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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Endurance test of helium circulator for HTR-PM - 18531

High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) will be built as the world's first pebble-bed type modular HTGR commercial demonstration plant according to Chinese National science and technology program. HTR-PM will be expected to begin generating electricity in the end of 2017 in Shidao bay of China. Helium circulator is one of critical component of HTR-PM, which circulates reactor coolant through the reactor core and steam generators. The helium circulator was designed to be a vertical layout and its rotor was supported by active magnet bearings (AMBs). The whole circulator including motor and cooler was sealed in a pressure vessel. Since it is a large mechanical and electrical equipment and the design is totally advanced integrated with many new technologies, experiment study of helium circulator would be significant for safe and reliable operation of HTR-PM plant in the future. For the lack of engineering experience of this equipment, it was necessary to manufacture a prototype for tests. Meanwhile, Engineering Test Facility- Helium Circulator. ETF-HC was built in Institute of Nuclear and New Energy Technology (INET) for the tests. The prototype is completely same as the real machine for the plant. After prototype fabrication, nitrogen test was firstly conducted in the factory by the manufacturer, designer and users. After successful factory test, the prototype was delivered to INET to be tested in ETF-HC. Although nitrogen simulation test had been completed successfully, helium endurance test has been continually carried out in ETF-HC for one and half years because of several sealing problems caused by helium. After solving all kinds of problem in the field, helium circulator run continuously 50 hours under rated condition. Furthermore, 50 service cycles test was conducted in ETF-HC. More than 500 operation hours was accumulated during endurance test. Throughout the final performance test, a conclusion can be made that the helium circulator performance met the design requirements and its feasibility, availability and reliability was verified. (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

A fully integrated 3D CFD model of the flow and heat transfer in a packed bed reactor - 18578

The physical phenomena associated with the modes of heat transfer within a packed pebble bed can be modelled by explicitly solving the associated partial differential equations using a finite volume method approach. Convection from the spheres to coolant, conduction inside and between spheres as well as radiation between various components of a packed bed reactor are the main modes of heat transfer, with several lesser, albeit contributing, phenomena also present. Advances both in software and computing power allow for a larger portion of a packed bed to be modeled with greater accuracy by integrating all applicable physics models in one simulation. A representative unstructured packed bed of uniform spheres was generated with a discrete element method (DEM) and the flow domain modelled using the CFD software package Star-CCM+, taking into account all primary modes of heat transfer as well as a user-specified approach to take secondary effects into account. Effects in the bulk, near-wall and wall-contact regions were investigated. By using such a fully integrated explicit model, a deeper understanding can be can be gained into the role of the underlying phenomena and their respective contributions to the overall heat transfer process. Fuel spheres were modeled explicitly with the conduction between spheres, bed and reflectors being prescribed by contact heat transfer theory as a function of temperature and compressive forces within the bed. The effective thermal conductivity (k{sub eff}), flow pressure drop, as well as temperature distributions in the spheres and coolant were calculated and compared with experimental results from published literature. Good agreement was found with experimental results and the significant contribution of radiation as well as inter-pebble conduction to the effective thermal conductivity could be observed from the results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Evaluation of fission product releases obtained from KueFA safety tests in light of the setup's accuracy - 18692

The Cold Finger Apparatus (KuehlFinger-Apparatur - KueFA) at the European Commission's JRC-Karlsruhe has been designed to assess the effects of Depressurization Loss of Forced Circulation (D-LOFC) accident scenarios on spherical irradiated High Temperature Reactor (HTR) fuel. At approximately 1600 deg. C, the reference maximum temperature for D-LOFC accidents, it has been observed that high-quality German HTR fuel pebbles release only a limited quantity of Cs but a significant amount of Ag. This paper revisits fission product release data obtained from the recently KueFA-tested fuel pebbles HFR-K5/3 and HFR-EU1/3 and presents associated investigations to improve the accuracy of the experimental setup. In light of these findings, implications for previously conducted KueFA heating tests are discussed, considering both experimental data as well as simulated release curves calculated with the Source Term Analysis Code system (STACY). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Recent Improvements in Pronghorn for Advanced Reactor Modeling

Pronghorn is a thermal-hydraulics computational tool developed using the Idaho National Laboratory's Multiphysics Object-Oriented Simulation Environment (MOOSE). It is designed to support Computational Fluid Dynamics (CFD) modeling, ranging from subchannel and porous media analysis to Reynolds Averaged Navier-Stokes (RANS) turbulence modeling. As an integral part of the MOOSE-based suite of tools, Pronghorn seamlessly couples with other MOOSE-based applications to simulate a variety of physical phenomena. This article highlights recent significant enhancements to Pronghorn's CFD modeling capabilities and demonstrates their application to advanced nuclear reactor designs. The recent improvements in Pronghorn primarily focus on modifications to its turbulence modeling capabilities, near-wall corrections and numerical schemes. In terms of turbulence modeling, the two-equation $k-\epsilon$ and $k-\omega$ SST models have been implemented and validated with both equilibrium and non-equilibrium wall treatments. Additionally, corrections for wall roughness, and curvature, and wall-channeling in pebble beds have been introduced in the near-wall modeling. These developments enable more accurate simulations of advanced nuclear reactors. Two case studies are presented in this work: a pool-type Molten Chloride Reactor and a salt-cooled Pebble-Bed High Temperature Reactor. In both cases, the previous models in Pronghorn are compared with the new implementations, demonstrating the improved accuracy achieved with the updated models.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The IAEA coordinated research programme on HTGR uncertainty analysis: Phase I status and Ex. I-1 prismatic reference results

The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied, in contrast to the historical approach where sensitivity analysis were performed and uncertainties then determined by a simplified statistical combination of a few important input parameters. New methodologies are currently under development in the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity. High Temperature Gas-cooled Reactor (HTGR) designs require specific treatment of the double heterogeneous fuel design and large graphite quantities at high temperatures. The IAEA has therefore launched a Coordinated Research Project (CRP) on HTGR Uncertainty Analysis in Modelling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the Chinese HTR-PM. Work has started on the first phase and the current CRP status is reported in the paper. A comparison of the Serpent and SCALE/KENO-VI reference Monte Carlo results for Ex. I-1 of the MHTGR-350 design is also included. It was observed that the SCALE/KENO-VI Continuous Energy (CE) k ∞ values were 395 pcm (Ex. I-1a) to 803 pcm (Ex. I-1b) higher than the respective Serpent lattice calculations, and that within the set of the SCALE results, the KENO-VI 238 Multi-Group (MG) k ∞ values were up to 800 pcm lower than the KENO-VI CE values. The use of the latest ENDF-B-VII.1 cross section library in Serpent lead to ~180 pcm lower k ∞ values compared to the older ENDF-B-VII.0 dataset, caused by the modified graphite neutron capture cross section. Furthermore, the fourth beta release of SCALE 6.2 likewise produced lower CE k∞ values when compared to SCALE 6.1, and the improved performance of the new 252-group library available in SCALE 6.2 is especially noteworthy. A SCALE/TSUNAMI uncertainty analysis of the Hot Full Power variant for Ex. I-1a furthermore concluded that the 238 U(n,γ) (capture) and 235 U(View the MathML source) cross-section covariance matrices contributed the most to the total k ∞ uncertainty of 0.58%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

SCALE Input and Result Files Supporting SCALE Inventory and Reactivity Analysis as Part of the Hermes 2021 PSAR Review

This dataset contains input and result files of computational simulations with the SCALE code system. The simulations cover radionuclide inventory and reactivity analyses of a fluoride salt-cooled high temperature pebble-bed reactor (PB-FHR), specifically the Hermes low-power PB-FHR demonstration reactor. Users wanting to reproduce results from this dataset are required to obtain a license to the SCALE code system for which details on the distribution can be found here: https://www.ornl.gov/scale/releases

equilibrium core↗
DOE OSTI2017

Identification and Characterization of Thermal Fluid Phenomena Associated with Selected Operating/Accident Scenarios in Modular High Temperature Gas-cooled Reactors

Dominant phenomena are identified and characterized for the loss-of-forced-convection (LOFC) scenario, the depressurized loss-of-forced-convection (DLOFC) scenario, and the moisture ingress scenario with relief valves available scenario. These scenarios are considered for the modular high temperature gas-cooled reactor (MHTGR) baseline design—with both a prismatic core and a pebble-bed core. The objectives of this effort, described in the above paragraph, were achieved by following a methodology identified by the U.S. Nuclear Regulatory Commission (NRC) and recommended in their Regulatory Guide 1.203. The methodology employed is described. For the above scenarios, baseline phenomena identification and ranking tables (PIRTs) generated by a NRC-sponsored effort in 2008 for the LOFC and DLOFC scenarios and a baseline PIRT generated by a Department of Energy-sponsored effort in 2011 by their Next Generation Nuclear Plant (NGNP) Moisture Ingress Assessment Committee were used to identify the dominant phenomena and existing levels-of-knowledge. Ten dominant phenomena (with only medium or low levels-of-knowledge) were identified for the LOFC and DLOFC scenarios including: (1) Inlet plenum stratification and plumes, (2) core coolant flow distribution, (3) radiant heat transfer from the top of the core to the upper vessel head, (4) reactor cavity air circulation and heat transfer, (5) reactor cavity cooling system behavior, (6) decay heat distributions, (7) core effective thermal conductivity, (8) molecular diffusion in air ingress DLOFC, (9) duct exchange flow, (10) phenomena that affect cavity gas composition for air-ingress. Five dominant phenomena, with high levels of knowledge, were identified for the moisture-ingress scenario with relief valves available including: (1) flow through break, (2) moisture level in the primary, (3) transport to moisture monitor, (4) moisture monitor instrument response, and (5) heat removal by shutdown cooling system. These phenomena are discussed and characterized. A listing of experiments—both completed and ongoing—is given with a view toward including data useful for further more detailed examination of the dominant phenomena listed in the previous paragraph. Note that only data for the LOFC and DLOFC scenarios are listed—since the moisture ingress scenario with relief valves available –have been identified as having adequate levels-of-knowledge by the PIRT committee. Finally, a preliminary computational fluid dynamics (CFD) calculation of inlet plenum stratification and plumes is described. Techniques employed for this calculation are typical for preliminary scoping calculations performed to validate numeric tools.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Horizontal Split Table Conceptual Design for Validation of Nuclear Data used in Advanced Reactors [Slides]

A methodology to create conceptual designs of benchmark critical experiments for advanced reactors nuclear data testing and validation was developed. A first concept was explored, pebble-bed high-temperature gas cooled reactor, based on the HTR-10 reactor. The very high correlation is a proof of concept that the design is similar to the application, and performing such critical experiments would help nuclear data testing and validation. This proof-of-concept was included in the IER-539 CED-1:Preliminary Design of a New Horizontal Split Table report. Other concepts could be explored if needed, such as Molten-salt reactor, Sodium-cooled fast reactor or Heat pipe reactors/Microreactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Recent Testing of ENDF/B-VIII.0 at ORNL [Slides]

This presentation provides an overview of comparisons of ENDF/B-VII.1 and ENDF/B-VIII.0 and uses VALID suite to test >600 cases with ENDF/B-VII.1 and ENDF/B-VIII.0 in KENO V.a/KENO-VI, although a limited number of cases have been tested with TSUNAMI. It provides a results summary of KENO V.a/KENO-VI with a pooled results summary. The presentation also discusses changes for LCT experiments, MCT experiments, MST experiments, and PST experiments. A few words about graphite are also provided and the presentation states that graphite performance investigated across ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0 for MSR and HTGR models. A brief summary of the results is presented to make sure the Validation Committee is aware of the work. Also presented is an Eigenvalue comparison with a graphite moderated system, and comparison between an HTR-10 pebble model ENDF/B-VII.1 vs. ENDF/B-VIII.0, which provides different graphite data. In conclusion, ORNL has recently completed a comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 using the VALID library. Work is also being performed for advanced reactor applications. Other efforts on-going and they are also trying to investigate more intermediate spectrum cases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Irradiated AGR 2 Compact 4-3-1 Examination Plan

This plan describes post-irradiation examination activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory on irradiated Compact 4-3-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with the Statement of Work for PIE and Safety Testing of Pebble Bed Modular Reactor Limited Fuel at ORNL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High Temperature Gas-cooled Reactors: Core Design

General Attributes of Modular Prismatic and Pebble Bed HTGRs Common features and physics Neutronics Thermal-Fluidics Plant Systems and Power Conversion Instrumentation and Control Normal Operation and Power Maneuvers

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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

Study on balance of axial force in helium circulator for HTR-PM - 18490

The helium circulator is one of the key equipment in the first loop of the High Temperature gas-cooled Reactor Pebble bed Module (HTR-PM). The active magnetic bearings are employed in the helium circulator, so that the inlet ring sealing clearance is about 5 to 10 times larger than that if the mechanical bearings are employed. Bigger clearance results in bigger leakage flow and bigger axial force. A new design of balance holes was proposed to reduce the axial aerodynamic force and the design principle was analyzed. To investigate the axial aerodynamic force due to leakage flow, numerical simulation of fluid dynamic was conducted in the impeller with two kinds of inlet ring sealing clearance and four types of balance holes, respectively. The calculation results shows that the balance holes with diameter 12 mm have high performance and that the designed balance holes reduce effectively about 78% of the axial aerodynamic force. The axial force was obtained from the measured real-time electric current of the axial active magnetic bearings in one helium circulator prototype. The comparison of experimental data with the CFD analysis results showed that the new design of balance holes was successful. (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

Reducing DLOFC fuel temperatures by mixing thorium with LEU in a single-zone six-pass fuel cycle in a PBMR-DPP-400 core - 18548

Many studies have been done in neutronics and thermal-hydraulic simulation of the standard 6- pass fuel recirculation scheme for the standard 9.6 wt % enriched, 9 g per fuel sphere low-enriched uranium (LEU) fuel in the PBMR-DPP-400, using different versions of the VSOP diffusion codes. Maximum DLOFC temperatures were all below the upper limit of 1600 deg. C. The DLOFC temperature is highly dependent on the peaking factor of the power; meaning a lower maximum DLOFC temperature can be obtained by suppressing the axial power peak and by moving the radial power peak towards the external reflector. In this study the standard 6-pass fuel recirculation was retained. The improvement strategy was thus attempted by means of flattening the axial power profile by mixing substantial amounts of thorium into the LEU fuel. The addition of thorium led to breading of substantial amounts of {sup 233}U. This led to slower depletion of enrichment with burn-up, which increased fuel reactivity and power densities near the bottom of the core and thus flattened the axial power profile. The effect was a reduction in maximum DLOFC temperature by 44 deg. C. The simulations were made using the VSOP- 99/05 diffusion code. It was further shown that the results obtained are also applicable to the Chinese HTR-PM and the proposed strategies for further improvement can be expected to produce even much better results in Block Reactors than in Pebble Bed Reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The coupled neutronics and thermo-fluid dynamics design characteristics of the XE-100 200 MW{sub TH} reactor - 18552

The Xe-100 is a high temperature helium-cooled, graphite moderated, pebble bed reactor with a 10 times through, multi-pass (MEDUL=MEhrfacDUrchLauf) fueling scheme. The design, characterized as a Generation IV reactor, is based on proven technology that has been demonstrated a number of times both experimentally and as commercial reactors. This proven foundation ensures high levels of design and technical readiness. In the overview provided of the design it is shown how the excess reactivity is limited by continuous refueling, whilst the selected core power density will ensure passive heat removal. In this way the intrinsic safety design features of a typical GEN IV design are guaranteed. It is shown that no upset event will lead to a significant amount of radiological release. The coupled core neutronic and thermo-fluid dynamics design for the Xe-100 is performed with the VSOP-A system of codes. For the equilibrium core, VSOP results demonstrate that the spherical fuel powers (maximum 1.9 kW) and operational temperatures (<900 deg. C) fall well within the envelope of the design criteria. Adequate reactivity control and long-term, cold shutdown are provided by two separate, diversely-actuated systems, while the overall negative reactivity temperature coefficient is illustrated over the total operational range. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Pronghorn Porous Media Model Validation with Pressure Drop Measurements

The verification and validation (V &V) of Pronghorn is imperative to assert its accuracy when predicting the fluid velocity, temperature, and pressure in high temperature gas-cooled reactors. This paper focuses on the validation of the Pronghorn implementation of the incompressible and compressible Navier-Stokes equations using the pressure drop measurements in the engineering-scale pebble bed facility at the Texas A &M university.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗