bk-Cf in the fuel cycle. It was found that though the Cm-Bk-Cf concentration in the equilibrium fuel cycle is dictated by the fast neutron spectrum, the time required to reach that equilibrium concentration is dictated by recycle, transmutation and decay storage dynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Differential Die-Away Instrument: Report on Initial Simulations of Spent Fuel Experiment

New Monte Carlo simulations of the differential die-away (DDA) instrument response to the assay of spent and fresh fuel helped to redefine the signal-to-Background ratio and the effects of source neutron tailoring on the system performance. Previously, burst neutrons from the neutron generator together with all neutrons from a fission chain started by a fast fission of 238 U were considered to contribute to active background counts. However, through additional simulations, the magnitude of the 238 U first fission contribution was found to not affect the DDA performance in reconstructing 239 Pu eff . As a result, the newly adopted DDA active background definition considers now any neutrons within a branch of the fission chain that does not include at least one fission event induced by a thermal neutron, before being detected, to be the active background. The active background, consisting thus of neutrons from a fission chain or its individual branches composed entirely of sequence of fast fissions on any fissile or fissionable nuclei, is not expected to change significantly with different fuel assemblies. Additionally, while source tailoring materials surrounding the neutron generator were found to influence and possibly improve the instrument performance, the effect was not substantial.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Studies of heavy ion reactions around Coulomb barrier

The angular distributions of fission fragments for the {sup 32}S+{sup 184}W reaction near Coulomb barrier energies were measured. The experimental fission excitation function is obtained. The measured fission cross sections are decomposed into fusion-fission, quasifission and fast fission contributions by the dinuclear system model. The hindrance to complete fusion both at small and large collision energies is explained. The fusion excitation functions of {sup 32}S+{sup 90,96}Zr in an energy range from above to below Coulomb barrier were measured and analyzed. The obvious effect of positive Q-value neutron transfers has been observed in the {sup 32}S+{sup 96}Zr system. In addition, the excitation functions of quasielastic scattering at a backward angle have been measured for the systems of {sup 16}O+{sup 208}Pb, {sup 196}Pt, {sup 184}W, and {sup 154,152}Sm at energies well below the Coulomb barrier. Considering the coupling effects, the extracted diffuseness parameters are close to the values extracted from the systematic analysis of elastic and inelastic scattering data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

ADRIANA project: Identification of research infrastructures for the SFR, within the frame of European industrial initiative for sustainable nuclear fission

Fast neutron reactors have a large potential as sustainable energy source. In particular, Sodium Fast Reactors (SFR) with a closed fuel cycle and potential for minor actinide burning may allow minimization of volume and heat load of high level waste and provide improved use of natural resources (as compared to only 1% energy recovery in the current once-through fuel cycle, with Thermal Reactors, such as EPR). The coordinating action ADRIANA (Advanced Reactor Initiative And Network Arrangement) has been initiated to set up a network dedicated to the construction and operation of research infrastructures in support of developments for the European Industrial Initiative for sustainable nuclear fission. The Project sets these objectives for the following reactor systems and related technologies: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR, including very high temperature technologies), Instrumentation, diagnostics and experimental devices, Irradiation facilities and hot laboratories, Zero power reactors. Among the fast reactor systems, the sodium cooled reactor has the most comprehensive technological basis as result of the experience gained from worldwide operation of several experimental, prototype and commercial size reactors, since the forties (see Appendix I). This concept is currently considered as the reference, within the European strategy. Innovations are needed to further enhance safety, reduce capital cost and improve efficiency reliability and operability, making the Generation IV SFR an attractive option for electricity production. Currently, in France, a moderate (500 to 600 MWe) power demonstrator named ASTRID (Advanced Sodium Test Reactor for Industrial Demonstration) has been proposed and endorsed by EU. Presently, the reference configuration is a pool concept. General R and D needs have been identified and experimental facilities required to satisfy these needs have been listed for the following domains: material and coolant, thermal-hydraulics and safety, severe accidents, systems and components validation, instrumentation and Education and Training. Then, the contributors have detailed the different types of experimental facilities, existing as well as lacking facilities. A ranking procedure of the facilities has been carried out, then a list of priorities and a road-map for investment has been established. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Advantages and applications of megawatt-sized heat-pipe reactors

Recently, worldwide interest in nuclear energy has focused on small reactors (10 to 300 MWe) to address emerging energy needs in remote locations. These designs are new to varying degrees but share similar approaches and common weaknesses with regard to primary heat rejection that differ little from reactor designs of the late 1950's. Here, an innovative concept, heat-pipe reactors, is discussed. The concept is unique in its simplicity and potential for safe, affordable, and reliable energy. Given the potential for reactors to meet worldwide energy needs and the pivotal role of heat rejection in overall reactor safety, the potential societal impact of this type of innovation is substantial. Heat-pipe-cooled, fast-spectrum reactors have been proposed for government applications requiring a robust, reliable, remotely controlled system with capacity much less than 1 MWe; however, they have not been designed for power ranges greater than 1 MWe. Los Alamos National Laboratory has initiated a study to design heat-pipe-cooled, fast-fission reactors and to generate a point design of a > 10-MWe-class machine suitable for next-generation compact reactors at remote locations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction are essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Optimal Arrangements of Tetracene Molecule Pairs for Fast Singlet Fission

In search for a qualitative understanding of the effects of molecular packing on singlet fission (SF) rate, a simplified version of the frontier orbital model is described and illustrated on a pair of tetracene molecules. To identify all favorable physically accessible pair geometries, all significant local maxima of the square of the electronic matrix element for SF have been located within the six-dimensional space of possible arrangements of two rigid bodies, using a grid of over 4.7 × 10 8 pair geometries. Those at which the molecules interpenetrate were excluded using a hard-sphere model. The effects of intermolecular interaction on the SF energy balance and thus its rate constant k SF were approximated using Marcus theory at each of the maxima using the same simplified version of the frontier orbital model. Starting at these local maxima, the pair geometries were optimized for maximum k SF and the 21 best are reported along with their computed Davydov splitting and triplet biexciton binding energies. The optimal pair structures at the resulting maxima follow qualitative rules published previously and further elaborated here.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2011

On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Short-lived fission product measurements from >0.1 MeV neutron-induced fission using boron carbide.

A boron carbide shield was designed, custom fabricated, and used to create a fast fission energy neutron spectrum. The fissionable isotopes 233, 235, 238U, 237Np, and 239Pu were separately placed inside of this shield and irradiated under pulsed conditions at the Washington State University 1 MW TRIGA reactor. A unique set of fission product gamma spectra were collected at short times (4 minutes to 1 week) post-fission. Gamma spectra were collected on single-crystal high purity germanium detectors and on Pacific Northwest National Laboratory's (PNNL's) Direct Simultaneous Measurement (DSM) system composed of HPGe detectors connected in coincidence. This work defines the experimental methods used to produce and collect the gamma data, and demonstrates the validity of the measurements. It is important to fully document this information so the data can be used with high confidence for the advancement of nuclear science and non-proliferation applications. The gamma spectra collected in these and other experiments will be made publicly available at https://spcollab.pnl.gov/sites/gammadata or via the link at http://rdnsgroup.pnl.gov. A revised version of this publication will be posted with the data to make the experimental details available to those using the data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Effects of fission yield data in the calculation of antineutrino spectra for U 235 ( 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 5–7 MeV, 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 0–7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Lastly, 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.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Final design for thermal epithermal experiments (TEX) with ZPPR plutonium/aluminium plates with polyethylene and tantalum - 13592

Lawrence Livermore National Laboratory has completed the final design for the first phase of the Thermal Epithermal experiments (TEX), which focused on critical configurations composed of plutonium/aluminum Zero Power Physics Reactor (ZPPR) plates moderated by polyethylene with and without a tantalum diluent. {sup 239}Pu and {sup 240}Pu were identified as being the number one and two nuclear data need by the nuclear criticality safety and nuclear data communities, with special emphasis placed on performance issues of the Pu cross sections in the intermediate energy region. Preliminary TEX design recommended tantalum as the first diluent to pursue for final design, as it represented a high priority data need, showed the highest sensitivity and largest contribution to uncertainty with the Pu ZPPR system, and a large number of high quality Ta plates from the ZPPR inventory in Idaho was readily available. The goals of the new critical experiments are 1) to create assemblies that span a wide range of fission energy spectra, from thermal (below 0.625 eV), through the intermediate (0.65 eV to 100 keV), and to fast energies (above 100 keV), and 2) to easily modify the assemblies to include high priority materials like tantalum. A total of ten critical configurations were designed as part of CED-2. Five TEX experiments (Experiments 1-5) were designed to establish baseline configurations with the Pu ZPPR plates covering the thermal, intermediate, and fast fission energy regimes. For all five experiments, the ZPPR plates will be arranged in layers of 24 plates (6 plates by 4 plates), resulting in approximately a 12 inch by 12 inch footprint. Multiple layers will be stacked together with varying thicknesses of interspersed polyethylene placed between the layers to fine tune the neutron spectrum of the assembly. The other five experiments (Experiments 6-10) will be similar to the five baseline experiments, except that a tantalum layer will be placed next to each plutonium plate layer, thus diluting the stack and allowing for tests of the neutron cross section for tantalum. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Advanced Monte Carlo modeling of prompt fission neutrons for thermal and fast neutron-induced fission reactions on {sup 239}Pu

Prompt fission neutrons following the thermal and 0.5 MeV neutron-induced fission reaction of {sup 239}Pu are calculated using a Monte Carlo approach to the evaporation of the excited fission fragments. Exclusive data such as the multiplicity distribution P({nu}), the average multiplicity as a function of fragment mass {nu}-bar(A), and many others are inferred in addition to the most used average prompt fission neutron spectrum {chi}(E{sub in},E{sub out}), as well as average neutron multiplicity {nu}-bar. Experimental information on these more exclusive data help constrain the Monte Carlo model parameters. The calculated average total neutron multiplicity is {nu}-bar{sub c}=2.871 in very close agreement with the evaluated value {nu}-bar{sub e}=2.8725 present in the ENDF/B-VII.0 library. The neutron multiplicity distribution P({nu}) is in very good agreement with the evaluation by Holden and Zucker. The calculated average spectrum differs in shape from the ENDF/B-VII.0 spectrum, evaluated with the Madland-Nix model. In particular, we predict more neutrons in the low-energy tail of the spectrum (below about 300 keV) than the Madland-Nix calculations, casting some doubts on how much scission neutrons contribute to the shape of the low-energy tail of the spectrum. The spectrum high-energy tail is very sensitive to the total kinetic energy distribution of the fragments as well as to the total excitation energy sharing at scission. Present experimental uncertainties on measured spectra above 6 MeV are too large to distinguish between various theoretical hypotheses. Finally, comparisons of the Monte Carlo results with experimental data on {nu}-bar(A) indicate that more neutrons are emitted from the light fragments than the heavy ones, in agreement with previous works.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2010

Induced-Fission Imaging of Nuclear Material

This paper presents initial results from development of the induced-fission imaging technique, which can be used for the purpose of measuring or verifying the distribution of fissionable material in an unopened container. The technique is based on stimulating fissions in nuclear material with 14 MeV neutrons from an associated-particle deuterium-tritium (D-T) generator and counting the subsequent induced fast fission neutrons with an array of fast organic scintillation detectors. For each source neutron incident on the container, the neutron creation time and initial trajectory are known from detection of the associated alpha particle of the d + t → α + n reaction. Many induced fissions will lie along (or near) the interrogating neutron path, allowing an image of the spatial distribution of prompt induced fissions, and thereby fissionable material, to be constructed. A variety of induced-fission imaging measurements have been performed at Oak Ridge National Laboratory with a portable, low-dose D-T generator, including single-view radiographic measurements and three-dimensional tomographic measurements. Results from these measurements will be presented along with the neutron transmission images that have been performed simultaneously. This new capability may have applications to a number of areas in which there may be a need to confirm the presence or configuration of nuclear materials, such as nuclear material control and accountability, quality assurance, treaty confirmation, or homeland security applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Evaluation of HEU-Beryllium Benchmark Experiments to Improve Computational Analysis of Space Reactors

An assessment was previously performed to evaluate modeling capabilities and quantify preliminary biases and uncertainties associated with the modeling methods and data utilized in designing a nuclear reactor such as a beryllium-reflected, highly-enriched-uranium (HEU)-O 2 fission surface power (FSP) system for space nuclear power. The conclusion of the previous study was that current capabilities could preclude the necessity of a cold critical test of the FSP; however, additional testing would reduce uncertainties in the beryllium and uranium cross-section data and the overall uncertainty in the computational models. A series of critical experiments using HEU metal were performed in the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. Of the hundreds of experiments, three were identified as fast-fission configurations reflected by beryllium metal. These experiments have been evaluated as benchmarks for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). Further evaluation of the benchmark experiments was performed using the sensitivity and uncertainty analysis capabilities of SCALE 6. The data adjustment methods of SCALE 6 have been employed in the validation of an example FSP design model to reduce the uncertainty due to the beryllium cross section data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Handbook of the Materials Properties of FeCrAl Alloys For Nuclear Power Production Applications

FeCrAl alloys are a class of alloys that have seen increased interest for nuclear power applications including as accident tolerant fuel cladding, structural components for fast fission reactors, and as first wall and blanket structures for fusion reactors. FeCrAl alloys are under consideration for these applications due to their inherent corrosion resistance, stress corrosion cracking resistance, radiation-induced swelling resistance, and high temperature oxidation resistance. A substantial amount of research effort has been completed to design, develop, and begin commercial scaling of FeCrAl alloys for nuclear power applications over the past half a century. These efforts have led to the development of an extensive database on material properties and process knowledge for FeCrAl alloys but not within a consolidated format. The following report is the first edition of a materials handbook to consolidate the state-of-the-art on FeCrAl alloys for nuclear power applications. This centralized database focuses solely on wrought FeCrAl alloys, oxide dispersion strengthened alloys, although discussed in brief, are not covered. Where appropriate, recommendations for applications of the data is provided and current knowledge gaps are identified.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Symmetry energy effects on fusion cross sections

We investigate the reaction path followed by heavy ion collisions with exotic nuclear beams at low energies. We focus on the interplay between reaction mechanisms, fusion vs. breakup (fast-fission, deep-inelastic), that in exotic systems is expected to be influenced by the symmetry energy term at densities around the normal value. The evolution of the system is described by a stochastic mean-field transport equation, where two parametrizations for the density dependence of symmetry energy (Asysoft and Asystiff) are implemented, allowing one to explore the sensitivity of the results to this ingredient of the nuclear interaction. The method described here, based on the event-by-event evolution of phase-space quadrupole collective modes, nicely allows us to extract the fusion probability at relatively early times, when the transport results are reliable. Fusion probabilities for reactions induced by {sup 132}Sn on {sup 64,58}Ni targets at 10 A MeV are evaluated. We obtain larger fusion cross sections for the more n-rich composite system, and, for a given reaction, in the Asysoft choice. Finally a collective charge equilibration mechanism (the dynamical dipole) is revealed in both fusion and breakup events, depending on the stiffnsess of the symmetry term just below saturation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Heat Transfer in a Molten Natural Uranium Nuclear Reactor With Thermal Storage

Previous papers describe a method for controlling a thermal nuclear reactor by moderator displacement between the reactor core and the reflector. The low neutron loss and high control range of this method enable a design for a molten natural uranium fueled and cooled heavy water moderated reactor. The design is geometrically similar to CANDU (Canadian Deuterium Uranium) reactors but with fewer, larger, vertically oriented fuel tubes. The large fuel tubes have a mixed spectrum which is largely thermal at the periphery of the tubes with a very hard spectrum in the large central region of the tube. This configuration provides a fast fission rate of around 15% (mostly of U-238) and an initial conversion ratio of around 1.15. It is a hypothetical reactor because it is unclear what materials could actually be used to fabricate the fuel tubes and other primary coolant plumbing. The molten uranium serves both as the fuel and the primary coolant and as such is continuously circulated through the fuel tubes in the reactor core, the primary heat exchange, and back to the the reactor core. This means that at any point in time all the fuel has the same isotopic configuration no matter where it is in the reactor or heat exchanger. Over time the isotopic configuration changes, but it changes uniformly for all the fuel. Uranium melts at 1135 deg. C, so a proposed temperature range for the circulating molten uranium is 1200 deg. C at the exit of the heat exchanger and entry to the core fuel tubes and 1400 deg. C at the exit from the core fuel tubes and the entry into the heat exchanger. The advantages of this design include: 1. Reduced nuclear proliferation concern because of natural uranium fuel. 2. High conversion ratio and large fuel mass provides 30 to 100 year fuel life. 3. No heavy metal waste because fissile content increases over the fuel life. 4. High output temperatures are suitable for high efficiency turbine generators. This paper discusses the heat transfer in the molten natural uranium fueled reactor described above and in the cited paper in the simple case and when the reactor is combined with a thermal storage unit which is composed of ordinary salt (NaCl). The proposed molten uranium fueled and cooled nuclear reactor works particularly well when coupled with a molten salt heat storage unit, a process heat user, and high Efficiency air-Brayton cycle turbine generators which can also add energy to the hot air output of the tin/air heat exchanges by burning some natural gas or hydrogen. The combination can provide highly variable electrical output to the grid to accommodate large daily and seasonal mismatches between electrical demand and supply from variable sources such as wind and solar. Future work includes modeling of the temperatures throughout the molten salt heat storage unit under varying charge and discharge cycles.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗