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DOE OSTI2026

Experimental investigation of a closed vapour box module for a divertor-like configuration in Magnum-PSI

Efficient management of extreme heat fluxes in the divertor region to extend the lifetime of the components remains a critical challenge for the realization of nuclear fusion-based power plants. Among the alternative concepts explored for the divertor region, the use of liquid metals, particularly lithium, is of interest due its ability to dissipate the incoming plasma heat flux through the vapour shielding effect (VS). In this work, we experimentally investigated a ‘closed’ configuration of a dedicated Vapour Box Module (VBM) in the linear plasma device Magnum-PSI. The goal of the experiments is to simulate the vapour box divertor environment conditions and assess its performance in terms of power mitigation and redistribution and lithium confinement. Initial testing without Li demonstrated the efficacy of a closed VBM structure in inducing detachment via neutral gas accumulation. Apertures which enabled non-condensing gas to be effectively pumped while ensuring lithium condensed on the inner surfaces were therefore added. With a lithium capillary porous structure target used, lithium is directly vaporized by the plasma, forming a dense lithium vapour cloud that interacts with the incoming plasma. This resulted in a significant reduction of the target temperature of at least 48%, together with a temperature locking effect, a phenomenon typically observed in the VS regime. Lithium vapour confinement within the VBM was strongly correlated with the wall temperature. Relatively cold walls promoted Li re-condensation and therefore improved Li confinement, although with the expected trade-off of increased hydrogenic retention on lithium-wetted surfaces. As the wall temperature increased, the confinement efficiency decreased, consistent with reduced Li re-condensation and thermally activated Li–H chemistry and remobilization at the walls. Diagnostic measurements through embedded thermocouples and calorimetry revealed that lithium vaporization and re-condensation processes also playedsignificant roles in plasma power dissipation. The results advance the case for a closed divertor chamber with direct lithium evaporation from the strike-points as a viable method to manage divertor heat fluxes in future fusion reactors.

Romano, Fabio [Dutch Institute for Fundamental Ene↗

DOE OSTI2023

A Conceptual Design of the Reactor Cavity Cooling System for the Horizontal Compact High Temperature Gas Reactor (HC-HTGR)

Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed HC-HTGR in 3 years and support its commercialization as a safe, low-cost HTGR. Argonne National Laboratory (Argonne) is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. This report documents the design study to derive a conceptual design study of the RCCS for the HC-HTGR. It includes the identification of the functions and requirements of the HC-HTGR RCCS, design analyses including high-level design consideration and the calculations for optimizing design space of the system with supporting component-level analysis to inform the material selection and performance of the water panel, the description of the conceptual design of the HC-HTGR RCCS derived based on the analyses results, and performance evaluation of the conceptual RCCS for the HC-HTGR. A detailed concept of the RCCS has been identified and high-level system requirements has been developed for the HC-HTGR. Design space focusing on the natural circulation loop portion of the RCCS has been investigated to optimize the system performance. The initial baseline dimensions were firstly derived based on the scoping calculations. A component level design analysis was conducted for the water panel to inform the material selection and to assess its conduction performance. A preliminary system-level performance analysis was performed for the 1/8th of the compartment of the initial baseline design of the RCCS using RELAP5-3D. To improve the system thermal performance, the RCCS design has been updated by exploring various design options by design parametric analyses. Based on the results, the conceptual design of the RCCS for the HC-HTGR has been derived, which satisfies the target performance of ~1 MWt at the elevated vessel wall temperature conditions. Transient simulations were conducted for the conceptual RCCS design for the HC-HTGR under various operation modes and heat load conditions using RELAP5-3D. The system dynamics in different operating states was investigated and the system performance under transients of interest was evaluated. The results demonstrated the overall system feasibility that the RCCS design maintains structures temperatures lower than maximum allowable temperature with sufficient system inventory without any active heat removal in the design process with certain transients addressed. The HC-HTGR RCCS will have additional design updates of subsystems or optimization of the system components during the preliminary and final design phases. Since the entire plant has not been integrated yet, this delivered conceptual design is subject to changes for integration, that require additional conceptual design activities and Quality and Assurance implementation (Q&A). The performance assessment of the RCCS for the HC-HTGR will be then revisited and optimized to finalize the system design, and the RCCS integrated primary system analysis will be utilized to simulate selective accident scenarios of interest where efforts are currently undergoing in the project.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

NOTCH EFFECT ON CREEP-FATIGUE BEHAVIOR OF ALLOY 617 AT ELEVATED TEMPERATURE

High-temperature reactor structural components are often under the complex multiaxial creep-fatigue (CF) loading conditions throughout the lifetime because of geometric and/or metallurgical discontinuities and complex loading paths. To assess the multiaxial CF deformation behavior and to evaluate the CF design rules in the ASME BPVC Section III, Division 5, Subsection HB, Subpart B, experimental and numerical studies are performed on Alloy 617 at 950°C using notch specimen geometries under CF loading in this study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Quantifying the effect of hydride microstructure on zirconium alloys embrittlement using image analysis

Since hydride microstructure can significantly influence Zr alloy nuclear fuel cladding’s ductility, a new metric has been developed to quantify hydride microstructure in 2D micrographs and relate it to crack propagation. As cladding failure usually results from a hoop stress, this new metric, called the Radial Hydride Continuous Path (RHCP), is based on quantifying the continuity of brittle hydride particles along the radial direction of the cladding tube. Compared to previous metrics, this approach more closely relates to the propensity of a crack to propagate radially through the cladding tube thickness. The RHCP takes into account hydride length, orientation, and connectivity to choose the optimal path for crack propagation through the cladding thickness. The RHCP can therefore be more closely linked to hydride embrittlement of the Zr alloy material, thus creating a relationship between material structure, properties, and performance. The new definition, along with previously proposed metrics such as the Radial Hydride Fraction (RHF), the Hydride Continuity Coefficient (HCC), and the Radial Hydride Continuity Factor (RHCF), have been implemented and automated in MATLAB. These codes are made available with this publication as supplementary materials. These metrics were verified by comparing their predictions of hydride morphology against expected values in simple cases, and the implementation of the new metric was validated by comparing its predictions with manual measurements of hydride microstructure performed on ImageJ. The RHCP was also validated against experimental measurements of fracture behavior and it was shown to correlate with cladding failure better than previous metrics. The information provided by these metrics will help accurately assess cladding integrity during operation, transportation, and storage.

36 MATERIALS SCIENCE↗

DOE OSTI2020

The Novel Hybrid Ab Initio Model of High-Performance Structural Alloys Design for Fossil Energy Power Plants

The current project developed a novel HT-CALPHAD/DFT approach, which can quickly design new high-performance structural alloys for the application of FE power plants. The PI will mainly take charge of high-throughput DFT simulations and computational thermodynamics of the selected multicomponent alloy systems for the FE power plant applications. At the end of the project, a novel hybrid model based on high throughput CALPHAD/DFT simulations and computational thermodynamics will be developed to provide guidance on how to identify multi-component new high-performance structural alloys with much less computational effort needed. It will address the extensive computation time needed for DFT on the new alloys design. In addition, it will also address the well-known headache of DFT, i.e. how to make the accurate prediction of the high-temperature equilibria. The novel hybrid model the PI proposed will not only be applied to the design of high-performance structural alloys in FE power plants but in many different applications, such as nuclear reactors. This hybrid modeling approach includes four sections: 1) HT-CALPHAD modeling of Al-Co-Cr-Ni-Fe system with FCC and BCC phase. In this section, 3561 non-equiatomic compositions were randomly generated in order to investigate the phase stability of single FCC and BCC phases. Meanwhile, we proposed a data screening procedure to screen out the good candidates within these compositions, considering the temperature range, average density, and melting temperature, etc. 2) Investigation of FCC-Cr lattice stability in Fe-Cr system. We systematically assessed the reliability of FCC-Cr lattice stability derived by DFT and CALPHAD approaches. Meanwhile, the Fe-Cr binary system was chosen as a case study to verify the Cr lattice stability obtained by both approaches. 3) High-throughput DFT modeling on elastic properties of Al-Co-Cr-Ni-Fe systems. We predicted and established the FCC quinary elastic constant database of the Al-Co-Cr-Fe-Ni systems at 0K by using special quasi-random structure (SQS) approach. The predictions will start with pure elements of Al-Co-Cr-Fe-Ni system and will be continued with binaries, ternaries, quaternaries, and finally the quinary compositions. 3) Modeling of temperature-dependent elastic properties in Al-Co-Cr-Ni-Fe systems. In this part, we predict the thermal expansion coefficient and elastic stiffness coefficient as a function of temperature by applying quasiharmonic approximation. With this approach, the elastic properties of HEAs at elevate temperature can be estimated.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2013

Preliminary safety evaluation of an aircraft impact on a near-surface radioactive waste repository

The aircraft impact accident has become very significant in the design of a nuclear facilities, particularly, after the tragic September 2001 event, that raised the public concern about the potential damaging effects that the impact of a large civilian airplane could bring in safety relevant structures. The aim of this study is therefore to preliminarily evaluate the global response and the structural effects induced by the impact of a military or commercial airplane (actually considered as a 'beyond design basis' event) into a near surface radioactive waste (RWs) disposal facility. The safety evaluation was carried out according to the International safety and design guidelines and in agreement with the stress tests requirements for the security track. To achieve the purpose, a lay out and a scheme of a possible near surface repository, like for example those of the El Cabril one, were taken into account. In order to preliminarily perform a reliable analysis of such a large-scale structure and to determine the structural effects induced by such a types of impulsive loads, a realistic, but still operable, numerical model with suitable materials characteristics was implemented by means of FEM codes. In the carried out structural analyses, the RWs repository was considered a 'robust' target, due to its thicker walls and main constitutive materials (steel and reinforced concrete). In addition to adequately represent the dynamic response of repository under crashing, relevant physical phenomena (i.e. penetration, spalling, etc.) were simulated and analysed. The preliminary assessment of the effects induced by the dynamic/impulsive loads allowed generally to verify the residual strength capability of the repository considered. The obtained preliminary results highlighted a remarkable potential to withstand the impact of military/large commercial aircraft, even in presence of ongoing concrete progressive failure (some penetration and spalling of the concrete wall) of the impacted area. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Evaluation of Advanced Signal Processing Techniques to Improve Detection and Identification of Embedded Defects

By the end of 1996, 109 Nuclear Power Plants were operating in the United States, producing 22% of the Nation’s electricity [1]. At present, more than two thirds of these power plants are more than 40 years old. The purpose of the U.S. Department of Energy Office of Nuclear Energy’s Light Water Reactor Sustainability (LWRS) Program is to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the operating lifetimes of nuclear power plants (NPPs) beyond 60 years [2]. The most important safety structures in an NPP are constructed of concrete. The structures generally do not allow for destructive evaluation and access is limited to one side of the concrete element. Therefore, there is a need for techniques and technologies that can assess the internal health of complex, reinforced concrete structures nondestructively. Previously, we documented the challenges associated with Non-Destructive Evaluation (NDE) of thick, reinforced concrete sections and prioritized conceptual designs of specimens that could be fabricated to represent NPP concrete structures [3]. Consequently, a 7 feet tall, by 7 feet wide, by 3 feet and 4-inch-thick concrete specimen was constructed with 2.257-inch-and 1-inch-diameter rebar every 6 to 12 inches. In addition, defects were embedded the specimen to assess the performance of existing and future NDE techniques. The defects were designed to give a mix of realistic and controlled defects for assessment of the necessary measures needed to overcome the challenges with more heavily reinforced concrete structures. Information on the embedded defects is documented in [4]. We also documented the superiority of Frequency Banded Decomposition (FBD) Synthetic Aperture Focusing Technique (SAFT) over conventional SAFT when probing defects under deep concrete cover. Improvements include seeing an intensity corresponding to a defect that is either not visible at all in regular, full frequency content SAFT, or an improvement in contrast over conventional SAFT reconstructed images. This report documents our efforts in four fronts: 1) Comparative study between traditional SAFT and FBD SAFT for concrete specimen with and without Alkali-Silica Reaction (ASR) damage, 2) improvement of our Model-Based Iterative Reconstruction (MBIR) for thick reinforced concrete [5], 3) development of a universal framework for sharing, reconstruction, and visualization of ultrasound NDE datasets, and 4) application of machine learning techniques for automated detection of ASR inside concrete. Our comparative study between FBD and traditional SAFT reconstruction images shows a clear difference between images of ASR and non-ASR specimens. In particular, the left first harmonic shows an increased contrast and sensitivity to ASR damage. For MBIR, we show the superiority of model-based techniques over delay and sum techniques such as SAFT. Improvements include elimination of artifacts caused by direct arrival signals, and increased contrast and Signal to Noise Ratio. For the universal framework, we document a format for data storage based on the HDF5 file format, and also propose a modular Graphic User Interface (GUI) for easy customization of data conversion, reconstruction, and visualization routines. Finally, two techniques for ASR automated detection are presented. The first technique is based on an analysis of the frequency content using Hilbert Transform Indicator (HTI) and the second technique employees Artificial Neural Network (ANN) techniques for training and classification of ultrasound data as ASR or non-ASR damaged classes. The ANN technique shows great potential with classification accuracy above 95%. These approaches are extensible to the detection of additional reinforced, thick concrete defects and damage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Multilayer Foil Manufacturability Final Report CRADA No. TSV-1246-96

The work performed under this CRADA was based on developments in multilayer technology made at Lawrence Livermore National Laboratory (LLNL) that enable advanced pyrotechnic material fabrication. We applied our understanding of the performance, composition, and microstructure of these materials to engineering design of application specific structures. The purpose of the project was to assess the feasibility of LLNL’s multilayer technology for application to large scale manufacturing of these advanced pyrotechnic materials for the automobile industry; address any manufacturing limitations; and propose approaches to improve manufacturability. Another purpose was to address issues associated with the design of application-specific materials for their subsequent manufacture, including the establishment of general guidelines for such a manufacturability assessment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS2024

Production and Characterization of Additively Manufactured Radiator Panels With Integral Branching Heat Pipes for High-Temperature Heat Rejection

Emerging concepts for fission surface power and nuclear electric propulsion necessitate lightweight, mechanically robust, and thermally efficient heat rejection radiators. State-of-the-art intermediate-temperature (~400 K) composite radiator assemblies have been developed based on titanium-water heat pipes bonded to metal, graphite, and carbon-fiber-based panels. NASA has identified a need for new radiator concepts that can operate at even higher temperatures (500 – 600 K), minimize thermal resistances and thermal stress failures at bond interfaces, and approach areal densities of 2 – 3 kg m -2 . To meet these needs, our team is developing additively manufactured (AM) radiator panels with integral branching wicking heat pipe networks. Water is selected as the working fluid for this temperature range. Based on simulations and thermal vacuum experiments, these branching embedded heat pipe networks can efficiently distribute heat over panels for finned surface efficiencies of η f >70% at T H = 500 K input heat. This paper first presents laser powder-bed fusion AM strategies to produce embedded porous structures for wicking heat pipes in Inconel 718 and titanium alloys (commercially pure and Ti-6Al-4V alloys). Post-build chemical and thermal treatments are described that yield hydrophilic wicking surfaces for operation with water. Transient rate-of-rise experiments with water and acetone are reported that yield estimates for AM wick porosity (ϵ), permeability (K), and effective pore radius ( r pore ). Based on the wick characterization results, small prototype radiator panels (75 × 125 mm) with integrated heat pipe networks were manufactured. Heat rejection performance data are presented from cold thermal vacuum testing, with heat input temperatures up to ~510 K. Future efforts will focus on improving heat pipe performance, optimizing radiator mass, and evaluating larger panels to assess scalability.

nuclear electric propulsion↗

DOE OSTI2026

Flaw Tolerance Assessment for DOE Standard SNF Dry Storage Canisters - 26550

The U.S. DOE has designed four spent nuclear fuel (SNF) dry storage canisters for storing DOE standardized SNFs. The DOE standard canisters are cylindrical shells with a diameter of 24 inches (610 m) or 18 inches (457 m), a wall thickness of 0.5 inches (12.7 m) or 0.375 inches (9.53 m), and a length of 15 feet (4.57 m) or 10 feet (3.05 m). These DOE canister geometries are completely different from commercial canisters. The latter may experience chloride-induced stress cracking corrosion (CI-SCC) because they are stored near coastal regions. The former may not experience CI-SCC but face different challenges because they are stored in the SNF storage facilities. Because of large residual stresses, mechanical flaws may occur in the DOE canisters during long-distance transportation or lifting handling. To date, only limited structural integrity analyses were carried out through drop tests on the DOE canisters, but a more general flaw tolerance assessment has not been performed. Therefore, the failure assessment diagram (FAD)-based fracture mechanics method, as codified by the latest API 579-1/ASME FFS-1-2021 Edition, is adopted in this work to assess surface flaw tolerance for DOE canisters under operation loading and welding residual stresses (WRS), where the new code-recommended WRS distributions are used. To more adequately consider the transverse distribution of WRS, an equivalent WRS distribution is proposed to account for the WRS reduction with distance from the weld centerline. Moreover, the closed-form solutions of stress intensity factor K, which serves as the crack driving force during subcritical crack growth, are developed from the tabular data of the K factors provided in API 579-1/ASME FFS-1 and used to determine more accurate flaw sizes at flaw instability. Subsequently, the Level 2 assessment procedures with 12 assessment steps, as codified and detailed in API 579-1 and ASME FFS-1, are followed to assess the flaw tolerance for the surface flaws in the DOE standard canisters with consideration of normal or accident operation loads combined with WRS. The assessment results show that the four designs of DOE standard canisters can tolerate all surface flaws that meet the code permitted maximum sizes of a flaw length of 8 inches (i.e., 200 mm) and a flaw depth of 80% wall thickness. This demonstrates that all designs of DOE standard canisters are robust and reliable.

DOE standard canister↗

DOE OSTI2020

Risk-Informed MultiPphysics BestEestimate Plus Uncertainties (BEPU) Application Development of RELAP5-3D Perturbation Model

The United States nuclear industry is facing a strong challenge to ensure maximum safety while enhancing economic benefit. Safety is a key parameter to all aspects related to light water reactor (LWR) nuclear power plants (NPPs), especially cost savings. Since the goal is to extend the lifetimes of these NPPs, the traditional deterministic safety concept may not guarantee a current economic asset. The Light Water Reactor Sustainability (LWRS) Program has been promoting a wide range of research and development (R&D) in this field to maximize the safety, economics, and performance of these NPPs through improved scientific understanding. One of the best practices to achieve this goal is to identify and optimize safety margins, which can lead to cost reduction. To do this, under the LWRS framework, the Risk-Informed Systems Analysis (RISA) Pathway will focus on the optimization of safety margin and minimization of uncertainties to ensure both safety and economics at the highest level. The RISA Pathway will provide enhanced capabilities for analyzing and characterizing LWR systems performance by developing and demonstrating methods, tools, and data to enable risk-informed margins management (RIMM). The goals of the RISA Pathway are twofold: (1) deploy the risk-informed tools and methods that enable better representation of safety margins and factors that contribute to cost and safety; and (2) conduct advanced risk assessment applications with industry to support margin management strategies that enable more cost-effective plant operation. The tools and methods provided by the RISA Pathway will support effective margin management for both active and passive safety systems, structures, and components (SSC) of an NPP. The tools and methods used in the RISA Pathway should have high confidence and highest technical maturity for and implementation to industry at its current setting. They should also have a capability to support risk-informed decision making for both probabilistic and deterministic elements of safety. The RISA Pathway will, therefore, perform a comprehensive assessment of verification and validation (V&V) status of RISA Toolkit to enhance credibility RISA Toolkit which be used by industry. This report summarizes RELAP5-3D development activities for best estimated and uncertainty (BEPU) capability to support ongoing RISA Pathway pilot projects for risk informed uncertainty quantification applications. Work scope includes (1) development of initial perturbation model for selected closure laws and (2) testing of suitable probability distribution functions (PDF) for the selected correlations. Developed BEPU modules are tested with selected separate effect cases and compared with conventional thermal-hydraulics model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Electromagnetic and Mechanical Analysis of the Coil Structure for the CLAS12 Torus for 12 GeV Upgrade

The torus magnet for the CLAS12 spectrometer is a 3.6 T superconducting magnet being designed and built as part of the Jefferson Lab 12 GeV Upgrade. The magnet consists of six coil case assemblies mounted to a cold central hub. The coil case assembly consists of an aluminum case and cover enclosing an epoxy vacuum impregnated coil pack. The coil pack consists of a 117 turn double-pancake winding wrapped with 2 layers of 0.635 mm thick copper cooling sheets. The coil case assembly is cooled by supercritical helium at 4.6 K. This report details the structural analysis of the coil case assembly and the assessment of the coil pack stresses. For the normal operation of the torus magnet, the coil case assembly was analyzed for cool down to 4.6 K and the Lorentz forces at normal operating current. In addition to the normal operating configuration, the coil case assembly was analyzed for Lorentz forces arising from coil misalignment and current imbalances. The allowable stress criteria for the magnet followed the approach of the ASME codes. Primary stresses were limited to the lesser of 2/3 times the yield strength or 1/3 times the ultimate tensile strength. Primary plus secondary stresses were limited to 3 times the primary stress allowable. The analysis was performed using ANSYS Maxwell to calculate the magneto-static loads and ANSYS Mechanical to calculate the stresses.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2014

Thermographic inspection of massive structures

Nondestructive Evaluation of concrete structures is a growing concern for the nuclear industry as well as for many other industries. As critical concrete components continue to age, the ability to assess the health and suitability for continued service has become a key consideration. In some cases, repair of these structures is difficult and expensive, while replacement is prohibitively expensive or, in some cases, not feasible. Therefore, the ability to inspect these key assets is a primary concern, especially in the nuclear industry. Due to the large size of containment buildings, cooling towers, and other large concrete assets, the ability to rapidly inspect for defects of concern is very desirable. Thermographic inspection appears to have the required ability to rapidly inspect large structures to ascertain the location and size of many of the defects of concern. This ability was demonstrated by performing a thermographic inspection of a large concrete dam in 2 days.

42 ENGINEERING↗

DOE OSTI2013

Preliminary Reactor Physics Assessment of the HTR Module with 14% Enriched UCO Fuel

The high temperature reactor (HTR) Module (Lohnert, 1990) is a graphite-moderated, helium cooled pebble bed design that has been extensively used as a reference template for the former South African (Matzner, 2004) and current Chinese (Zhang et al., 2009) HTR programs. This design utilizes spherical fuel elements packed into a dynamic pebble bed, consisting of tri-structural isotropic (TRISO) coated uranium oxide (UO2) 500 µm fuel kernels with a U-235 enrichment of 7.8% and a heavy metal loading of 7 g per pebble. This fuel type was previously qualified for use in Germany for pebble bed HTRs, as well as undergoing re-qualification in South Africa for the PBMR project. It is also the fuel type being tested for use in the high temperature reactor (HTR-PM) under construction in China. In the United States, however, a different TRISO fuel form is the subject of a qualification program. The U.S. experience with HTRs has been focused upon the batch-fueled prismatic reactor in which TRISO particles are embedded in cylindrical compacts and stacked inside the graphite blocks which comprise the core. Under this type of operating regime, a smaller TRISO with a different composition and enrichment performs better than the fuel historically used in PBRs. Fuel kernels and compacting techniques more suited to prismatic core duty are currently being developed and qualified under the U.S. Department of Energy's Advanced Gas Reactor (AGR) fuel development program and in support of the Next Generation Nuclear Plant project. Interest in the pebble bed concept remains high, however, and a study was undertaken by the authors to assess the viability of using AGR fuel in a pebble bed reactor. Using the German HTR Module as the reference plant, key neutronic and thermal-hydraulic parameters were compared between the nominal design and one fueled with the fuel that is the focus of the AGR program.

99 GENERAL AND MISCELLANEOUS↗

NASA NTRSDate not supplied

Production and Characterization of Additively Manufactured Radiator Panels with Integral Branching Heat Pipes for High-Temperature Heat Rejection

Emerging concepts for fission surface power and nuclear electric propulsion necessitate lightweight, mechanically robust, and thermally efficient heat rejection radiators. State-of-the-art intermediate-temperature (~400 K) composite radiator assemblies have been developed based on titanium-water heat pipes bonded to metal, graphite, and carbon-fiber-based panels. NASA has identified a need for new radiator concepts that can operate at even higher temperatures (500 – 600 K), minimize thermal resistances and thermal stress failures at bond interfaces, and approach areal densities of 2 – 3 kg m -2 . To meet these needs, our team is developing additively manufactured (AM) radiator panels with integral branching wicking heat pipe networks. Water is selected as the working fluid for this temperature range. Based on simulations and thermal vacuum experiments, these branching embedded heat pipe networks can efficiently distribute heat over panels for finned surface efficiencies of η f >70% at T H = 500 K input heat. This paper first presents laser powder-bed fusion AM strategies to produce embedded porous structures for wicking heat pipes in Inconel 718 and titanium alloys (commercially pure and Ti-6Al-4V alloys). Post-build chemical and thermal treatments are described that yield hydrophilic wicking surfaces for operation with water. Transient rate-of-rise experiments with water and acetone are reported that yield estimates for AM wick porosity (ϵ), permeability (K), and effective pore radius (r pore ). Based on the wick characterization results, small prototype radiator panels (75 × 125 mm) with integrated heat pipe networks were manufactured. Heat rejection performance data are presented from cold thermal vacuum testing, with heat input temperatures up to ~510 K. Future efforts will focus on improving heat pipe performance, optimizing radiator mass, and evaluating larger panels to assess scalability.

thermal management↗

NASA NTRS2008

Reliability Assessment of Graphite Specimens under Multiaxial Stresses

An investigation was conducted to predict the failure strength response of IG-100 nuclear grade graphite exposed to multiaxial stresses. As part of this effort, a review of failure criteria accounting for the stochastic strength response is provided. The experimental work was performed in the early 1990s at the Oak Ridge National Laboratory (ORNL) on hollow graphite tubes under the action of axial tensile loading and internal pressurization. As part of the investigation, finite-element analysis (FEA) was performed and compared with results of FEA from the original ORNL report. The new analysis generally compared well with the original analysis, although some discrepancies in the location of peak stresses was noted. The Ceramics Analysis and Reliability Evaluation of Structures Life prediction code (CARES/Life) was used with the FEA results to predict the quadrants I (tensile-tensile) and quadrant IV (compression-tension) strength response of the graphite tubes for the principle of independent action (PIA), the Weibull normal stress averaging (NSA), and the Batdorf multiaxial failure theories. The CARES/Life reliability analysis showed that all three failure theories gave similar results in quadrant I but that in quadrant IV, the PIA and Weibull normal stress-averaging theories were not conservative, whereas the Batdorf theory was able to correlate well with experimental results. The conclusion of the study was that the Batdorf theory should generally be used to predict the reliability response of graphite and brittle materials in multiaxial loading situations.

Sookdeo, Steven↗

DOE OSTI2015

Thermophysical properties of U 3 Si 2 to 1773K

Use of U 3 Si 2 in nuclear reactors requires accurate thermophysical property data to capture heat transfer within the core. Compilation of the limited previous research efforts focused on the most critical property, thermal conductivity, reveals extensive disagreement. Assessment of this data is challenged by the fact that the critical structural and chemical details of the material used to provide historic data is either absent or confirms the presence of significant impurity phases. This study was initiated to fabricate high purity U 3 Si 2 to quantify the coefficient of thermal expansion, heat capacity, thermal diffusivity, and thermal conductivity from room temperature to 1773 K. Here, the datasets provided in this manuscript will facilitate more detailed fuel performance modeling to assess both current and proposed reactor designs that incorporate U 3 Si 2 .

36 MATERIALS SCIENCE↗

NASA NTRS2017

Effect of Mechanical and Thermal Loading Histories on Residual Properties of SiCf/SiC Ceramic Matrix Composites

Silicon Carbide based Ceramic Matrix Composites (CMCs) are attractive materials for use in high-temperature structural applications in the aerospace and nuclear industries. Under high stresses and temperatures, creep degradation is the dominant damage mechanism in CMCs. Consequently, chemical vapor infiltration (CVI) SiCf/SiC ceramic matrix composites (CMC) incorporating SylramicTM-iBN SiC fibers coated with boron nitride (BN) interphase and CVI-SiC matrix were tested to examine creep behavior in air at a range of elevated temperatures of (2200 - 2700 F). Samples that survived creep tests were evaluated via RT fast fracture tensile tests to determine residual properties, with the use of acoustic emission (AE) to assess stress dependent damage initiation and progression. Microscopy of regions within the gage section of the tested specimens was performed. Observed material degradation mechanisms are discussed.

creep↗