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At least 1,279 records · Page 71
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

use of cross-section covariance information in uncertainty analysis at modeling HTGR neutron physics - 18551

This work is aimed at improvement of high temperature gas cooled reactors (HTGR) neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group binary library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for HTGR was generated. This library can be used for estimation of uncertainties associated with nuclear data in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k{sub ∞}) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code (technology SUW) and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. The eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: {sup 10}B(capt), {sup 238}U(n,γ), ν{sub 5}, {sup 235}U(n,γ), {sup 238}U(el), {sup nat}C(el), {sup 235}U(fiss)-{sup 235}U(n,γ) and {sup 235}U(fiss). It is shown that the main contribution in the uncertainty of multiplication factor of the fuel block model gives the reaction {sup 10}B(capt) ((n,α)+(n,γ)). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Lessons learned from the commissioning of the THTR 300 - 18575

Although the commissioning of the THTR 300, the only large pebble bed High Temperature Gas-Cooled Reactor (HTGR) power plant, occurred more than 30 years ago, many lessons could be learned that would directly benefit the design, construction and startup of future pebble bed HTGRs. This report includes a review of the THTR 300 design and an overview of the commissioning program utilized to bring the plant from initial criticality to full power. Selected plant test results are compared to predicted plant behavior from 40% to 100% plant load. Lessons learned from the commissioning of the THTR are also included. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

FM-BEM simulation of 3-D microvoid and microcrack graphite models with Python and ABAQUS/CAE pre-processing - 18603

The mechanical property and structural integrity of graphite structure have an important significance on the safe operation of high temperature gas cooled reactor. The graphite is porous media, the structural deficiencies such as microcracks and microvoids in it have great influence on its material mesoscopic structure and mechanical parameter. The shape and size distribution of microvoids and microcracks at different oxidation degree have a great influence on the overall performance. In order to simulate the geometric structure of the oxidized graphite, we used Python script to pre-process in Abaqus/CAE and this leads to different graphite models with different oxidation degrees. Then, these models were set with our pre-designed force boundary conditions and displacement boundary conditions. The models information was output to a FMBEM solver written by C++ code and the graphite s equivalent young s modulus and other mechanic parameters were calculated. In this paper, based on the 3- D microvoid and microcrack graphite models simulation, we studied the influence of the change of mesoscopic structure on graphite's mechanical property. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The IAEA coordinated research project on modular HTGR safety design: status and outlook - 18608

The IAEA Coordinated Research Project (CRP) on Modular High Temperature Gas-cooled Reactor (HTGR) Safety Design started in December 2014 and is now in its second year of implementation. The project investigates and makes proposals on safety design criteria by not only considering the existing water cooled reactors safety requirements for nuclear power plant design but also making use of past and current experience of HTGR licensing in the IAEA member states. Two approaches are being followed to establish the criteria. The first is to develop and apply a safety design process to specific reactor de-signs. The second is to use the existing IAEA specific safety requirements for designs based on water cooled reactor technology and then through expert judgment and review propose changes that is needed to make it applicable to HTGRs. The detailed project content and a description of modular HTGRs and its safety features are given. This informs the comments made on the applicability of the existing requirements and the examples of newly pro-posed HTGR specific safety requirements under development. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

IS Process Hydrogen Production Test for Components and System Made of Industrial Structural Material - (I) - Bunsen and HI Concentration Section - 18628

Japan Atomic Energy Agency (JAEA) has been conducting research and development on a hydrogen production system as one of the heat applications of High Temperature Gas-Cooled Reactor. The iodine-sulfur (IS) process is one of the most promising thermochemical methods because of its potential for large-scale CO{sub 2} free hydrogen production. In the IS process, hydrogen is produced through three chemical reactions: Bunsen reaction, hydriodic acid (HI) decomposition and sulfuric acid (H{sub 2}SO{sub 4}) decomposition. High corrosion resistance is required of the component materials in the IS process because the high-temperature strong acids flow in this process. JAEA has fabricated a 30 NL/h - H{sub 2}-scale apparatus made of glass and Teflon which boast corrosion-resistance for the acids, and has succeeded a continuous operation for a week. However, for practical use on a larger scale, the components have to be made of industrial corrosion-resistance materials. As a next step, we have fabricated the new test-facility which can produce H{sub 2} of 100 NL/h-scale for the demonstration of the continuous operation by using the industrial materials. In advance of the continuous operation, we investigated the performance of the components in each section of the IS process. In this presentation, the test results of Bunsen and HI concentration sections will mainly be discussed. (authors)

08 HYDROGEN↗
DOE OSTI2016

IS Process Hydrogen Production Test for Components and System Made of Industrial Structural Material (II) - H{sub 2}SO{sub 4} decomposition, HI Distillation, and HI Decomposition Section - 18629

Japan Atomic Energy Agency (JAEA) has been conducting research and development on the thermochemical iodine sulfur (IS) process for nuclear-powered hydrogen production. The IS process is one of the promising candidates of heat application of High-Temperature Gas-Cooled Reactors. The IS process consists of three chemical reactions: Bunsen reaction, hydriodic acid (HI) decomposition and sulfuric acid (H{sub 2}SO{sub 4}) decomposition. It is necessary for component materials to have high corrosion resistance because high-temperature strong acids such as HI and H{sub 2}SO{sub 4} are used in the IS process. JAEA, so far, achieved to produce hydrogen for one week continuously with a hydrogen production rate of approximately 30 NL/h by a test apparatus made of glass and fluoro-resin material. As a next step toward practical application, we test-fabricated main components, that is, Bunsen reactor, sulfuric acid decomposer, and HI decomposer, made of industrial materials and investigated their integrity in the practical corrosive environments of the IS process. Based on their results, we have constructed a 100 NL/h-H{sub 2}-scale test apparatus made of industrial materials. This test apparatus is divided by chemical reaction and unit operation into five sections: Bunsen section, H{sub 2}SO{sub 4} decomposition section, HI concentration section, HI distillation section, and HI decomposition section. First, we investigated performance of components in each section. This report discusses test results of the H{sub 2}SO{sub 4} decomposition section, HI distillation section, and HI decomposition section. (authors)

08 HYDROGEN↗
DOE OSTI2016

Research on reconstruction method for pebble-detecting CT system for PB-HTGR - 186633

In the pebble-bed high temperature gas-cooled reactor (PB-HTGR), pebble dynamics in the reactor is important for safe operation. However, the experimental research at present were mostly implemented in two-dimensional models or limited to phenomenological methods. To investigate pebble dynamics using a three-dimensional experimental method, we designed a platform with a pebble-detecting CT system exactly as the proportion of 1:5 with the actual HTR-PM. Multi-slice helical CT can provide cross-sectional images of fuel pebbles at any place during motion, allowing for pebble trajectories to be obtained during the whole unloading process. The purpose of this paper is to design appropriate algorithms to reconstruct cross-sectional images of pebbles for the system. Two methods applicable to distinct situations were discussed and compared. Based on this, a fast reconstruction method was introduced and proved to be the most suitable solution for the system to get pebble trajectories rapidly. The pebble-detecting principle and the effectiveness of the fast reconstruction method were demonstrated by a verification experiment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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

Proper orthogonal decomposition analysis of parallel twin jets measured using particle image velocimetry - 18685

The mixing of turbulent jets has been found in several locations of High Temperature Gas-cooled Reactors (HTGRs) including upper and lower plenums, where jets impinge into open space from coolant channels. In the present work, a parallel twin jets experimental facility was used to generate velocity field of mixing jets by means of Particle Image Velocimetry (PIV), which can be eventually used for validation of Computational Fluid Dynamics (CFD) simulations. The facility has two fixed parallel rectangular jet nozzles (5.8 mm in width, 87.8 mm in length) submerged in the middle of the water tank, which allows an assumption of free jet mixing. In order to compare the experimental data produced to existing data, Re = 4000 {approx} 5000 at the jet outlet was selected. In the past, several statistically meaningful data have been reported using Laser Doppler Velocimetry (LDV) and Hot Wire Anemometer (HWA), which can support Reynolds Averaged Navier-Stokes turbulent models. However, recent CFD simulations using Large Eddy Simulation (LES) require high resolution experimental data in time and space. Thus, in this study, in addition to statistically enough number of data sets for RANS models validation, high temporal and spatial resolution data were collected for the validation of LES calculations. In order to study the turbulent structure Reynolds decomposition and Proper Orthogonal Decomposition (POD) were applied to the data obtained. These decomposition analyses visualized multi-scale vortices and the transportation of eddies. In terms of twin jets structure, the present experimental data showed good agreement with other techniques for the important characteristic parameters including the length of the converging region and the location of the merging point. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Investigation of nuclear graphite emissivity behaviors at various temperatures and oxidation degrees - 18763

Air Ingress into the primary loop is generally considered as one of the most likely severe accidents that could occur in High Temperature Gas-cooled Reactors (HTGRs). With air ingress, the graphite materials would be oxidized resulting in decreased graphite structure strength and release of harmful gases. As the main passive heat transfer mechanism during air ingress, heat transfer by radiation has high efficiency due to high temperature fuels in HTGRs. However, it is found that oxidation of fuel cladding affects the radiation heat transfer of graphite components. The purpose of the current study is to propose an empirical model that predicts the normal spectral emissivity ({lambda}=1.55 {mu}m) of the IG-110 at various temperatures and oxidation degrees. In this study, the emissivities of IG-110 at 0.0 burn-off, 4.44% burn-off and 8.90% burn-off were measured by employing IR thermometer and K-type thermocouples at four different temperatures corresponding to 600 deg. C, 700 deg. C, 800 deg. C and 900 deg. C. An empirical model of the spectral emissivity as a function of temperature and oxidation degree was proposed and validated against experimental data. In addition, the emissivity measurements revealed that, during air ingress, HTGRs would experience degraded radiating heat transfer due to graphite oxidation over certain temperature range. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Feature selection based on PCA and the transient identification in nuclear power plants - 289

For the sake of enhancing safety and achieving more economic benefits, it is very important to timely and correctly identify the transients during the operation of nuclear power plants. There are thousands of monitoring signals in a nuclear power plant. It is difficult to complete the calculations timely if all the related signals are chosen for an algorithm of transient identification. Moreover, the irrelevant state variables may also do harm to the correctness of the identification process. Thus, one of the key tasks for transient identification is to figure out effective methods to extract the features from state variables. In this paper, we will address an efficient method for feature selection and transient identification based on principal component analysis (PCA). The effectiveness of the method are illustrated by the tests based on the data from the simulator designed for the High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM), which is under construction in China and the construction will be completed around 2017. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Analysis on the Heat Transfer Phenomena Inside the Scaled-Down VHTR RCCS Riser Experiment

Very High Temperature gas-cooled Reactor (VHTR) that is one of GEN-IV reactors has been developed for the purpose of hydrogen production. VHTR is characterized by high operating temperature and stepped-up inherent safety given by passive safety systems such as Reactor Cavity Cooling System (RCCS). In case the active heat removal system losses its capability, RCCS significantly contributes to the inherent safety of VHTR by removing the decay heat transferred from the reactor core to the reactor vessel. Therefore, the performance of RCCS is the key parameter to ensure the safety of the reactor. Several computation and experiment researches on the performance of RCCS have been conducted. There are two types of RCCS depending on its coolant; one is the air-cooled type and the other is water-cooled type. This study focused on the former type. Due to the excessive height of the prototype RCCS, experiment facilities were constructed in reduced scale. Korea Atomic Energy Research Institute (KAERI) constructed the quarter-scale RCCS experiment facility, NACEF and Argonne National Laboratory (ANL) has constructed half-scale experiment facility, NSTF. University of Wisconsin (UW) also has the quarter-scale experiment facility and Texas A and M University has conducted separate effect test on the jet effect at the upper plenum of RCCS. From the reduced scale experiment result, the performance of RCCS such as a heat removal rate of RCCS and a peak temperature of reactor vessel were evaluated. When the reduced-scale result is applied to the prototype, however, a distortion could be brought about due to the reduced height of RCCS in the experiments. Especially, if the heat transfer regime inside the riser which roles as a heat exchanger to discharge the decay heat into the atmosphere is not identical, the riser wall temperature and the consequent reactor vessel temperature of the prototype could be incorrectly predicted. Thus, the convective heat transfer phenomena inside the riser duct both in the reduced scale and in the prototype is necessary to be investigated. In this study, the preliminary analysis on the heat transfer phenomena inside the RCCS riser was conducted. The subjects of the analysis were the prototype RCCS riser of PMR 200, and its scaled-down experiment facilities, NACEF. First, the geometry, flow and thermal conditions of the riser in PMR200 RCCS and NACEF were introduced with a scaling method. And then, the heat transfer regimes and the heat transfer coefficient corresponding to each condition were predicted using correlation and CFD calculation. Finally, planned activities including the experiment was introduced. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Historical Savannah River Site Tritium Releases, Pathways, and Transport

† Releases of tritium from the production reactors were due almost entirely to releases of moderator, either as liquid or vapor from the reactor systems and from the heavy water rework facility. Small amounts could have been released from the reactor blanket gas system and from irradiated lithium-bearing control rods and targets stored underwater in the disassembly basins (Longtin et al. 1973; Jacober et al. 1973). ‡Only a part of the tritium released to the seepage and retention basins leave the site. Of the 3,015,000 Ci of tritium released to the basins, 27% (0.81M Ci) evaporated from the basin surfaces, 40% (1.2M Ci) migrated to outcrops and entered surface streams, and 33% (0.99M Ci) remains under the basins.

Chiswell, Steven R.↗
DOE OSTI2023

Modeling a Gas-Cooled Microreactor Balance of Plant for the Virtual Test Bed

High Temperature Gas-cooled Reactors (HTGRs) are one of the technology considered for microreactor designs because of their high performance and technological maturity. Several designs are being developed. These microreactor cores are connected to power conversion system (PCS) through a heat exchanger. For transient analysis, the feedback of the PCS or of the primary loop can be important, therefore, it is necessary to develop robust models for the balance of plant that can be coupled to a core model. In this work, a balance of plant model using the MOOSE Thermal Hydraulics module (THM) is presented and demonstrated for startup and load-follow transients. This model can be seamlessly coupled with any MOOSE-based multi-physics core model using the MOOSE multi-app approach.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

AGR Fuel Oxidation Testing Summary and Plans

This presentation describes the goals, scope, and plans for oxidation testing of (primarily irradiated) AGR TRISO fuels and related materials (such as graphite and graphitic matrix material). This presentation will be given at the annual program review meeting for the Gas-cooled Reactor program hosted at INL starting on July 16.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

RELAP5-3D Modeling of the OECD-NEA HTTF Benchmark

HTTF at Oregon State University (OSU) Reference: General Atomics’ modular high-temperature gas-cooled reactor Helium cooled, electrically heated (2.2 MW) Prismatic graphite blocks in the core and reflectors Alumina ceramic blocks are used to simulate the core and top and bottom reflectors One-fourth scale in length and diameter Most of the coolant channels in the core are full scale Lower pressure compared to the prototype reactor (0.7 MPa) Over 500 instruments Designed primarily to investigate depressurized (DCC) and pressurized (PCC) conduction cooldown transients

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

RELAP5-3D HTGR Validation Work at Idaho National Laboratory

Prismatic block-type high-temperature gas-cooled reactors (HTGRs) were built in the United States decades ago, and now advanced reactor vendors are seeking to deploy them again for a variety of applications. Deploying these reactors requires modelling and simulation tools that have been validated against conditions representative of the HTGR application. Idaho National Laboratory (INL) is leading the execution an of HTGR thermal hydraulics benchmark to accelerate the validation of thermal hydraulics modelling and simulation tools for these applications. That benchmark is based on a facility called the High Temperature Test Facility (HTTF). This work provides an overview of work conducted at INL over the last 2 years to validate RELAP5-3D against data from HTTF. This presentation shows results from multiple RELAP5-3D models and an HTTF experiment to assess the impact of certain modelling assumptions on results. The contents of this talk sit on the cutting edge of RELAP5-3D validation for HTGR analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Code Benchmark of Depressurized Conduction Cooldown Transient in the High Temperature Test Facility

This paper presents results from modeling of a depressurized conduction cooldown (DCC) transient at the High Temperature Test Facility (HTTF) as part of the OECD-NEA Thermal Hydraulics Code Validation Benchmark for High-Temperature Gas-Cooled Reactors using HTTF Data . This paper briefly describes the benchmark and the models being used. It then presents a comparison of steady state and transient results based on the Problem 2 Exercise 1A and 1B definitions. We compare block and helium temperature distributions, mass flow distribution, and energy balance in steady state. All models show comparable mass flow distributions and energy balances. The temperatures within the core and outer regions are comparable in all models too, but inner reflector temperatures can vary significantly. Despite that, we find that the models are in good agreement for the full-power steady state. In the DCC, we look at block temperature at the core midplane and RCCS water exit temperature. The INL and ANL models are found to be in excellent agreement with one another on block temperature over time, while the agreement when the KAERI and NRG models are added into consideration is good. Differences in the transient heat removal from the RCCS cause the differences in block temperature over time in these models. The CNL models show similar trends to the INL, ANL, KAERI, and NRG models, but the temperatures are high because the volumes used in calculating the average temperature include the heater rods in the CNL models only. The HUN-REN model shows results that suggest significantly lower heat removal in the RCCS which merit further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗