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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Towards Realistic and High Fidelity Models for Nuclear Reactor Power Synthesis Simulation with Self-Powered Neutron Detectors

As presented in this report, a weighting function–based inferencing method is being applied to synthesize the power distribution in next-generation and university research reactors based on simulated self power neutron detector (SPND) responses. The overall goal is to assess the impacts of sensor uncertainty and true power distribution perturbations on the error in the synthesized power distribution. Regarding sensor uncertainty, the NuScale Small Modular Reactor (SMR) and the Westinghouse AP1000 serve as testbeds for analyzing the impact of varying the sensor uncertainty, as well as varying the number of sensors per sensor string in the reactor core. The reactor models are informed by Monte Carlo N-Particle (MCNP) neutron flux tallies. For the NuScale SMR and Westinghouse AP1000, the SPND response functions (i.e., the response of the SPNDs to individual segments of fuel) were determined homogeneously. Regarding an analysis of power distribution perturbation detection, the Texas A&M Testing, Research, Isotopes, General Atomics Reactor (TAMU TRIGA) reactor was used as a demonstration case with one particular arrangement of SPNDs; the response functions for this reactor model were determined heterogeneously, making this a uniquely high-fidelity demonstration of perturbation detection. Finally, SPND models generated in the Geometry and Tracking 4 (Geant4) code have been generated and tested for comparison with traditionally implemented analytical SPND models, with the intent for Geant4 integration with the full methodological framework. SPND current outputs as a function of distance from some fuel assembly segment in the NuScale SMR are compared with the analytically determined currents. Results from the sensor uncertainty simulations for the NuScale SMR and AP1000 indicate that the average error in the inferred power distribution on the fuel assembly segment level is reasonably low, being slightly less than the random uncertainty applied to all respective SPNDs in both cores. For example, if all SPNDs in the core have a random uncertainty of 5%, then the corresponding fuel assembly segment level error (i.e. difference between the true and inferred local power) is ~2–3%. However, the maximum error in the inferred power distribution on the fuel assembly segment level can be considerably high (>15%) when SPND random uncertainties start to exceed ~3%. In general, the average and maximum errors in the inferred power distribution were slightly higher in the AP1000 as opposed to the NuScale SMR for the sensor string configurations considered herein. Another result determined from analysis of the sensor uncertainty simulations was that increasing the number of SPNDs per string does not clearly reduce inferred power distribution error and can in fact make the error large in some cases; however, this assessment may skewed due to imposed iteration limits. Results from the perturbation detection demonstration using the high-fidelity TAMU TRIGA model indicate that, given the arrangement of 17 SPND strings and 4 SPNDs per string considered herein, there is a clear, provable ability to infer a localized Gaussian-type peak perturbation in the 3D power distribution. Such a perturbation was detected with an average fuel assembly segment level error of 0.19%, and the general visualization of the detected perturbation clearly indicates that the magnitude and shape were appropriately resolved. Finally, the electrical current output generated by the Geant4 modeled SPND indicates significant magnitude differences than the analytically modeled SPND, demonstrating the need for accurate SPND models which account for finite sensor geometry effects to inform the power synthesis work described herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Codes and standards for ceramic composite core materials for High Temperature Reactor applications

Fiber-reinforced ceramic matrix composites are attractive for high-temperature nuclear applications due to excellent thermal and mechanical properties as well as reasonable-to-outstanding radiation resistance. Over the past 20 years, the use of ceramic matrix composite applications expanded to many commercial non-nuclear industries as fabrication and application of the technologies mature. The ASME Boiler Pressure Vessel Code, under Section III Division 5, provides the design and construction rules for High Temperature Reactor components. It published the first rules for ceramic matrix composites to be used for reactor core components. The rules lay out the quality requirements together with the design and materials criteria for the use and application of silicon carbide- and carbon-based matrix material technologies. As with the established graphite rules, the ceramic composite material rules are structured in Subsection HH (from Section III), that addresses the criteria for class SN nonmetallic core components. The code rules rely heavily on the development and publication of standards for composite specification, classification, and testing of mechanical, thermal, and other properties. These test methods are developed in ASTM Committee C28 on Advanced Ceramics, with a current focus on ceramic composite tubes. This article describes the detail of the composites code, the design methodology and similarities to the graphite code, the guidance for the development of specifications for ceramic composites (for nuclear applications) including recent standard developments, and it mentions the next steps to support licensing aspects by validating the code with benchmarking data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for validation. Comparisons between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data are discussed in detail.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Thermal and hydraulic performance of volumetrically heated triply periodic minimal surface heaters

Advanced heat transfer surfaces, mainly Triply Periodic Minimal Surfaces (TPMSs) have great potential to increase heat transfer performance over traditional heat transfer surfaces due to their tortuous flow path and higher surface area-volume ratio. The emergence of additive manufacturing of nuclear fuel motivates experimental investigation of such advanced geometries in the context of a nuclear reactor core. Here, this study employed a novel method to explore the performance of gyroid and diamond TPMS geometries. Volumetrically heated heaters were additively manufactured from conductive polymer filament, and airflow was used to simulate the convective heat transfer from the core within a nuclear reactor. The mass flow and volumetric generation rates were varied, local Nusselt number correlations were obtained using embedded thermocouples to measure the solid and fluid temperature, and friction factor correlations were developed using differential pressure measurements. The gyroid TPMS presented a friction factor 1.5 times higher than the diamond TPMS and 90 times that of the rod bundle. TPMS geometries showed Nusselt numbers 8–10 times higher than that of the rod bundle, also maintaining colder internal temperatures, which suggested that they could sustain 250–275 % higher power density than the rod bundle for a given centerline temperature. Hence, having the potential to greatly increase the safety margins and power output of a nuclear reactor core with similar volume.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Indication of Reactor ν ¯ e Disappearance in the Double Chooz Experiment

The Double Chooz Experiment presents an indication of reactor electron antineutrino disappearance consistent with neutrino oscillations. A ratio of 0.944 ± 0.016 (stat) ± 0.040 (syst) observed to predicted events was obtained in 101 days of running at the Chooz Nuclear Power Plant in France, with two 4.25 GW th reactors. The results were obtained from a single 10 m 3 fiducial volume detector located 1050 m from the two reactor cores. The reactor antineutrino flux prediction used the Bugey4 measurement as an anchor point. The deficit can be interpreted as an indication of a non-zero value of the still unmeasured neutrino mixing parameter sin 2 2θ 13 . Analyzing both the rate of the prompt positrons and their energy spectrum we find sin 2 2θ 13 = 0.086 ± 0.041 (stat) ± 0.030 (syst), or, at 90% CL, 0.015 < sin 2 2θ 13 < 0.16.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Predictive Maturity of Multi-Scale Simulation Models for Fuel Performance

The project proposed to provide a Predictive Maturity Framework with its companion metrics that (1) introduce a formalized, quantitative means to communicate information between interested parties, (2) provide scientifically dependable means to claim completion of Validation and Uncertainty Quantification (VU) activities, and (3) guide the decision makers in the allocation of Nuclear Energy’s resources for code development and physical experiments. The project team proposed to develop this framework based on two complimentary criteria: (1) the extent of experimental evidence available for the calibration of simulation models and (2) the sophistication of the physics incorporated in simulation models. The proposed framework is capable of quantifying the interaction between the required number of physical experiments and degree of physics sophistication. The project team has developed this framework and implemented it with a multi-scale model for simulating creep of a core reactor cladding. The multi-scale model is composed of the viscoplastic self-consistent (VPSC) code at the meso-scale, which represents the visco-plastic behavior and changing properties of a highly anisotropic material and a Finite Element (FE) code at the macro-scale to represent the elastic behavior and apply the loading. The framework developed takes advantage of the transparency provided by partitioned analysis, where independent constituent codes are coupled in an iterative manner. This transparency allows model developers to better understand and remedy the source of biases and uncertainties, whether they stem from the constituents or the coupling interface by exploiting separate-effect experiments conducted within the constituent domain and integral-effect experiments conducted within the full-system domain. The project team has implemented this procedure with the multi- scale VPSC-FE model and demonstrated its ability to improve the predictive capability of the model. Within this framework, the project team has focused on optimizing resource allocation for improving numerical models through further code development and experimentation. Related to further code development, we have developed a code prioritization index (CPI) for coupled numerical models. CPI is implemented to effectively improve the predictive capability of the coupled model by increasing the sophistication of constituent codes. In relation to designing new experiments, we investigated the information gained by the addition of each new experiment used for calibration and bias correction of a simulation model. Additionally, the variability of ‘information gain’ through the design domain has been investigated in order to identify the experiment settings where maximum information gain occurs and thus guide the experimenters in the selection of the experiment settings. This idea was extended to evaluate the information gain from each experiment can be improved by intelligently selecting the experiments, leading to the development of the Batch Sequential Design (BSD) technique. Additionally, we evaluated the importance of sufficiently exploring the domain of applicability in experiment-based validation of high-consequence modeling and simulation by developing a new metric to quantify coverage. This metric has also been incorporated into the design of new experiments. Finally, we have proposed a data-aware calibration approach for the calibration of numerical models. This new method considers the complexity of a numerical model (the number of parameters to be calibrated, parameter uncertainty, and form of the model) and seeks to identify the number of experiments necessary to calibrate the model based on the level of sophistication of the physics. The final component in the project team’s work to improve model calibration and validation methods is the incorporation of robustness to non-probabilistic uncertainty in the input parameters. This is an improvement to model validation and uncertainty quantification stemming beyond the originally proposed scope of the project. We have introduced a new metric for incorporating the concept of robustness into experiment-based validation of numerical models. This project has accounted for the graduation of two Ph.D. students (Kendra Van Buren and Josh Hegenderfer) and two M.S. students (Matthew Egeberg and Parker Shields). One of the doctoral students is now working in the nuclear engineering field and the other one is a post-doctoral fellow at the Los Alamos National Laboratory. Additionally, two more Ph.D. students (Garrison Stevens and Tunc Kulaksiz) who are working towards graduation have been supported by this project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1975

Nuclear system that burns its own wastes shows promise

A nuclear fission energy system, capable of eliminating a significant amount of its radioactive wastes by burning them, is described. A theoretical investigation of this system conducted by computer analysis, is based on use of gaseous fuel nuclear reactors. Gaseous core reactors using a uranium plasma fuel are studied along with development for space propulsion.

Atchison, K.↗
DOE OSTI2016

Adaptive Sparse-Grid Stochastic Collocation Uncertainty Quantification Convergence for Multigroup Diffusion

Advanced methods in uncertainty quantification for numerical models in computational physics continue to gain acceptance in nuclear modeling. Previously, the efficiency of sparse-grid stochastic collocation in comparison with Monte Carlo for uncertainty quantification through convergence studies was demonstrated. An adaptive method for anisotropic sparse grid collocation and its convergence compared to previous efforts is considered. The physical system in consideration is a two-dimensional quarter-core reactor, consisting of 5 materials distributed in 121 regions. The two-group neutron diffusion criticality approximation is solved. Vacuum boundaries are applied on the top and right, and reflecting boundaries on the bottom and left. The material cross sections, neutron multiplication factors, and diffusion coefficients are potential sources of uncertainty. Uniformly-distributed uncertainty within 5% of the reference values is introduced for material properties. The distribution of the uncertain parameters make up the uncertainty space Γ included in R{sup N}, where N is the number of uncertain parameters. The input parameter uncertainties are independently distributed. As expected, the convergence of Monte Carlo on the benchmark as a function of computational solves is nearly linear and lethargic compared to the other methods. The static isotropic index set, in which each input dimension is treated with equal importance, converges much more efficiently than Monte Carlo, but lacks the benefit of emphasizing some dimensions over others. The anisotropic case is most efficient, where first-order sensitivity information was used to determine anisotropic weights for polynomials to use in the polynomial chaos expansion. Interestingly, the adaptive algorithm converges slightly less efficiently than the anisotropic case, while still performing better than the isotropic case. Because the adaptive case has to determine the importance of each input dimension, it occasionally predicts inaccurately and wastes computation solves compared to the more ideal anisotropic case. However, in circumstances when an ideal anisotropy is not understood a priori, the adaptive algorithm may be optimally efficient in representing the original model. Given the potential loss of efficiency when using a poorly-chosen anisotropic weighting, the adaptive algorithm may be an effective choice for initial uncertainty quantification in problems with a low-dimensionality input space. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Benchmark for evaluation and validation of reactor simulations (BEAVRS)

Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Systems Engineering (SE) Approach of Developing A Passive Safety Injection System (PSIS) to Cope with APR1400 Extended Station Black-Out (SBO) Accident Using RELAP5

On 11 March 2011 the power supply and cooling of the three Fukushima Daiichi reactors were disabled by a major earthquake followed by a relatively high tsunami waves. The earthquake damaged the electrical grid which is the normal power supply for the reactors, then the tsunami waves disabled the emergency diesel generators of the Fukushima Daiichi reactors, which left the plant with no alternating current (AC) source. This condition is called 'Station Black-Out'. The reactors were tripped immediately, but due to the nuclear physics phenomena, the reactor cores were still producing thermal power, which is generated by the decay of radioactive fission products in the reactor core, known as decay heat. The decay heat had not been circulated to an external heat sink and eventually the accumulative heat produced in the core led to damage to the reactor cores. After the Fukushima Daiichi accident, Design Extension Condition (DEC) such as total loss of feed water (TLOFW), and prolonged Station Black-Out (SBO) attracted the attention of nuclear experts. The attention focuses on the defense in depth concept, redundancy and sustainability of the systems, and development of the passive safety systems, which rely on natural physics laws. The topic of this paper is to discuss the possibility of attaching a Passive Safety Injection System (PSIS) to the APR1400. The PSIS is a safety system which provides adequate cooling water to the steam generators after the loss of the Auxiliary Feed Water System (AFWS) during an extended SBO accident. The PSIS will be available while the power supply of the plant is being recovered and the active safety systems take over the functionality. The anticipated PSIS will be automatically actuated to enhance the engineered safety feature and to reduce the human errors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the premier irradiation facilities in the world. The ATR hosts experiments from a variety of laboratories, companies, and disciplines. Because of its flexibility in providing customized irradiation conditions to experiments, such as coolant chemistry, temperature, and reactor power, analysis for experiments, both from a programmatic and safety perspective, can be a challenging endeavor. Because of these experiment variables, there is always an effort to improve upon the existing analysis methodology in order to provide the sponsor of any given experiment with the most accurate predictions possible. The paper discusses the effort to improve upon the methodology for loading the ATR driver core for experiment analysis. Experiment analysis at INL is generally broken up into three separate disciplines: neutronics, thermal-hydraulics, and structural. The focus of this paper is neutronics analysis. After the experiment has a preliminary design, the neutronics analyst will create a model in the Monte Carlo N-Particle (MCNP) transport code to calculate fuel and material heating rates from MCNP tallies and provide those to the thermal-hydraulics analyst, and calculate reactivity worths, perturbation on the axial fission profile of the ATR driver core by the experiment, void worths, and experiment backup worths. Also, there are usually several programmatic parameters that the experiment sponsor requests that the neutronics analyst calculates. These can include heat generation rates, burnup and displacements per atom (DPA). In order to accurately analyze these parameters, an important input into the physics analyst's MCNP model is the loading of the ATR driver core. The ATR driver core consists of 40, 19-plate highly-enriched uranium (HEU) fuel elements, arranged in a serpentine shape, as pictured in Figure 1. The ATR driver core is responsible for the vast majority of the thermal power of the reactor and the nuclear interactions with the experiments being irradiated in ATR. A unique feature of ATR is the 'power tilt' across the reactor. A power tilt means that the Northwest, Northeast, Southwest, and Southeast Lobes of ATR frequently operate at different powers, providing different irradiation conditions for the experiments in and around the lobes of the reactor. In order to achieve lobe power splits, different fuel elements for each of the 40 positions are chosen by ATR Reactor Engineering so that lobe power splits can be maintained during an operating cycle. A typical fuel loading for an ATR cycle consists of 14 to 18 new fuel elements and 2 to 26 depleted fuel elements. These loadings are typically chosen from three different types of fuel elements. All three ATR fuel element types share a common design, spelled out in Chapter 4 of the ATR Safety Analysis Report (SAR). The fuel element meat is HEU UAlx dispersion fuel sandwiched between aluminum cladding with a length of 121.92 cm (4 ft), with 19 plates per fuel element. The three variations of this common design are designated by the first two letters in their serial numbers: XA, YA, and NB. XA elements are most commonly used. These elements have nominally 1075 g of uranium-235 and 0.66 g of boron-10. The NB elements have the same uranium-235 loading as the XA elements, but no boron in any plates. These elements are used when excess reactivity is needed. The YA elements have the same loading as the XA elements, except that the outermost plate (Plate 19) has no boron or uranium. These elements are used to mitigate aged reflector damage. In order to accurately model the ATR driver core, all three types of fuel elements must be accounted for in the MCNP model. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Calculation of the neutron diffusion equation by using Homotopy Perturbation Method

The distribution of the neutrons in a nuclear fuel element in the nuclear reactor core can be calculated by the neutron diffusion theory. It is the basic and the simplest approximation for the neutron flux function in the reactor core. In this study, the neutron flux function is obtained by the Homotopy Perturbation Method (HPM) that is a new and convenient method in recent years. One-group time-independent neutron diffusion equation is examined for the most solved geometrical reactor core of spherical, cubic and cylindrical shapes, in the frame of the HPM. It is observed that the HPM produces excellent results consistent with the existing literature.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2013

Interim Safe Storage of Plutonium Production Reactors at the US DOE Hanford Site - 13438

Nine plutonium production reactors located on DOE's Hanford Site are being placed into an Interim Safe Storage (ISS) period that extends to 2068. The Environmental Impact Statement (EIS) for ISS [1] was completed in 1993 and proposed a 75-year storage period that began when the EIS was finalized. Remote electronic monitoring of the temperature and water level alarms inside the safe storage enclosure (SSE) with visual inspection inside the SSE every 5 years are the only planned operational activities during this ISS period. At the end of the ISS period, the reactor cores will be removed intact and buried in a landfill on the Hanford Site. The ISS period allows for radioactive decay of isotopes, primarily Co-60 and Cs-137, to reduce the dose exposure during disposal of the reactor cores. Six of the nine reactors have been placed into ISS by having an SSE constructed around the reactor core. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Physics and Environmental Assessments for a Heavy-Water Moderated, Molten Uranium Reactor Concept

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 fuel tubes contain molten natural uranium as both the fuel and the coolant instead of fuel rods and water coolant. 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 assessment will examine three perspectives: reactor physics, regional grid management, and global resource utilization and impact. This assessment will examine three perspectives: reactor physics, regional grid management, and global resource utilization and impact. The proposed reactor type can generate much of the future CO{sub 2} emission free electric power demand with a low impact on total uranium required to be mined, zero heavy metal waste produced, zero enrichment capability required, and capacity (with thermal storage) to meet fluctuating demand. There are engineering issues to be addressed before such a reactor can be built, but thanks to the demand for high efficiency gas turbines and other high temperature applications, high temperature materials and technology are developing rapidly. An operating temperature of 1400 degrees C does not seem unreasonable in the near future. The implied advantages of this reactor type compared to other existing or proposed reactors suggest that further work on the issues raised would be well worthwhile. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Conformal core cooling and containment structure

A nuclear reactor includes a pressure vessel and a nuclear reactor core comprising fissile material disposed inside the pressure vessel at the bottom of the pressure vessel. A secondary core containment structure includes a containment basket comprising insulation with a maximum stable temperature of at least 2200K cladded by steel. The bottom of the pressure vessel and the nuclear reactor core are disposed inside the containment basket with the containment basket spaced apart from the bottom of the pressure vessel by a clearance gap. The containment basket may comprise zirconia insulation cladded by steel. In some embodiments the clearance gap between the containment basket and the bottom of the pressure vessel is no larger than one meter. The secondary core containment structure may further comprise conduits arranged to inject water into the clearance gap between the containment basket and the bottom of the pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Conformal core cooling and containment structure

A nuclear reactor includes a pressure vessel and a nuclear reactor core comprising fissile material disposed inside the pressure vessel at the bottom of the pressure vessel. A secondary core containment structure includes a containment basket comprising insulation with a maximum stable temperature of at least 2200K cladded by steel. The bottom of the pressure vessel and the nuclear reactor core are disposed inside the containment basket with the containment basket spaced apart from the bottom of the pressure vessel by a clearance gap. The containment basket may comprise zirconia insulation cladded by steel. In some embodiments the clearance gap between the containment basket and the bottom of the pressure vessel is no larger than one meter. The secondary core containment structure may further comprise conduits arranged to inject water into the clearance gap between the containment basket and the bottom of the pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗