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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
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 OSTI2013

Decommissioning for the containment vessel of HTR-PM: Monte Carlo simulation and long-time predictions

A high temperature gas cooled reactor-pebble bed module (HTR-PM) is currently under design and construction in China. Calculations of induced radioactivity in the concrete containment vessel of HTR-PM were performed in order to predict the decommissioning problems. The main activation products and the ambient dose equivalent rate corresponding to 40 years of operations followed by 30 days/1 year/10 years of decay, were calculated by using the Monte Carlo particle transport code FLUKA. Finally, according to the Chinese National Standard and on the basis of the knowledge gained from this study, the classifications of the radioactive solid wastes were discussed, and some suggestions for the working duration of the employees were given. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2012

The improvement of the method of equivalent cross section in HTR

The Method of Equivalence Cross-Sections (MECS) is a combined transport-diffusion method. By appropriately adjusting the diffusion coefficient of homogenized absorber region, the diffusion theory could yield satisfactory results for the full core model with strong neutron absorber material, for example the control rod in High temperature gas cooled reactor (HTR). Original implementation of MECS based on 1-D cell transport model has some limitation on accuracy and applicability, a new implementation of MECS based on 2-D transport model are proposed and tested in this paper. This improvement can extend the MECS to the calculation of twin small absorber ball system which have a non-circular boring in graphite reflector and different radial position. A least-square algorithm for the calculation of equivalent diffusion coefficient is adopted, and special treatment for diffusion coefficient for higher energy group is proposed in the case that absorber is absent. Numerical results to adopt MECS into control rod calculation in HTR are encouraging. However, there are some problems left. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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↗
DOE OSTI2023

Options, Initial Design Requirements, Estimated Costs, Reactor Commitments, and Potential Uses of a Graphite Leadout Type Experiment Supporting Various Commercial HTR Vendors

Multiple commercial High Temperature Reactor (HTR) vendors and nuclear graphite suppliers would benefit by collaborating on a new irradiation capsule(s) that would include graphite grades not included within the AGC Experiment. This new irradiation capsule(s) would provide data to answer vendor graphite licensing issues. Rather than spending money (and especially) time in designing separate irradiation capsules for each designer, the capsule(s) would be used for multiple graphite and composite designs to maximize efficiency and promote multiple HTR designs. However, the primary motivation for assisting vendors with this new irradiation capsule(s) is lack of availability in the existing Material Test Reactors (MTRs). Cost reduction is a secondary goal. A common, collaborative, capsule design can be achieved for graphite and composites due to the similarity of different grades. Irradiation, disassembly, shipping, and PIE costs would be cost-shared by all users. It is anticipated that interest would extend across all DOE campaigns (micro-Rx, SMR, GCR, MSR, etc.) due to the similar requirements for all graphite grades.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Temperature modeling for analysis and design of the sintering furnance in HTR fuel type of ball

One of the factors that determine the safety of the operation of the sintering furnace fuel HTR ball is the temperature distribution in the ceramic tube furnace. The temperature distribution must be determined at design stage. The tube has a temperature of 1600 °C at one end and about 40 °C at the other end. The outside of the tube was cooled by air through natural convection. The tube is a furnace ceramic tube which its geometry are 0.08, 0.09 and 0.5 m correspondingly for the inner tube diameter, outer tube diameter and tube length. The temperature distribution of the tube is determined by the natural convection coefficient (NCF), which is difficult to be calculated manually. The determination of NCF includes the Grasshoff, Prandtl, and Nusselt numbers which is a function of the temperature difference between the surrounding air with the ceramic tube. If the temperature vary along the tube, the complexity of the calculations increases. Thus the proposed modeling was performed to determine the temperature distribution along the tube and heat transfer coefficient using a self-developed software which permit the design process easier.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Multiphysics Analyses of the Protected and Unprotected Loss of Forced Cooling Accidents in the HTR-PM

Here, we present the multiphysics simulation results for the protected pressurized and depressurized loss of forced cooling (PLOFC and DLOFC) events in the High-Temperature gas-cooled Reactor--Pebble-bed Module (HTR-PM) equilibrium core using the Griffin-Pronghorn coupled code system. Additionally, this paper discusses the strategy for estimating the spontaneous fission neutron source needed for unprotected events and re-criticality calculations. The solutions of the protected PLOFC and DLOFC events were verified against similar solutions obtained for temperature evolutions from the open literature. Both the average and maximum pebble surface temperatures behaved as expected during the DLOFC and remained below 1800 K. The PLOFC results are highly dependent on the ability to resolve the natural circulation in the core, which is impacted by the mesh resolution in Pronghorn. Furthermore, we present the results of the unprotected DLOFC transient to predict the timing of the re-criticality event, which occurred 47 hours after the onset of the transient, and the new steady-state power of 1.3 MW.

42 - ENGINEERING↗
DOE OSTI2022

Verification of the AGREE and Serpent for the steady state HTR-10 benchmark problems

This paper provides results from steady state HTR-10 benchmark calculations performed using AGREE and Serpent that are compared to experimental results as well as calculations performed by INET. The purpose of completing this benchmark is to validate AGREE and Serpent for the prediction of HTGR operation so that they may ultimately be used to support licensing and deployment efforts of advanced reactors. The benchmark consists of several problems ranging from control rod worth calculations to k{sub eff} calculations at various temperatures. Overall, AGREE and Serpent show good agreement with the reference solutions and are effectively able to predict HTGR operation for a variety of steady state cases. The largest difference from the reference was for the initial core single rod worth, which is possibly due to the larger core helium cavity causing inaccuracies in the diffusion calculation whereas the control rod worth determined by Serpent and AGREE is much closer to the reference result for the full core loading. Moreover, in every benchmark problem, increasing the number of energy groups in the cross sections results in improved agreement of the AGREE result with the reference, although it also causes an increase in computation time. For problems with experimental data available, the accuracy of results generated by Serpent and AGREE is comparable to results obtained by other benchmark participants. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗