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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Temperature sensitivity of the equilibrium neutronics and accident analysis of the HTR-10

Pebble-Bed High-Temperature Gas-cooled Reactors (PB-HTGR) are moderated by the graphite in the fuel pebbles and the graphite reflector surrounding the pebble-bed. Because graphite is by far the most abundant material in PB-HTGRs and the primary moderator, accurate modeling of the graphite material, including density, impurities, and temperatures, is crucial for accurate computational modeling and simulation of these reactors. While main characteristics of the graphite components are often known, the local temperature is less well known and often averaged over all components. Here, this work studies the impact of considering accurate temperature profiles in the graphite material on the generation of a small PB-HTGR model at the state of equilibrium operation and on short-term accident progression. The fuel compositions for the PB-HTGR were determined using a jump-in equilibrium modeling method, the Axial Radial Zone Equilibrium Modeling (AR-ZEM) method. In contrast to previous work, the AR-ZEM method was used considering thermal-hydraulic feedback from the MELCOR code to determine temperatures of the fuel pebbles and the surrounding graphite reflector. The consideration of an axial and radial temperature profile in the core and reflector, as opposed to uniform material temperatures, had an impact of almost 1,300 pcm on the equilibrium core eigenvalue and caused significant differences in the discharged plutonium fuel inventory with up to 4.9% and 11.0% for Pu-239 and Pu-242, respectively. To assess the impact on short-term accident progression, two Anticipated Transient Without SCRAM (ATWS) events, a Pressurized Loss of Forced Coolant (PLOFC) and a Control Rod Withdrawal (CRW) with loss of flow, were simulated with MELCOR. The use of temperature profiles in the equilibrium core models did not reveal a significant impact on the temperature, power, or reactivity responses during the transients. In conclusion, a need for consideration of accurate temperature profiles, in particular for the graphite reflector, was found for the generation of equilibrium PB-HTGRs core models using jump-in methods, but detailed temperature profiles may not be necessary when performing conservative transient analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Analysis of the ASME Code Rules for Subsection III-5-HHB (Composite Materials) for Current HTR Design Requirements

This document includes the critical analysis review of the American Society of Mechanical Engineers (ASME) Section III Division 5 Subsection HH Subpart B (HHB), including Mandatory Appendices, that was published in 2023. In the context of this document, reference to “the code” is specific to this subsection unless otherwise specified. A specific composites task group within the ASME Nonmetallic Design and Materials Working Group, with the support of external experts, was established to perform a gap analysis review. The significant findings are summarized here. The committee response with suggested action items are detailed in the body of the report.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Comparison between Flownex and Corona thermal-hydraulic models for a single HTR prismatic fuel assembly - 18544

Computational tools and methodologies that can perform thermal-hydraulic analyses to reasonably predict the temperature distribution for the whole core of HTRs are useful for operational and safety analyses. In this paper, Flownex (a system code based on a one-dimensional thermal-fluid network methodology) and CORONA (Core Reliable Optimization and thermo-fluid Network Analysis) thermal-hydraulic models and results for a vertically stacked single fuel assembly will be compared. In Flownex, a basic unit cell containing part of the graphite and the associated fuel elements and coolant channels was developed. CORONA uses a basic unit cell concept for generating computational grids using several types of basic unit cells to model the heat transfer through the hexagonal blocks and a 1-D formulation to model the flow through the coolant channels. In CORONA, computational grids are generated using a combination of pre-generated mesh cells. The coolant channels and fuel rods are modelled explicitly. Flownex in contrast, models fuel and coolant channels implicitly by defining representative coolant channels, fuel rods and coated particle layers associated with each unit cell. Flownex can therefore model temperatures at coated particle level and also through each graphite block. The computational mesh employed by Flownex is much coarser than the mesh employed by CORONA. In both models, the helium gap between the fuel compact and the graphite block is modelled as stagnant. Flownex results were found to be of the same order of magnitude as those obtained using CORONA. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

AGC-4 Graphite Preirradiation Data Analysis Report

The Advanced Reactor Technology (ART) Graphite R&D program is conducting an extensive graphite irradiation program to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs. Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of new HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance is required for the physical, mechanical, and thermal properties of each major graphite grade with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change—where the graphite physically changes (i.e., first shrinking and then expanding with increasing neutron dose). This disparity in material volume change can produce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphite is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment. The AGC experiment is currently underway to determine the in service behavior of these new graphite grades for HTR and molten salt reactor designs. This experiment will examine properties and behavior of nuclear grade graphite over a large spectrum of temperatures, irradiation fluencies and applied stress levels that are expected to cause irradiation creep strains within a HTR graphite component. Irradiation data are provided through the AGC test series, which is comprised of six planned capsules irradiated in the Advanced Test Reactor (ATR) in a large flux trap located at Idaho National Laboratory (INL). Each irradiation capsule consists of over 400 graphite specimens that are characterized before and after irradiation to determine the irradiation induced material properties changes and life-limiting irradiation creep rate for each graphite grade.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Analysis and Optimization of the Recompression Cycle with High Temperature Recuperator Bypass for Concentrating Solar Power Applications

This work analyzes the sCO2 recompression with high temperature recuperator (HTR) bypass power cycle for use in concentrating solar power (CSP) systems. CSP operation differs from other thermal power plants in that CSP must balance between maximizing the heat transfer fluid (HTF) temperature difference and maximizing the cycle thermal efficiency, which typically are inversely related. Large HTF temperature differences reduce the size and cost of thermal energy storage (TES), improve the solar receiver efficiency, and require lower mass flow rates that reduce pumping power required to elevate the HTF to the receiver. The recompression cycle with HTR bypass potentially offers improved thermal efficiency with larger HTF temperature differences as compared to the recompression cycle, and it has fewer turbomachinery components than the partial cooling cycle. The recompression with HTR bypass cycle adds a second lower temperature primary heat exchanger which transfers heat from the HTF to the fraction of sCO2 flow that bypasses the HTR. We developed a model to compare the recompression with HTR bypass cycle to the recompression and partial cooling cycles. A sweep of design parameters including bypass fraction, recompression fraction, recuperator conductance, and pressure ratio is used to form a pareto-optimal front with the cycle thermal efficiency and HTF temperature difference as objectives. An optimization routine has also been developed to find optimal design point parameters for a target HTF temperature difference. The performance of the recompression with HTR bypass cycle is compared with recompression and partial cooling cycles.

concentrating solar power↗
DOE OSTI2012

Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

AGC 2 Irradiated Material Properties Analysis

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

36 MATERIALS SCIENCE↗
DOE OSTI2016

AGC 2 Irradiation Creep Strain Data Analysis

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

AGC-3 Irradiation Creep Strain Data Analysis

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete irradiated properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

42 ENGINEERING↗
DOE OSTI2016

The hydrogen behavior analysis in the water-ingress accident of the 250 MW pebble-bed modular high temperature gas-cooled reactor - 18639

The demonstration power plant of Chinese 250 MW high temperature gas-cooled reactor pebbled-bed module (HTR-PM) which is being under construction in the Shidao Bay of Shandong province has made breakthrough in all of the critical technologies, and is willing to operate in 2017. HTR-PM does not have the risk of large amount of hydrogen generation resulted from the zirconium water reaction due to its special design of graphite core structure and helium coolant of the primary circuit, however, considering that the HTR-PM has a small containment and does not design the deliberate ignitions which is the main hydrogen management measure in light water reactors, it is still necessary to study the hydrogen behavior in the water-ingress accident in which hydrogen will probably be generated. In this paper, the hydrogen generation amount and its distribution in the HTR-PM containment are analyzed with the TINTE code and GASFLOW code respectively in the water ingress accident caused by the steam generator heating tubes rupture. The calculation results indicate that there are total about 16.6 kg and 21 kg hydrogen produced due to graphite corrosion in the typical design basis water ingress accident and beyond design basis accident. The containment sub-pressure ventilation system makes sure the hydrogen concentration in the containment is less than 0.2% during the accident. There is no hydrogen explosion risk in the HTR-PM. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Mono-2-ethylhexyl phthalate inhibits human extravillous trophoblast invasion via the PPARγ pathway

Concerns over the adverse reproductive outcomes in human have been raised, more evidence including the underlying mechanism are required. Since extravillous trophoblast (EVT) invasion is an important physiological step during early development, the effects of mono-2-ethylhexyl phthalate (MEHP), the bioactive metabolite of DEHP, on EVT invasion were investigated using Matrigel-coated transwell chambers and cell line HTR-8/SVneo. In the transwell-based invasive assay, MEHP exposure inhibited EVT invasion as judged by decreased invasion index. Further analysis showed that MEHP exposure significantly inhibited the activity of matrix metalloproteinase-9 (MMP-9), which is an important positive regulator of EVT invasion. Meanwhile, the protein levels of tissue inhibitor matrix metalloproteinase-1 (TIMP-1), one key negative regulator of EVT invasion, were upregulated by MEHP treatment. Finally, inactivation of PPARγ pathway by either PPARγ inhibitors or PPARγ shRNA knockdown rescued the MEHP-induced inhibited invasion of HTR-8/SVneo cells, which is accompanied by the recovery of inhibited MMP-9 expression. The present study provides the evidence that MEHP exposure inhibits trophoblast invasion via PPARγ at concentrations comparable to those found in humans, which provides an insight in understanding the mechanisms of DEHP-associated early pregnancy loss. - Highlights: • MEHP inhibits HTR-8/SVneo cell invasion. • MEHP exposure imbalanced the expression of MMP-9/TIMP-1 in HTR-8/SVneo cells. • This effect of MEHP is mediated via the PPARγ pathway.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

Long non-coding RNA DLX6-AS1 is upregulated in preeclampsia and modulates migration and invasion of trophoblasts through the miR-376c/GADD45A axis

Highlights: • lncRNA DLX6-AS1 is upregulated in placentas from women with preeclampsia. • DLX6-AS1 inhibits trophoblastic migration and invasion. • DLX6-AS1 affects trophoblastic migration and invasion via miR-376c/GADD45A pathway. Long non-coding RNA (lncRNA) are key regulatory molecules that are implicated in diverse biological processes and human diseases, including preeclampsia. However, their expression and functions in the progression of preeclampsia remains largely unclear. In this study, lncRNA DLX6-AS1 was confirmed to be significantly upregulated in the placentas of patients with preeclampsia, compared with normal controls. Overexpression of DLX6-AS1 dramatically decreased proliferation, migration and invasion in trophoblast JEG3 and HTR-8/SVneo cells. Luciferase activity and RNA pull-down assays showed that DLX6-AS1 interacted with miR-376c, and that overexpression of DLX6-AS1 significantly reduced expression of miR-376c in HTR-8/SVneo cells. Also, miR-376c targeted and downregulated GADD45A in HTR-8/SVneo cells. Overexpression of GADD45A effectively attenuated a miR-376c-induced increase in the proliferation, migration and invasion of HTR-8/SVneo cells. Further research showed that DLX6-AS1-induced inhibition of trophoblast proliferation, migration and invasion was effectively neutralized by GADD45A knockdown and miR-376c overexpression. Taken together, these findings suggest that DLX6-AS1 may contribute to preeclampsia by impairing proliferative, migratory and invasive abilities of trophoblasts via the miR-376c/GADD45A axis.

60 APPLIED LIFE SCIENCES↗