Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “HTR”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Enhanced electrical penetration components meet the high safety standards required for China's HTR-PM Shidaowan and generation mPower small modular reactors - 300

Enhanced glass-to-metal feedthroughs will be used to carry key instrumentation and control (I and C) signals in China's high-temperature, gas-cooled pebble-bed modular reactor (HTR-PM) at Shidaowan, and in Small Modular Reactor (SMR) prototyping. Chinergy Co., Ltd and Jiamusi Electric Machine Co., Ltd will use the feedthroughs, also called electrical penetration assemblies (EPAs), at the Shidaowan twin-reactor high temperature reactor (HTR). These feedthroughs are manufactured using specialized processes that make them far more robust than counterparts made with organic or ceramic materials. Improved EPAs manufactured by SCHOTT are the only feedthrough solution that can be used in the primary loop of an HTR because organic seals cannot withstand the high temperatures and pressures of this environment. Glass-to-metal seals (GTMS) are made using inorganic, non-aging materials which enable a virtually unlimited lifespan in high-temperature, high-radiation and high-pressure applications. This I and C technology package is being enhanced by the development and testing of new mineral-insulated (MI) cable connectors and EPAs made of crystallized glass-ceramic combined with superalloys. The EPAs have a maintenance-free lifespan of over 60 years, preventing any outage time related to these components and subsequently reducing costs associated with shutdowns. They are far more robust than organic seals and provide increased safety at a lower lifetime cost. Resistance against high pressure and high temperature are especially ideal qualities that make EPAs a valuable choice for environments with high standards for fire protection, such as nuclear plants, subsea oil drilling, or any other underwater application. Due to their significant offering of safety-relevant qualities, the EPAs have also been chosen by BWXT for implementation in its mPower SMR. Small modular reactor requirements are complex and very challenging but are met by the new EPA components, which will help form a strong I and C safety chain. Unlike ordinary compression glass-to-metal seals, GTMS with crystallized glass-ceramic have the ability to withstand temperatures of over 350 deg. C / 662 deg. F and operating pressures of 3500 bar. Furthermore, their lifetime is qualified for up to 60 years, including pressure and radiation requirements in accordance with the latest IEC/IEEE standards. Higher-specification EPAs and MI cable connectors have been designed into severe accident I and C instruments, including advanced hydrogen and oxygen sensors that exceed the challenging requirements of both HTR and SMR applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering effects during pebble loading. Core 4 was determined to be acceptable benchmark experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Summary and experience feedback on the restart and power operation after a long term shutdown of HTR-10 - 18460

After a long term shutdown, 10 MW high temperature gas-cooled reactor (HTR-10) has been restarted in 2014. The main content of this process has been summarized and presented. The preparation in the reactor shutdown stage, the characteristic tests after the restart, and the engineering practice and experience during the power operation of HTR-10, have been introduced in details. The experience feedback on the restart process and power operation stage of HTR-10 have been proposed, including the equipment maintenance, technical transformation, routine test and inspection, modification of operational limits and conditions, and personnel training, etc. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

The HTR-Proteus Benchmark: Analysis and Use as a Verification and Validation Case

This presentation outlines the evaluation and application of the HTR-Proteus benchmark as a verification and validation (V&V) case for advanced reactor modeling tools. The work supports the U.S. Department of Energy’s HALEU Availability Program (HAP) and the joint DOE/NRC DNCSH project, which aims to reduce criticality safety uncertainties in commercial-scale HALEU fuel cycle and transportation systems. The HTR-Proteus experiments, conducted at the Paul Scherrer Institute, provide high-fidelity data for TRISO-fueled, graphite-moderated pebble bed reactors with high neutron leakage—conditions relevant to HALEU transport scenarios. This study focuses on Cores 4.2 and 4.3 of the HTR-Proteus benchmark, analyzing key sources of uncertainty including pebble packing, TRISO particle positioning, and core height. Using Project Chrono for realistic pebble geometries and Serpent, SHIFT, and MCNP for neutronics simulations, the study quantifies the impact of these uncertainties on the effective multiplication factor (keff). Results show that a sample size of 110 pebble configurations is sufficient to converge keff, with ±30 pcm uncertainty due to packing randomness. TRISO positioning and core height variations also significantly influence keff, highlighting the importance of detailed modeling in V&V efforts. The benchmark serves as a valuable test case for validating the Griffin reactor physics code and improving confidence in HALEU system simulations.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

HTR solver: An open-source exascale-oriented task-based multi-GPU high-order code for hypersonic aerothermodynamics

In this study, the open-source Hypersonics Task-based Research (HTR) solver for hypersonic aerothermodynamics is described. The physical formulation of the code includes thermochemical effects induced by high temperatures (vibrational excitation and chemical dissociation). The HTR solver uses high-order TENO-based spatial discretization on structured grids and efficient time integrators for stiff systems, is highly scalable in GPU-based supercomputers as a result of its implementation in the Regent/Legion stack, and is designed for direct numerical simulations of canonical hypersonic flows at high Reynolds numbers. Additionally, the performance of the HTR solver is tested with benchmark cases including inviscid vortex advection, low- and high-speed laminar boundary layers, inviscid one-dimensional compressible flows in shock tubes, supersonic turbulent channel flows, and hypersonic transitional boundary layers of both calorically perfect gases and dissociating air.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

Seismic fragility analysis for lateral supports of HTR-PM RPV - 18521

Since Fukushima accident, there has been increasing interest in China to study the risks of nuclear power plants under seismic events that are beyond the design basis. Two methodologies are commonly being adopted for this purpose, namely the seismic probabilistic risk assessment (SPRA) and the seismic margin assessment (SMA). Fragility analysis plays an important role as it not only provides fragility curves as input for the SPRA, but also can deduce HCLPF values for the SMA. In this paper, seismic fragilities of the lateral supports of the HTR-PM reactor pressure vessel (RPV) are studied. The lateral supports are crucial to the safety of HTR-PM RPV under seismic events. As they are located at a high elevation in the reactor building, it is expected they are one of the major contributors to the seismic risk of HTR-PM. The lateral supports are first briefly introduced in terms of structural features, load conditions and stress acceptance criteria in the engineering design. Their fragilities are then analyzed by identifying basic fragility analysis variables and determining values in terms of the median and standard deviations. The associated fragility curves are finally calculated and a HCLPF value is reported. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Benchmark Evaluation of the HTR-PROTEUS Absorber Rod Worths (Core 4)

PROTEUS was a zero-power research reactor at the Paul Scherrer Institute (PSI) in Switzerland. The critical assembly was constructed from a large graphite annulus surrounding a central cylindrical cavity. Various experimental programs were investigated in PROTEUS; during the years 1992 through 1996, it was configured as a pebble-bed reactor and designated HTR-PROTEUS. Various critical configurations were assembled with each accompanied by an assortment of reactor physics experiments including differential and integral absorber rod measurements, kinetics, reaction rate distributions, water ingress effects, and small sample reactivity effects [1]. Four benchmark reports were previously prepared and included in the March 2013 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [2] evaluating eleven critical configurations. A summary of that effort was previously provided [3] and an analysis of absorber rod worth measurements for Cores 9 and 10 have been performed prior to this analysis and included in PROTEUS-GCR-EXP-004 [4]. In the current benchmark effort, absorber rod worths measured for Core Configuration 4, which was the only core with a randomly-packed pebble loading, have been evaluated for inclusion as a revision to the HTR-PROTEUS benchmark report PROTEUS-GCR-EXP-002.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

HTR-1.3 solver: Predicting electrified combustion using the hypersonic task-based research solver

Here this manuscript presents an updated open-source version of the Hypersonics Task-based Research (HTR) solver. The solver, whose main features are presented in Di Renzo et al. (2020) and Di Renzo & Pirozzoli (2021), is designed for direct numerical simulation of reacting flows at high Reynolds numbers. This new version extends the applications of the HTR solver to turbulent combustion in the presence of external electric fields. In particular, a new distributed Poisson solver compatible with heterogeneous architectures has been incorporated in the algorithm to compute the electric potential distribution in bi-periodic configurations. The drift fluxes of the electrically charged species are now included in the transport equations using a targeted essentially non-oscillatory scheme. A verification of these new features of the solver is provided using one-dimensional burner stabilized flames, whereas a three dimensional turbulent flame is utilized to discuss the scalability of the proposed numerical tool.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

Comparison of Homogeneous and Heterogeneous CFD Fuel Models for Phase I of the IAEA CRP on HTR Uncertainties Benchmark

Computational Fluid Dynamics (CFD) evaluation of homogeneous and heterogeneous fuel models was performed as part of the Phase I calculations of the International Atomic Energy Agency (IAEA) Coordinate Research Program (CRP) on High Temperature Reactor (HTR) Uncertainties in Modeling (UAM). This study was focused on the nominal localized stand-alone fuel thermal response, as defined in Ex. I-3 and I-4 of the HTR UAM. The aim of the stand-alone thermal unit-cell simulation is to isolate the effect of material and boundary input uncertainties on a very simplified problem, before propagation of these uncertainties are performed in subsequent coupled neutronics/thermal fluids phases on the benchmark. In many of the previous studies for high temperature gas cooled reactors, the volume-averaged homogeneous mixture model of a single fuel compact has been applied. In the homogeneous model, the Tristructural Isotropic (TRISO) fuel particles in the fuel compact were not modeled directly and an effective thermal conductivity was employed for the thermo-physical properties of the fuel compact. On the contrary, in the heterogeneous model, the uranium carbide (UCO), inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers of the TRISO fuel particles are explicitly modeled. The fuel compact is modeled as a heterogeneous mixture of TRISO fuel kernels embedded in H-451 matrix graphite. In this study, a steady-state and transient CFD simulations were performed with both homogeneous and heterogeneous models to compare the thermal characteristics. The nominal values of the input parameters are used for this CFD analysis. In a future study, the effects of input uncertainties in the material properties and boundary parameters will be investigated and reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Effects of bypass flow and power change and deviation on performance of thermal mixing structure of HTR-PM - 18424

The temperature of bypass flow not flowing through the reactor core is obviously lower than that of the main coolant helium in a High Temperature Gas-cooled Reactor (HTGR). In addition, change and deviation of reactor power because of power regulation will introduce higher radial temperature difference of coolant helium from reactor to steam generator. In order to investigate the effects of bypass flow and power change and deviation on radial temperature difference of the helium flow entering the steam generator, model experiments and simulation calculations are proposed to evaluate the mixing performance of the thermal mixing structure at HTR-PM reactor core outlet. The experiments on Test Facility Hot Gas Mixing (TF-HGM) are carried out to observe the influences of bypass flow and reactor power deviation on mixing performance. In addition, some simulation calculations are conducted for the main mixing structure of the TF-HGM facility and the thermal mixing structure of HTR-PM. According to the results of experiments and simulation calculations, the maximum radial temperature difference at the outlet of the mixing structure fulfils the requirement by steam generator, less than 30 deg. C ({+-}15 deg. C), under all considered conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Study on Water-ingress Accident during the Initial Core and the Running-in Phase of HTR-PM - 18429

Water ingress accident is one of the typical accidents with potential hazard to the high temperature gas-cooled reactor (HTGR), which will cause reactivity introduction, as well as the chemical corrosion of the graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes. Therefore high attention has been paid to water ingress phenomenon during the design of the HTGR to verify its inherent safety characteristics. The online continuous discharge and multi-pass of the fuel elements is another important characteristics for the pebble-bed HTGR which can effectively increase the reactor operation availability. From the initial core, a long period of time, namely the running-in phase, is needed to reach the final long-term stable equilibrium core. During this period of time, the core states are totally different, including the number of the fuel elements, the fuel enrichment, the accumulation of the fission product, the temperature feedback coefficient of the reactivity, and so on. In this paper, based on the design of 250 MW Pebble-bed Module High Temperature gas-cooled Reactor (HTR-PM) and a candidate proposal for its initial core and running-in phase, considering three typical states during the running-in phase, the design basis accident of a double-ended guillotine break of one heating tube has been analyzed with the TINTE code, and the results are compared with that of the equilibrium core. Some important parameters during the transient process, including the power, the temperature of the fuel element, the pressure of the primary coolant, as well as the graphite corrosion rate, are studied. To a certain extent, the results can verity the design of the initial core and the running-in phase, and also further prove the safety feature of the HTR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Design of dehumidification test of HTR-PM - 18461

Moisture in the primary loop is harmful for the normal operation of the Modular High Temperature Gas-cooled Reactor (MHTGR). A dehumidification test to remove the moisture in the primary loop of the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) reactor was proposed. This test consisted of two stages, stage A and stage B. The test process of stage A was divided into six steps including vacuuming, inflating, heating, dehumidifying, cooling and venting, while the test process of stage B only included the first five steps of stage A. Nitrogen and helium were respectively used as the dehumidification gas to carry away the moisture in the graphite in stage A and stage B, which were circulated and heated by the helium blower. What s more, preliminary calculations to heating and dehumidifying were implemented. Variations of water content of the graphite in the dehumidification process were obtained, and the times spent for different dehumidification targets were also worked out. It needs 19.5 days to remove 80% of the moisture in the graphite in stage A when the dehumidification temperature is 150 deg. C, and 14.3 days are needed to make water content in the graphite decrease from 20% to 5% in stage B. The results show that the test method proposed is feasible and effective, and offer references to the design, test and operation of the HTR-PM. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Using CFD tools to model separate effects of HTR air ingress accident phenomena - 18508

The air ingress phenomena under the HTR conditions represent the important area of nuclear safety. Best practice guidelines for the use of CFD in nuclear safety applications are mainly based on water cooled reactors while in HTR the gas is used as a coolant. The ambition of this paper is to define the best recommendations and description of needed settings of CFD models solving the separate effects of air ingress accident phenomena. The JAERI experiments deal with phenomena of air ingress during the early stages of the accident. Experiments investigate the basic features of the flow behavior of the gas mixture and regard the phenomena of the molecular diffusion and the natural convection of the gas mixture with the graphite oxidation reaction. The separate effects simulations of air ingress accident were performed using CFD code ANSYS Fluent. Whole graphite oxidation process including all three main chemical reactions was successfully modeled. The new findings and useful information were gathered to be extrapolated to the integral effects level for the following phase. The role of the computational mesh and all important input parameters were discussed in detail and many in sensitivity studies were performed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The development of 3D physical program with integrated control rod calculation for HTR - 18525

The accurate calculation of the control rod worth in pebble bed high temperature gas cooled reactor(HTR) requires special treatment, because the control rod system of HTR is located outside the active zone. The current version VSOP did not calculate the control rod value directly, but use a compromise method called the Equivalent Boron Concentration (EBC). To obtain the value of the control rod with higher precision directly, this paper embedded the discontinuity factor (DF) corrected diffusion calculation into the original procedure of the VSOP. The computational procedure to obtain the DF on the side of the absorber region was optimized. While an appropriate coupling method has been developed to take into account the neutron leakage feedback, data transform and so on. The new program can achieve accurate control rod calculation and more realistic 3D power distribution during the physical calculation. The numerical results demonstrated the correctness of the new program preliminarily. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

fuel cycle cost analysis of Th/U MOX fuel in a block-type HTR - 18619

With a rapid development of nuclear energy in these years, the supply of nuclear fuel is becoming more and more important. Thorium regains attention because of its abundant reserves and excellent physical properties. As a next generation reactor, the high temperature reactor (HTR) is a promising choice for thorium utilization because of its deeper burnup, harder spectrum and better inherent safety. Therefore, a 600 MWth 108-column modular high temperature reactor (MHR) core model with 20%UO{sub 2}-80%ThO{sub 2}; MOX fuel is adapted. A traditional two-step calculation scheme, lattice calculation then core calculation, is used to get the neutronic performance of equilibrium cycle including uranium enrichment and mass of fuel. Using these data, the fuel cycle cost of HTR can be calculated by the levelized lifetime cost method. In order to take into account the variation of unit price, a well-established, standard statistical analysis technique is used by Monte Carlo simulation, (thus, the probability distribution of fuel cycle cost can be obtained. Furthermore, sensitivity calculations are made to analyze the impact on the fuel cycle cost of variations in the unit prices for each fuel cycle component. The result shows that the most influential parameters are unit price of natural uranium, enrichment and discount rate. Finally a comparison of fuel cycle cost between 100%UO{sub 2} and 20%UO{sub 2}-80%ThO{sub 2} MOX fuel is made. The latter is 11% cheaper than the former. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preliminary study on applying the method of characteristics to pebble-bed HTR - 18637

Neutron transport in high temperature pebble-bed gas-cooled reactor, which is a promising generation IV reactor candidate with cylindrical core, graphite moderator and spherical TRISO coated particles fuel, is of great interest. However, the stochastic method namely Monte Carlo faces with tremendous simulation of particles and the conventional two-step method featured by homogenization only can give the scalar flux. In this paper, method of characteristics is trying to be applied to pebble bed reactor because of its advantages in arbitrary geometry and the ability to obtain solution in neutronically large domains. Some challenges occur when it comes to HTR. First, the amount of calculation and the memory requirement are huge due to the large quantities of spherical fuel and stochastic distribution of pebbles. Second, the arrangement of characteristics and ray tracing process is more complex comparing to LWR with cuboid geometry because the stochastic distribution of the spherical assemblies and the longer neutron mean free path in pebble bed HTR. Third, the treatment of the multiscale sphere fuel need be detail studied to determine the subdivision of flat source region. At present, preliminary research on this subject is undergoing. This paper is concentrated on the modeling of the pebble-bed reactor core by constructive solid geometry (CSG) which allows complex spatial models to be built using boolean operations, characteristics arrangement and the treatment of boundary condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Helium production prediction in HTR fuel using Monte Carlo MCB system - 18713

The integrated burnup calculation code MCB deals with the complexity of burnup process, which occurs in HTR and requires application of an integrated calculation system with consideration of spatial effects of full heterogeneous reactor model with continuous energy representation of cross section and the thermo-hydraulic coupling. MCB automatically calculates helium production in all possible reaction or decay channels. The methodology of He calculation in HTR using MCB is described. It includes transmutation trajectories analysis in few batch reloading scenarios, which allows us for integrated analysis of helium production sources. The result of performed analysis is presented showing the sources of strong contribution to the helium production rate. Statistical fluctuation effects on Monte Carlo modeling of nuclear reactors as well as the uncertainty assessment are carried out, which includes the error propagation in the depletion calculations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Multi-Modular Coordination Control of HTR-PM600 Plant - 190

HTR-PM600 is a high temperature gas-cooled reactor plant based on the multi-modular scheme, i.e. six modular high temperature gas-cooled reactor (MHTGR) driving a common secondary loop system. Six MHTGR-based nuclear steam supplying system (NSSS) modules are tightly coupled by the secondary fluid flow network (FFN) which contains the secondary side of steam generators, feedwater system, main steam system, turbine generator system. The secondary FFN control, which focuses on the feedwater flowrates and main steam pressure, plays a central role in multi-modular coordination of HTR-PM600 plant. In this paper, a differential algebraic model of this secondary FFN is first proposed. Then, the corresponding flowrate-pressure control strategy is proposed, which takes the simple form of proportional-integral feedback laws. Numerical simulation results show both the feasibility and satisfactory transient performance of this controller. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗