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Sample records for temperature reactor vhtr

  1. Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; G. Baccaglini; S. Ball; T. Burchell; B. Corwin; T. Fewell; M. Labar; P. MacDonald; P. Rittenhouse; Russ Vollam; F. Southworth

    2003-01-01

    The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.

  2. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

  3. Materials for the very high temperature reactor (VHTR): a versatile nuclear power station for combined cycle electricity and heat production

    Energy Technology Data Exchange (ETDEWEB)

    Hoffelner, W

    2005-07-01

    The International Generation IV Initiative provides a research platform for the development of advanced nuclear plants which are able to produce electricity and heat in a combined cycle. Very high-temperature gas-cooled reactors are considered as near-term deployable plants meeting these requirements. They build on high-temperature gas-cooled reactors which are already in operation. The main parts of such an advanced plant are: reactor pressure vessel, core and close-to-core components, gas turbine, intermediate heat exchanger, and hydrogen production unit. The paper discusses the VHTR concept, materials, fuel and hydrogen production based on discussions on research and development projects addressed within the generation IV community. It is shown that material limitations might restrict the outlet temperature of near-term deployable VHTRs to about 950 {sup o}C. The impact of the high temperatures on fuel development is also discussed. Current status of combined cycle hydrogen production is elaborated on. (author)

  4. Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, DT

    2005-12-15

    A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of forced

  5. Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Kawaji, Masahiro [City College of New York, NY (United States); Valentin, Francisco I. [City College of New York, NY (United States); Artoun, Narbeh [City College of New York, NY (United States); Banerjee, Sanjoy [City College of New York, NY (United States); Sohal, Manohar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-12-21

    The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

  6. Numerical investigation of the High Temperature Reactor (VHTR) using computational fluid dynamics; Investigacao numerica do Reator de Alta Temperatura (VHTR) utilizando fluidodinamica computacional

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Joao Pedro C.T.A.; Santos, Andre A. Campagnole dos; Mesquita, Amir Z., E-mail: jpctap@cdtn.br, E-mail: aacs@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG),Belo Horizonte, MG (Brazil). Lab. de Termo-Hidraulica

    2013-07-01

    This work consists to evaluate and continue the study that is being developed in the Laboratory of Thermo-Hydraulics of the CNEN/CDTN (Centro de Desenvolvimento da Tecnologia Nuclear), aiming to validate the methods and procedures used in the numerical calculations of fluid flow in fuel elements of the core of the VHTR.

  7. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  8. A preliminary neutronic evaluation and depletion study of VHTR and LS-VHTR reactors using the codes: WIMSD5 and MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Fabiano C.; Pereira, Claubia; Costa, Antonella Lombardi; Veloso, Maria Auxiliadora Fortini [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2009-07-01

    It is expected that, in the future, besides electricity generation, reactors should also develop secondary activities, such as hydrogen generation and seawater desalinization. Generation IV reactors are expected to possess special characteristics, like high safety, minimization of radioactive rejects amount and ability to use reprocessed fuel with non-proliferating projects in their cycles. Among the projects of IV generation reactors available nowadays, the (High Temperature Reactors) HTR, are highlighted due to these desirable characteristics. Under such circumstances, such reactor may be able to have significant higher thermal power ratings to be used for hydrogen production, without loose of safety, even in an emergency. For this work, we have chosen two HTR concepts of a prismatic reactor: (Very High Temperature Reactor) VHTR and the (Liquid Salted -Very High Temperature Reactor) LS-VHTR. The principal difference between them is the coolant. The VHTR uses helium gas as a coolant and have a burnup of 101,661 MWd/THM while the LS-VHTR uses low-pressure liquid coolant molten fluoride salt with a boiling point near 1500 de C working at 155,946 MWd/THM. The ultimate power output is limited by the capacity of the passive decay system; this capacity is limited by the reactor vessel temperature. The goal was to evaluate the neutronic behavior and fuel composition during the burnup using the codes (Winfrith Improved Multi-Group Scheme) WIMSD5 and the MCNPX2.6. The first, deterministic and the second, stochastic. For both reactors, burned fuel type 'C' coming from Angra-I nuclear plant, in Brazil, was used with 3.1% of initial enrichment, burnup to 33,000 MWd/THM using the ORIGEN2.1 code, divided in three steps of 11,000 MWd/THM, with an average density power of 37.75 MWd/THM and 5 years of cooling in pool. Finally, the fuel was reprocessed by Purex technique extracting 99.9% of Pu, and the desired amount of fissile material (15%) to achieve the final mixed

  9. Dynamic Simulation of Temperature Transition on the Secondary Helium Loop of a VHTR-SI Hydrogen Production System

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Ji Woon; Shin, Young Joon; Lee, Tae Hoon; Lee, Ki Young; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Youn, Cheong [Chungnam National University, Daejeon (Korea, Republic of)

    2011-10-15

    A sulfur.iodine (SI) cycle coupled to a Very High Temperature Gas Cooled Reactor (VHTR) is one of the leading candidates of thermochemical cycles for hydrogen production. The SI cycle can be divided into three sections based on the chemical reactions: a Bunsen reaction (Section 1), sulfuric acid concentration and decomposition (Section 2), and a hydrogen iodine concentration and decomposition (Section 3). The heat required in the SI cycle can be supplied through an intermediate heat exchanger (IHX) by the VHTR. On the other hand, helium is used as a high-temperature energy carrier gas between the VHTR and the IHX or IHX and the SI cycle. In the SI cycle, the chemical reactors that receive thermal energy from the helium are a sulfuric acid vaporizer, sulfuric acid decomposer, sulfuric trioxide decomposer, and hydriodic acid decomposer including a pre-heating part. To simulate the dynamic behavior of the VHTR-SI cycle, the Korea Atomic Energy Research Institute -Dynamic Simulation Code (KAERI-DySCo) based on the Visual C++ has been prepared by the KAERI research group in 2010. KAERI-DySCo is integration application software, which includes several code modules that can solve the dynamic problem of the seven chemical reactors in the VHTR-SI cycle. In this paper, the dynamic behavior of the temperature transition on the secondary helium loop of the SI cycle has been simulated using KAERI-DySCo

  10. 850/sup 0/C VHTR plant technical description

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    This report describes the conceptual design of an 842-MW(t) process heat very high temperature reactor (VHTR) plant having a core outlet temperature of 850/sup 0/C (1562/sup 0/F). The reactor is a variation of the high-temperature gas-cooled reactor (HTGR) power plant concept. The report includes a description of the nuclear heat source (NHS) and of the balance of reactor plant (BORP) requirements. The design of the associated chemical process plant is not covered in this report. The reactor design is similar to a previously reported VHTR design having a 950/sup 0/C (1742/sup 0/F) core outlet temperature.

  11. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  12. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Science.gov (United States)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  13. Strategy of VHTR Realization

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    High temperature gas cooled reactor has been developed since 1956. Fundamental idea of a gas cooled reactor is to achieve high temperature which is suitable for high efficiency application such as electricity generation. The core is composed of ceramics, graphite blocks which are mechanical stable up to very high temperature. Fuel is ceramics, TRISO ( tri-isotropic coated micro particle) whose dense coating layers work as small radioactivity containment. Coolant is inert gas, helium, which is stable chemically, neutronically, and thermal hydraulically. Several test reactors such as DRE, PB-1, FSV, AVR, THTR, HTTR, HTR-10 were built and demonstrated their safety. Large GA-HTR, RSA-PBMR projects are canceled and US-NGNP project is idling. Only Chinese HTR-PM demonstrator is under construction. HTGR has long history of development. For realization and market penetration, VHTR community should look at niche market such as carbon free energy supply to industry complex, electric power for small grid, carbon free hydrogen production, power source for space colony. Technology Readiness Level should be advanced to get proper investment from industry. For this, cooperation between international R and D institutions is required. Clearly divided role between universities, research institutions, and industries will reduce complication and shorten VHTR realization day.

  14. Development of essential technology for VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Kang, G. H.; Koo, G. H.; Kim, Y. W.; Kim, D. H.; Kim, D. Jong; Kim, D. O.; Kim, D. Jin; Kim, M. C.; Kim, M. C.

    2012-04-15

    The research tasks performed in this project can be classified into five categories; high temperature material of VHTR reactor and components for hydrogen production, the nuclear graphite for the core material, the essential technologies for VHTR components, Process Heat Exchanger (PHE) fabrication, and gas loop for PHE verification tests. Research tasks on high temperature materials of VHTR reactor and components include creep properties of super alloy for high temperature components, properties of a modified 9Cr-1Mo alloy, fabrication and properties of in-core ceramic composites, and corrosion properties of the materials for the sulfuric acid decomposer. The echnologies of graphitization evaluation, nondestructive defect detection, and impurity analysis were developed in field of nuclear graphites. The properties of graphites were evaluated by tests using small specimen test. The abroad status of graphite machining was reviewed. Review about the status of VHTR components, structural sizing and analysis for hot gas duct, thermal sizing of IHX were performed in the field of the essential technologies for VHTR components. The surface modification process with ion beam mixing was optimized and evaluated for the fabrication of process heat exchanger (PHE). The secondary sulfuric acid loop was designed and constructed in the gas loop. The lab-scale PHE test was performed in the gas loop.

  15. Topical report : NSTF facilities plan for water-cooled VHTR RCCS : normal operational tests.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C. P.; Lomperski, S.; Aeschlimann, R. W.; Nuclear Engineering Division

    2006-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the gas-cooled Very High Temperature Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept.

  16. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  17. Prismatic VHTR neutronic benchmark problems

    Energy Technology Data Exchange (ETDEWEB)

    Connolly, Kevin John, E-mail: connolly@gatech.edu [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States); Rahnema, Farzad, E-mail: farzad@gatech.edu [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States); Tsvetkov, Pavel V. [Department of Nuclear Engineering, Texas A& M University, College Station, TX (United States)

    2015-04-15

    Highlights: • High temperature gas-cooled reactor neutronics benchmark problems. • Description of a whole prismatic VHTR core in its full heterogeneity. • Modeled using continuous energy nuclear data at a representative hot operating temperature. • Benchmark results for core eigenvalue, block-averaged power, and some selected pin fission density results. - Abstract: This paper aims to fill an apparent scarcity of benchmarks based on high temperature gas-cooled reactors. Within is a description of a whole prismatic VHTR core in its full heterogeneity and modeling using continuous energy nuclear data at a representative hot operating temperature. Also included is a core which has been simplified for ease in modeling while attempting to preserve as faithfully as possible the neutron physics of the core. Fuel and absorber pins have been homogenized from the particle level, however, the blocks which construct the core remain strongly heterogeneous. A six group multigroup (discrete energy) cross section set has been developed via Monte Carlo using the original heterogeneous core as a basis. Several configurations of the core have been solved using these two cross section sets; eigenvalue results, block-averaged power results, and some selected pin fission density results are presented in this paper, along with the six-group cross section data, so that method developers may use these problems as a standard reference point.

  18. A Research Status on High-Temperature Creep of Alloy 617 for Use in VHTR System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woo-Gon; Park, Jae-Young; Kim, Eung-Seon; Kim, Yong-Wan; Kim, Min-Hwan [KAERI, Daejeon (Korea, Republic of); Kim, Seon-Jin [Pukyong National Univ., Busan (Korea, Republic of)

    2016-05-15

    In this study, a research status on creep works of Alloy 617 conducting at KAERI was introduced and summarized. Various experimental creep data and creep constants obtained in the air/helium environments and base/weld metals were presented and discussed using various creep equations and parameters. The draft Code Case is a modification from ASME Section III Subsection NH that was put forth by a special task force of the ASME subgroup that deals with elevated temperature design. The primary intended application of the draft Code Case is a VHTR. Presently, various creep data for Alloy 617 are being accumulated through Generation-IV forum (GIF) Material Handbook Database of a next-generation nuclear plant research and development. As per this, a new Alloy 617 Code Case is planned to be approved by 2017. However, to do so, various creep data and creep constants in air/helium environments, and base/weld metals etc. should be obtained to help draft the new Code Case, and creep behavior should be investigated through systematic analysis of a wide range of creep temperature and stress conditions. Using various creep equations and parameters, the creep constants were determined for design use of Alloy 617. The stress of the He environment was more reduced than that of the air one. As the stress increases, the creep rate of WM was significantly lower than that of BM. The reason for this was that the rupture elongation of WM was largely reduced compared with that of BM.

  19. New experimental device for VHTR structural material testing and helium coolant chemistry investigation - High Temperature Helium Loop in NRI Rez

    Energy Technology Data Exchange (ETDEWEB)

    Berka, Jan, E-mail: bej@cvrez.cz [Research Centre Rez, Ltd, Husinec-Rez 130, 25068 Rez (Czech Republic); Institute of Chemical Technology Prague, Technicka 1905, 16628 Prague 6 (Czech Republic); Matecha, Josef, E-mail: josef.matecha@ujv.cz [Nuclear Research Institute Rez plc., Husinec-Rez 130, 25068 Rez (Czech Republic); Cerny, Michal [Institute of Chemical Technology Prague, Technicka 1905, 16628 Prague 6 (Czech Republic); Viden, Ivan, E-mail: ivan.viden@vscht.cz [Institute of Chemical Technology Prague, Technicka 1905, 16628 Prague 6 (Czech Republic); Sus, Frantisek [Research Centre Rez, Ltd, Husinec-Rez 130, 25068 Rez (Czech Republic); Nuclear Research Institute Rez plc., Husinec-Rez 130, 25068 Rez (Czech Republic); Hajek, Petr [Nuclear Research Institute Rez plc., Husinec-Rez 130, 25068 Rez (Czech Republic)

    2012-10-15

    The High Temperature Helium Loop (HTHL) is an experimental device for simulation of VHTR helium coolant conditions. The purpose of the HTHL is structural materials testing and helium coolant chemistry investigation. In the HTHL pure helium will be used as working medium and its main physical parameters are 7 MPa, max. temperature in the test section 900 Degree-Sign C and flow rate 37.8 kg/h. The HTHL consists of an active channel, the helium purification system, the system of impurities dosage (e.g. CO, CO{sub 2}, H{sub 2}, H{sub 2}O, O{sub 2}, N{sub 2}, and CH{sub 4}) and the helium chemistry monitoring system (sampling and on-line analysis and determination of impurities in the helium flow). The active channel is planned to be placed into the core of the experimental reactor LVR-15 which will serve as a neutron flux source (max. 2.5 Multiplication-Sign 10{sup 18} n/m{sup 2} s for fast neutrons). The HTHL is now under construction. Some of its main parts are finished, some are still being produced (active channel internals, etc.), some should be improved to work correctly (the helium circulatory compressor); certain sub-systems are planned to be integrated to the loop (systems for the determination of moisture and other impurities in helium, etc.). The start of the HTHL operation is expected during 2011 and the integration of the active channel into the LVR-15 core during 2012.

  20. Evaluation of the DRAGON code for VHTR design analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-12

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

  1. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  2. A probabilistic consequence assessment for a very high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

    2017-02-15

    Currently, fossil fuel is globally running out. If current trends continue, crude oil will be depleted in 20 years and natural gas in 40 years. In addition, the use of fossil resource has increased emissions of green gas such as carbon dioxide. Therefore, there has been a strong demand in recent years for producing large amounts of hydrogen as an alternative energy [1]. To generate hydrogen energy, very high temperature more than 900 C is required but this level is not easy to reach. Because a Very High Temperature Reactor (VHTR), one of next generation reactor, is able to make the temperature, it is regarded as a solution of the problem. Also, VHTR has an excellent safety in comparison with existing and other next generation reactors. Especially, a passive system, Reactor Cavity Cooling System (RCCS), is adopted to get rid of radiant heat in case of accidents. To achieve variety requirements of new designed-reactors, however, it needs to develop new methodologies and definitions different with existing method. At the same time, an application of probability safety assessment (PSA) has been proposed to ensure the safety of next generation NPPs. For this, risk-informed designs of structures have to be developed and verified. Particularly, the passive system requires to be evaluated for its reliability. The objective of this study is to improve safety of VIITR by conducting risk profile.

  3. Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

    2016-10-15

    A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

  4. The use of a very high temperature nuclear reactor in the manufacture of synthetic fuels

    Science.gov (United States)

    Farbman, G. H.; Brecher, L. E.

    1976-01-01

    The three parts of a program directed toward creating a cost-effective nuclear hydrogen production system are described. The discussion covers the development of a very high temperature nuclear reactor (VHTR) as a nuclear heat and power source capable of producing the high temperature needed for hydrogen production and other processes; the development of a hydrogen generation process based on water decomposition, which can utilize the outputs of the VHTR and be integrated with many different ultimate hydrogen consuming processes; and the evaluation of the process applications of the nuclear hydrogen systems to assess the merits and potential payoffs. It is shown that the use of VHTR for the manufacture of synthetic fuels appears to have a very high probability of making a positive contribution to meeting the nation's energy needs in the future.

  5. VHTR, ADS, and PWR spent nuclear fuel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salome, J.A.D.; Cardoso, F.; Velasquez, C.E.; Pereira, F.; Pereira, C. [Departamento de Engenharia Nuclear - Escola de Engenharia Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP: 31270-901 (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores - CNPq, Rio de Janeiro (Brazil); Barros, G.P. [Comissao Nacional de Energia Nuclear - CNEN, Rua General Severiano 82, Botafogo, Rio de Janeiro, RJ, CEP: 22290-040 (Brazil)

    2016-07-01

    The aim of this study is to analyze and compare the discharged-spent fuel of 3 types of nuclear systems: a Very High-Temperature Gas Reactor (VHTR), a lead-cooled Accelerator-Driven System (ADS) and a standard Pressurized Water Reactor (PWR). The two first systems, VHTR, and ADS were designed to use reprocessed fuels. UREX+ and GANEX techniques were used for the reprocessing processes respectively. The fuel burnup simulated for the systems in other works have been used to obtain the final composition of the spent fuel discharged. After discharge, the radioactivity, the radiotoxicity, and the decay heat were evaluated through the ORIGEN 2.1 code until 10{sup 7} years and compared to the literature. The spent nuclear waste (SNF) coming from reprocessing techniques and burned up in advanced reactors show that the radiotoxicity decreases below a conventional SNF from a typical PWR for the time studied. The VHTR and ADs have higher values of radioactivity, radiotoxicity and decay heat, because of the greater concentrations of plutonium and curium in these reactors than in the PWR. Fission products have the greatest contribution for the first 25 years over the parameters studied for a PWR. The most harmful fission products are: Ba{sup 137m}, Tc{sup 99}, I{sup 129} and Nb{sup 93m} and for actinides is the plutonium and curium.

  6. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  7. Critical evaluation of high-temperature gas-cooled reactors applicable to coal conversion

    Energy Technology Data Exchange (ETDEWEB)

    Spiewak, I.; Jones, J.E. Jr.; Rittenhouse, P.L.; DeStefano, J.R.; Delene, J.G.

    1975-12-01

    A critical review is presented of the technology and costs of very high-temperature gas-cooled reactors (VHTRs) applicable to nuclear coal conversion. Coal conversion processes suitable for coupling to reactors are described. Vendor concepts of the VHTR are summarized. The materials requirements as a function of process temperature in the range 1400 to 2000/sup 0/F are analyzed. Components, environmental and safety factors, economics and nuclear fuel cycles are reviewed. It is concluded that process heat supply in the range 1400 to 1500/sup 0/F could be developed with a high degree of assurance. Process heat at 1600/sup 0/F would require considerably more materials development. While temperatures up to 2000/sup 0/F appear to be attainable, considerably more research and risk were involved. A demonstration plant would be required as a step in the commercialization of the VHTR.

  8. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  9. Emissivity of Candidate Materials for VHTR Applicationbs: Role of Oxidation and Surface Modification Treatments

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Allen, Todd; Anderson, Mark; Cao, Guoping; Kulcinski, Gerald

    2011-07-25

    The Generation IV (GEN IV) Nuclear Energy Systems Initiative was instituted by the Department of Energy (DOE) with the goal of researching and developing technologies and materials necessary for various types of future reactors. These GEN IV reactors will employ advanced fuel cycles, passive safety systems, and other innovative systems, leading to significant differences between these future reactors and current water-cooled reactors. The leading candidate for the Next Generation Nuclear Plant (NGNP) to be built at Idaho National Lab (INL) in the United States is the Very High Temperature Reactor (VHTR). Due to the high operating temperatures of the VHTR, the Reactor Pressure Vessel (RPV) will partially rely on heat transfer by radiation for cooling. Heat expulsion by radiation will become all the more important during high temperature excursions during off-normal accident scenarios. Radiant power is dictated by emissivity, a material property. The NGNP Materials Research and Development Program Plan [1] has identified emissivity and the effects of high temperature oxide formation on emissivity as an area of research towards the development of the VHTR.

  10. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

  11. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  12. Computer Program for Equipment Sizing on the Secondary Helium Loop of a VHTR-SI Hydrogen Production System

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Ji Woon; Kim, Ji Hwan; Shin, Young Joon; Lee, Tae Hoon; Lee, Ki Young; Kim, Yong Wan; Chang, Jong Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Youn, Cheong [Chungnam National University, Daejeon (Korea, Republic of)

    2010-10-15

    The Sulfur-Iodine (SI) process coupled to a very high temperature gas-cooled reactor (VHTR) is well known as a feasible technology to produce hydrogen. The SI process can be divided into three sections based on the chemical reactions a Bunsen reaction (Section 1), a Sulfur acid concentration and decomposition (Section 2) and a hydrogen iodine concentration and decomposition (Section 3). The heat required in the SI process can be supplied through an intermediate heat exchanger (IHX) by the VHTR. On the other hand, Helium is used as a high temperature energy carrier gas between the VHTR and the IHX or the IHX and the SI process. In the SI process, chemical reactors that receive thermal energy from the helium are: A sulfuric acid vaporizer, a sulfuric acid decomposer, a sulfuric trioxide decomposer, and a hydriodic acid decomposer including pre-heating par. The sizing of such chemical reactors is essentially required for dynamic simulation of the VHTR-SI process. In this paper, a sizing program which can calculate the sizing of each chemical reactor to satisfy mass and energy balance at a given operation condition has been introduced. A graphic user interface (GUI) to visualize calculated sizing results of the chemical reactors has been added for the effective program operation

  13. Designing a TAC thermometer from a VHTR graphite structure

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James A., E-mail: James.Smith@INL.gov; Kotter, Dale, E-mail: James.Smith@INL.gov [Fuel Performance and Design, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Garrett, Steven L.; Ali, Randall A. [Graduate Program in Acoustics, Penn State University, State College, PA (United States)

    2015-03-31

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. Very High Temperature Reactors are pushing the in core temperatures even higher. A unique sensing approach will be discussed to address the necessary high temperature measurements. Thermoacoustic thermometry exploits high temperatures and uses materials that are immune to the effects of ionizing radiation to create a temperature sensor that is self-powered and wireless. In addition, the form-factor for the Thermoacoustic Thermometer (TACT) can be designed to be integrated within common in-pile structures. There are no physical moving parts required for TACT and the sensor is self-powered, as it uses the nuclear fuel for its heat source. TACT data will be presented from a laboratory prototype mimicking the design necessary for a VHTR graphite structure.

  14. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    Energy Technology Data Exchange (ETDEWEB)

    Marvin, M.D.

    1978-10-31

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated. (FS)

  15. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  16. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  17. A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ed Gorski; Dennis Harrell; Finis Southworth

    2004-09-01

    The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory

  18. Fuel particles for high temperature reactors; Combustibles a particules pour reacteurs a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Pheip, M. [CEA Cadarache (DEN/CAD/DEC/SESC/LIPA), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles; Masson, M. [CEA Valrho, Dept. Radiochimie et Procedes, 30 (France); Perrais, Ch. [CEA Cadarache (DEN/DEC/SPUA), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles; Pelletier, M. [CEA Cadarache (DEN/DEC/SESC), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles

    2007-07-15

    The concept of fuel particles with a millimeter size was born at the end of the 1950's and is the reference concept of high or very high temperature gas-cooled reactors (HTR/VHTR). The specificity of this fuel concerns its fine divided structure, its all-ceramic composition and its micro-confining properties with respect to fission products. These 3 properties when combined together allow the access to high temperatures and to a high level of safety. This article presents: 1 - the general properties of particle fuels; 2 - the fabrication and control of fuel elements: nuclei elaboration processes, vapor deposition coating of nuclei, shaping of fuel elements, quality control of fabrication; 3 - the fuel particles behaviour under irradiation: mechanical and thermal behaviour, behaviour and diffusion of fission products, ruining mode; 4 - the reprocessing of particle fuels: stakes and options, direct storage, separation of constituents, processing of carbonous wastes; 5 - conclusion. (J.S.)

  19. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  20. Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

    Energy Technology Data Exchange (ETDEWEB)

    Fok, Alex

    2013-10-30

    The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the model to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.

  1. High temperature reactors for cogeneration applications

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl [Forschungszentrum Juelich (Germany). IEK-6; Allelein, Hans-Josef [Forschungszentrum Juelich (Germany). IEK-6; RWTH Aachen (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik (LRST)

    2016-05-15

    There is a large potential for nuclear energy also in the non-electric heat market. Many industrial sectors have a high demand for process heat and steam at various levels of temperature and pressure to be provided for desalination of seawater, district heating, or chemical processes. The future generation of nuclear plants will be capable to enter the wide field of cogeneration of heat and power (CHP), to reduce waste heat and to increase efficiency. This requires an adjustment to multiple needs of the customers in terms of size and application. All Generation-IV concepts proposed are designed for coolant outlet temperatures above 500 C, which allow applications in the low and medium temperature range. A VHTR would even be able to cover the whole temperature range up to approx. 1 000 C.

  2. TRISO fuel thermal simulations in the LS-VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, Mario C.; Scari, Maria E.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F., E-mail: marc5663@gmail.com, E-mail: melizabethscari@yahoo.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2017-07-01

    The liquid-salt-cooled very high-temperature reactor (LS-VHTR) is a reactor that presents very good characteristics in terms of energy production and safety aspects. It uses as fuel the TRISO particles immersed in a graphite matrix with a cylindrical shape called fuel compact, as moderator graphite and as coolant liquid salt Li{sub 2}BeF{sub 4} called Flibe. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core by performing different simplifications to represent the reactor core and the fuel compact under steady-state conditions, starting the modeling from a single fuel element, until complete the studies with the entire core model developed in the RELAP5-3D code. Two models were considered for representation of the fuel compact, homogeneous and non-homogeneous models, as well as different geometries of the heat structures was considered. The aim to develop several models was to compare the thermal hydraulic characteristics resulting from the construction of a more economical and less discretized model with much more refined models that can lead to more complexes analyzes to representing TRISO effect particles in the fuel compact. The different results found, mainly, for the core temperature distributions are presented and discussed. (author)

  3. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz

    2012-04-01

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  4. In Situ Measurements of Spectral Emissivity of Materials for Very High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    G. Cao; S. J. Weber; S. O. Martin; T. L. Malaney; S. R. Slattery; M. H. Anderson; K. Sridharan; T. R. Allen

    2011-08-01

    An experimental facility for in situ measurements of high-temperature spectral emissivity of materials in environments of interest to the gas-cooled very high temperature reactor (VHTR) has been developed. The facility is capable of measuring emissivities of seven materials in a single experiment, thereby enhancing the accuracy in measurements due to even minor systemic variations in temperatures and environments. The system consists of a cylindrical silicon carbide (SiC) block with seven sample cavities and a deep blackbody cavity, a detailed optical system, and a Fourier transform infrared spectrometer. The reliability of the facility has been confirmed by comparing measured spectral emissivities of SiC, boron nitride, and alumina (Al2O3) at 600 C against those reported in literature. The spectral emissivities of two candidate alloys for VHTR, INCONEL{reg_sign} alloy 617 (INCONEL is a registered trademark of the Special Metals Corporation group of companies) and SA508 steel, in air environment at 700 C were measured.

  5. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

    2006-12-15

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA.

  6. A Preliminary Investigation of Rapid Depressurization Phenomena Following a Sudden DLOFC in a VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Richard C. Martineau; Ray A. Berry; Dana A. Knoll

    2009-03-01

    Air ingress has been identified as a potential threat for Very High Temperature gas-cooled Reactors (VHTR). Reactor components constructed of graphite will, at high temperatures, produce exothermic reactions in the presence of oxygen. The danger lies in the possibility of fuel element damage and core structural failure. Previous investigations of air ingress mechanisms have focused on thermal and molecular diffusion, density-driven stratified flow, and natural convection. Here, we investigate the possibility of a rapid ingress of air due to a Taylor wave expansion after a hypothetical sudden loss of coolant accident (LOCA) scenario in a VHTR. Our analysis starts with a one-dimensional shock tube simulation to simply illustrate the development of a Taylor wave with resulting reentrant flow. Then, a simulation is performed of an idealized two-dimensional axisymmetric representation of the lower plenum of General Atomics GT-MHR subjected to a hypothetical catastrophic break of the hot duct. Analysis shows the potential for significant and rapid air ingress into the reactor vessel in the case of a large break in the cooling system.

  7. VHTR Construction Ripple Effect using Inter-Industry Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, T. H.; Lee, K. Y.; Shin, Y. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As a part of a VHTR economic analysis, we have studied the VHTR construction cost and operation and maintenance cost. However, it is somewhat difficult to expect the exact cost due to insufficient reference data and experience. As a result, we propose quantitative analysis techniques for ripple effects such as the production inducement effect, added value inducement effect, and employment inducement effect for VHTR 600MWt x 4 module construction and operation ripple effect based on NOAK. This paper presents a new method for the ripple effect and preliminary ripple effect consequence. We proposed a ripple effect analysis method using a time series and inter-industry table. As a result, we can predict that a 600MWth x 4 module VHTR reactor construction will bring about a 43771 employment effect, 24160 billion KRW production effect, and 4472 billion added value effect for 22 years. It is necessary to use the sub-account values of an inter-industry table to obtain a more precise effect result. However, the methodology can be applied with minor modification to another reactor type.

  8. Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

  9. Thermal Shield and Reactor Structure Temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Collier, A.R.

    2001-07-31

    The purpose of this report is to present reactor structure and thermal shield temperature data taken during P-3 and P-5 cycles and compare them with design calculations in order to predict temperatures at higher power levels.

  10. WAVE PROPAGATION in the HOT DUCT of VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz; Jim C. P. Liou

    2013-07-01

    In VHTR, helium from the reactor vessel is conveyed to a power conversion unit through a hot duct. In a hypothesized Depressurized Conduction Cooldown event where a rupture of the hot duct occurs, pressure waves will be initiated and reverberate in the hot duct. A numerical model is developed to quantify the transients and the helium mass flux through the rupture for such events. The flow path of the helium forms a closed loop but only the hot duct is modeled in this study. The lower plum of the reactor vessel and the steam generator are treated as specified pressure and/or temperature boundary to the hot duct. The model is based on the conservation principles of mass, momentum and energy, and on the equations of state for helium. The numerical solution is based on the method of characteristics with specified time intervals with a predictor and corrector algorithm. The rupture sub-model gives reasonable results. Transients induced by ruptures with break area equaling 20%, 10%, and 5% of the duct cross-sectional area are described.

  11. Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)

    2007-07-15

    Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.

  12. VHTR-based Nuclear Hydrogen Plant Analysis for Hydrogen Production with SI, HyS, and HTSE Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper, analyses of material and heat balances on the SI, HyS, and HTSE processes coupled to a Very High Temperature gas-cooled Reactor (VHTR) were performed. The hydrogen production efficiency including the thermal to electric energy ratio demanded from each process is found and the normalized evaluation results obtained from three processes are compared to each other. The currently technological issues to maintain the long term continuous operation of each process will be discussed at the conference site. VHTR-based nuclear hydrogen plant analysis for hydrogen production with SI, HyS, and HTSE facilities has been carried out to determine the thermal efficiency. It is evident that the thermal to electrical energy ratio demanded from each hydrogen production process is an important parameter to select the adequate process for hydrogen production. To improve the hydrogen production efficiency in the SI process coupled to the VHTR without electrical power generation, the demand of electrical energy in the SI process should be minimized by eliminating an electrodialysis step to break through the azeotrope of the HI/I{sub 2}/H{sub 2}O ternary aqueous solution.

  13. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

    Energy Technology Data Exchange (ETDEWEB)

    William J. O’Donnell; Donald S. Griffin

    2007-05-07

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  14. Helium-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  15. Scoping Analyses on Tritium Permeation to VHTR Integarted Industrial Application Systems

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim

    2011-03-01

    Tritium permeation is a very important current issue in the very high temperature reactor (VHTR) because tritium is easily permeated through high temperature metallic surfaces. Tritium permeations in the VHTR-integrated systems were investigated in this study using the tritium permeation analysis code (TPAC) that was developed by Idaho National Laboratory (INL). The INL TPAC is a numerical tool that is based on the mass balance equations of tritium containing species and hydrogen (i.e. HT, H2, HTO, HTSO4, TI) coupled with a variety of tritium sources, sink, and permeation models. In the TPAC, ternary fission and thermal neutron caption reactions with 6Li, 7Li 10B, 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including high temperature electrolysis (HTSE) and sulfur-iodine processes.

  16. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Guzina, Bojan [Univ. of Minnesota, Minneapolis, MN (United States); Kunerth, Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-30

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  17. Investigation on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin

    2013-10-22

    Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the

  18. Development of the HELIOS/MASTER 2-Step Procedure for the Prismatic VHTR Physics Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, Jin Young; Noh, Jae Man

    2007-05-15

    A new physics analysis procedure has been developed for a prismatic very high temperature gas-cooled reactor based on a conventional two-step procedure for the PWR physics analysis. The HELIOS and MASTER codes were employed to generate the coarse group cross sections through a transport lattice calculation, and to perform the 3-dimensional core physics analysis by a nodal diffusion calculation, respectively. Physics analysis of the prismatic VHTRs involves particular modeling issues such as a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment and state parameters. Double heterogeneity effect was considered by using a recently developed reactivity equivalent physical transformation method. Neutron streaming effect was quantified through 3-dimensional Monte Carlo transport calculations by using the MCNP code. Strong core-reflector interaction could be handled by applying an equivalence theory to the generation of the reflector cross sections. The effects of a spectrum shift could be covered by optimizing the coarse energy group structure. A two-step analysis procedure was established for the prismatic VHTR physics analysis by combining all the methodologies described above. The applicability of our code system was tested against core benchmark problems. The results of these benchmark tests show that our code system is very accurate and practical for a prismatic VHTR physics analysis.

  19. Neutronic analysis stochastic distribution of fuel particles in Very High Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Ji, Wei

    The Very High Temperature Gas-Cooled Reactor (VHTR) is a promising candidate for Generation IV designs due to its inherent safety, efficiency, and its proliferation-resistant and waste minimizing fuel cycle. A number of these advantages stem from its unique fuel design, consisting of a stochastic mixture of tiny (0.78mm diameter) microspheres with multiple coatings. However, the microsphere fuel regions represent point absorbers for resonance energy neutrons, resulting in the "double heterogeneity" for particle fuel. Special care must be taken to analyze this fuel in order to predict the spatial and spectral dependence of the neutron population in a steady-state reactor configuration. The challenges are considerable and resist brute force computation: there are over 1010 microspheres in a typical reactor configuration, with no hope of identifying individual microspheres in this stochastic mixture. Moreover, when individual microspheres "deplete" (e.g., burn the fissile isotope U-235 or transmute the fertile isotope U-238 (eventually) to Pu-239), the stochastic time-dependent nature of the depletion compounds the difficulty posed by the stochastic spatial mixture of the fuel, resulting in a prohibitive computational challenge. The goal of this research is to develop a methodology to analyze particle fuel randomly distributed in the reactor, accounting for the kernel absorptions as well as the stochastic depletion of the fuel mixture. This Ph.D. dissertation will address these challenges by developing a methodology for analyzing particle fuel that will be accurate enough to properly model stochastic particle fuel in both static and time-dependent configurations and yet be efficient enough to be used for routine analyses. This effort includes creation of a new physical model, development of a simulation algorithm, and application to real reactor configurations.

  20. Failure Predictions for Graphite Reflector Bricks in the Very High Temperature Reactor with the Prismatic Core Design

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Gyanender, E-mail: sing0550@umn.edu [Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States); Fok, Alex [Minnesota Dental Research in Biomaterials and Biomechanics, School of Dentistry, University of Minnesota, 515, Delaware St. SE, Minneapolis, MN 55455 (United States); Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States); Mantell, Susan [Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States)

    2017-06-15

    Highlights: • Failure probability of VHTR reflector bricks predicted though crack modeling. • Criterion chosen for defining failure strongly affects the predictions. • Breaching of the CRC could be significantly delayed through crack arrest. • Capability to predict crack initiation and propagation demonstrated. - Abstract: Graphite is used in nuclear reactor cores as a neutron moderator, reflector and structural material. The dimensions and physical properties of graphite change when it is exposed to neutron irradiation. The non-uniform changes in the dimensions and physical properties lead to the build-up of stresses over the course of time in the core components. When the stresses reach the critical limit, i.e. the strength of the material, cracking occurs and ultimately the components fail. In this paper, an explicit crack modeling approach to predict the probability of failure of a VHTR prismatic reactor core reflector brick is presented. Firstly, a constitutive model for graphite is constructed and used to predict the stress distribution in the reflector brick under in-reactor conditions of high temperature and irradiation. Fracture simulations are performed as part of a Monte Carlo analysis to predict the probability of failure. Failure probability is determined based on two different criteria for defining failure time: A) crack initiation and B) crack extension to near control rod channel. A significant difference is found between the failure probabilities based on the two criteria. It is predicted that the reflector bricks will start cracking during the time range of 5–9 years, while breaching of the control rod channels will occur during the period of 11–16 years. The results show that, due to crack arrest, there is a significantly delay between crack initiation and breaching of the control rod channel.

  1. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  2. Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Hawari, Ayman [North Carolina State Univ., Raleigh, NC (United States). Dept. of Nuclear Engineering; Ougouag, Abderrafi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-07-08

    This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

  3. Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO{sub 2} emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO{sub 2} emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus{sup TM} Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition.

  4. Heat Exchanger Design Options and Tritium Transport Study for the VHTR System

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim

    2008-09-01

    This report presents the results of a study conducted to consider heat exchanger options and tritium transport in a very high temperature reactor (VHTR) system for the Next Generation Nuclear Plant Project. The heat exchanger options include types, arrangements, channel patterns in printed circuit heat exchangers (PCHE), coolant flow direction, and pipe configuration in shell-and-tube designs. Study considerations include: three types of heat exchanger designs (PCHE, shell-and-tube, and helical coil); single- and two-stage unit arrangements; counter-current and cross flow configurations; and straight pipes and U-tube designs in shell-and-tube type heat exchangers. Thermal designs and simple stress analyses were performed to estimate the heat exchanger options, and the Finite Element Method was applied for more detailed calculations, especially for PCHE designs. Results of the options study show that the PCHE design has the smallest volume and heat transfer area, resulting in the least tritium permeation and greatest cost savings. It is theoretically the most reliable mechanically, leading to a longer lifetime. The two-stage heat exchanger arrangement appears to be safer and more cost effective. The recommended separation temperature between first and second stages in a serial configuration is 800oC, at which the high temperature unit is about one-half the size of the total heat exchanger core volume. Based on simplified stress analyses, the high temperature unit will need to be replaced two or three times during the plant’s lifetime. Stress analysis results recommend the off-set channel pattern configuration for the PCHE because stress reduction was estimated at up to 50% in this configuration, resulting in a longer lifetime. The tritium transport study resulted in the development of a tritium behavior analysis code using the MATLAB Simulink code. In parallel, the THYTAN code, previously performed by Ohashi and Sherman (2007) on the Peach Bottom data, was revived

  5. Development Of An Experiment For Measuring Flow Phenomena Occurring In A Lower Plenum For VHTR CFD Assessment

    Energy Technology Data Exchange (ETDEWEB)

    D. M. McEligot; K.G. Condie; G. E. Mc Creery; H. M. Mc Ilroy

    2005-09-01

    The objective of the present report is to document the design of our first experiment to measure generic flow phenomena expected to occur in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In the process, fabrication sketches are provided for the use of CFD (computational fluid dynamics) analysts wishing to employ the data for assessment of their proposed codes. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. One aspect of the complex flow in a prismatic VHTR is being addressed: flow and thermal mixing in the lower plenum ("hot streaking" issue). Current prismatic VHTR concepts were examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses were applied to determine key non-dimensional parameters and their magnitudes over this operating range. The flow in the lower plenum can locally be considered to be a situation of multiple jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentum-dominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other.

  6. Preliminary Conceptual Design and Development of Core Technology of Very High Temperature Gas-Cooled Reactor Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jong Hwa; Kang, H. S.; Gil, C. S. and others

    2006-05-15

    For the nuclear hydrogen production system, the VHTR technology and the IS cycle technology are being developed. A comparative evaluation on the block type reactor and the pebble type reactor is performed to decide a proper nuclear hydrogen production reactor. 100MWt prismatic type reactor is tentatively decided and its safety characteristics are roughly investigated. Computation codes of nuclear design, thermo-fluid design, safety-performance analysis are developed and verified. Also, the development of a risk informed design technology is started. Experiments for metallic materials and graphites are carried out for the selection of materials of VHTR components. Diverse materials for process heat exchanger are studied in various corrosive environments. Pyrolytic carbon and SiC coating technology is developed and fuel manufacturing technology is basically established. Computer program is developed to evaluate the performance of coated particle fuels.

  7. Development of Tritium Permeation Analysis Code and Tritium Transport in a High Temperature Gas-Cooled Reactor Coupled with Hydrogen Production System

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2010-06-01

    Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems, including high temperature electrolysis and sulfur-iodine processes.

  8. Liquid Salt Heat Exchanger Technology for VHTR Based Applications

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark; Sridhara, Kumar; Allen, Todd; Peterson, Per

    2012-10-11

    The objective of this research is to evaluate performance of liquid salt fluids for use as a heat carrier for transferring high-temperature process heat from the very high-temperature reactor (VHTR) to chemical process plants. Currently, helium is being considered as the heat transfer fluid; however, the tube size requirements and the power associated with pumping helium may not be economical. Recent work on liquid salts has shown tremendous potential to transport high-temperature heat efficiently at low pressures over long distances. This project has two broad objectives: To investigate the compatibility of Incoloy 617 and coated and uncoated SiC ceramic composite with MgCl2-KCl molten salt to determine component lifetimes and aid in the design of heat exchangers and piping; and, To conduct the necessary research on the development of metallic and ceramic heat exchangers, which are needed for both the helium-to-salt side and salt-to-process side, with the goal of making these heat exchangers technologically viable. The research will consist of three separate tasks. The first task deals with material compatibility issues with liquid salt and the development of techniques for on-line measurement of corrosion products, which can be used to measure material loss in heat exchangers. Researchers will examine static corrosion of candidate materials in specific high-temperature heat transfer salt systems and develop an in situ electrochemical probe to measure metallic species concentrations dissolved in the liquid salt. The second task deals with the design of both the intermediate and process side heat exchanger systems. Researchers will optimize heat exchanger design and study issues related to corrosion, fabrication, and thermal stresses using commercial and in-house codes. The third task focuses integral testing of flowing liquid salts in a heat transfer/materials loop to determine potential issues of using the salts and to capture realistic behavior of the salts in a

  9. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, September 23, 1976--December 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This report presents the results of work performed from September 23, 1976 through December 31, 1976 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  10. Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving VHTR Efficiency and Testing Material Compatibility - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh

    2006-06-01

    Generation IV reactors will need to be intrinsically safe, having a proliferation-resistant fuel cycle and several advantages relative to existing light water reactor (LWR). They, however, must still overcome certain technical issues and the cost barrier before it can be built in the U.S. The establishment of a nuclear power cost goal of 3.3 cents/kWh is desirable in order to compete with fossil combined-cycle, gas turbine power generation. This goal requires approximately a 30 percent reduction in power cost for stateof-the-art nuclear plants. It has been demonstrated that this large cost differential can be overcome only by technology improvements that lead to a combination of better efficiency and more compatible reactor materials. The objectives of this research are (1) to develop a supercritical carbon dioxide Brayton cycle in the secondary power conversion side that can be applied to the Very-High-Temperature Gas-Cooled Reactor (VHTR), (2) to improve the plant net efficiency by using the carbon dioxide Brayton cycle, and (3) to test material compatibility at high temperatures and pressures. The reduced volumetric flow rate of carbon dioxide due to higher density compared to helium will reduce compression work, which eventually increase plant net efficiency.

  11. Characteristics of Modified 9Cr-1Mo Steel for Reactor Pressure Vessel of Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Ryu, W. S.; Han, Chang Hee; Yoon, J. H.; Chang, Jong Hwa

    2004-11-15

    Many researches and developments have been progressed for the construction of VHTR by 2020 in Korea. Modified 9Cr-1Mo steel has been receiving attention for the application to the reactor pressure vessel material of VHTR. We collected and analyzed the research data for modified 9Cr-1Mo steel in order to understand the characteristics of modified 9Cr-1Mo steel. The modified 9Cr-1Mo steel is a modified alloy system similar to conventional 9Cr-1Mo grade ferritic steel. Modifications include additions of vanadium, niobium, and nitrogen, as well as lower carbon content. In this report, we summarized the change of microstructure and mechanical properties after tempering, thermal aging, and irradiation. Modified 9Cr-1Mo steel has high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. But the irradiation embrittlement behavior of modified 9Cr-1Mo steel should be evaluated and the evaluation methodology also should be developed. At the same time, the characteristics of weldment which is the weak part in pressure vessel should be evaluated.

  12. Numerical and experimental investigation on labyrinth seal mechanism for bypass flow reduction in prismatic VHTR core

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su-Jong, E-mail: paper80@snu.ac.r [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Sang-Moon [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Tak, Nam-il; Kim, Min-Hwan [Korea Atomic Energy Research Institute, 150-1 Deokjin-Dong, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of); Kim, Kwang-Yong [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of)

    2013-09-15

    Highlights: • Bypass flow reduction method was developed by applying labyrinth seal mechanism. • Grooves on side walls of replaceable reflector block were made. • Design of the grooved wall of the reflector block was optimized by the RSA method. • The flow resistance of the bypass gap rose from 18.04 to 26.24 by the optimization. • The bypass ratios at the inlet and outlet were reduced by 36.19% and 14.66%, respectively. -- Abstract: Core bypass flow in block type very high temperature reactor (VHTR) occurs due to the inevitable gaps between the hexagonal core blocks for the block installation and refueling. Since the core bypass flow affects the reactor safety and efficiency, it should be minimized to enhance the core thermal margin. In this regard, the core bypass flow reduction method applying the labyrinth seal mechanism was developed and optimized by using the single-objective shape optimization method. Response surface approximation (RSA) method was adopted as the optimization method. Side wall of the replaceable reflector block was redesigned and response surface approximate model was adopted to optimize the shape of the reflector wall. Computational fluid dynamics (CFD) analyses were carried out not only to assess the limitation of existing method of bypass flow reduction, but also to optimize the design of a newly developed reduction method. The experiment with Seoul National University (SNU) multi-block experimental facility was performed to demonstrate the performance of the reduction method. It was found that the effect of the existing bypass flow reduction method by sealing the bypass gap exit was restricted nearby the lower region of the core. However, the flow resistance factor of the bypass gap increased from 18.04 to 26.24 by the optimized reduction method. The results of the performance test showed that the bypass flow distribution was reduced throughout the entire core regions. The bypass flow ratios at the inlet and the outlet were

  13. Initial VHTR accident scenario classification: models and data.

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Feldman, E. E.; Pointer, W. D.; Wei, T. Y. C.; Nuclear Engineering Division

    2005-09-30

    Nuclear systems codes are being prepared for use as computational tools for conducting performance/safety analyses of the Very High Temperature Reactor. The thermal-hydraulic codes are RELAP5/ATHENA for one-dimensional systems modeling and FLUENT and/or Star-CD for three-dimensional modeling. We describe a formal qualification framework, the development of Phenomena Identification and Ranking Tables (PIRTs), the initial filtering of the experiment databases, and a preliminary screening of these codes for use in the performance/safety analyses. In the second year of this project we focused on development of PIRTS. Two events that result in maximum fuel and vessel temperatures, the Pressurized Conduction Cooldown (PCC) event and the Depressurized Conduction Cooldown (DCC) event, were selected for PIRT generation. A third event that may result in significant thermal stresses, the Load Change event, is also selected for PIRT generation. Gas reactor design experience and engineering judgment were used to identify the important phenomena in the primary system for these events. Sensitivity calculations performed with the RELAP5 code were used as an aid to rank the phenomena in order of importance with respect to the approach of plant response to safety limits. The overall code qualification methodology was illustrated by focusing on the Reactor Cavity Cooling System (RCCS). The mixed convection mode of heat transfer and pressure drop is identified as an important phenomenon for Reactor Cavity Cooling System (RCCS) operation. Scaling studies showed that the mixed convection mode is likely to occur in the RCCS air duct during normal operation and during conduction cooldown events. The RELAP5/ATHENA code was found to not adequately treat the mixed convection regime. Readying the code will require adding models for the turbulent mixed convection regime while possibly performing new experiments for the laminar mixed convection regime. Candidate correlations for the turbulent

  14. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and

  15. CFD analysis of the VHTR prismatic core with variation of geometry parameters

    Energy Technology Data Exchange (ETDEWEB)

    Lira, Carlos A.B.O.; Paiva, Pedro P.D.S., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

    2017-11-01

    The Very High Temperature Reactor is a thermal, graphite moderated and helium cooled nuclear reactor. The purpose of this work is to study the behavior of the VHTR by means of parametric analysis, altering the energy generation profile in the fuel blocks and the influence of modifications in the geometry itself. The coolant flow through the coolant channels and by-pass channels were analyzed in a 1/12{sup th} section of a fuel block column. Geometry was used with by-pass channels of different dimensions, besides one that had only the cooling channels, without by-pass channel. It has been found that the existence of a by-pass flow induces an increase in the temperature gradient in the fuel block. Comparative studies were performed between the results obtained in simulations carried out with different profiles of thermal energy generation (uniform and sinusoidal) in the fuel channels. It was verified that when there is the same total thermal energy generation in the fuel block, the maximum temperature observed in each of the materials is smaller for the generation with sinusoidal profile. Computer simulations were performed using a geometry with a central channel with the same diameter as the others to verify the hypothesis that the existence of a temperature gradient in the fuel block, with the highest temperature at the center and the lowest temperature being at the periphery of this block, is due to the smaller dimension of the coolant channel located in the center of this block. The results obtained confirm the hypothesis. (author)

  16. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to

  17. Development and Verification of Tritium Analyses Code for a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim

    2009-09-01

    A tritium permeation analyses code (TPAC) has been developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in the VHTR systems including integrated hydrogen production systems. A MATLAB SIMULINK software package was used for development of the code. The TPAC is based on the mass balance equations of tritium-containing species and a various form of hydrogen (i.e., HT, H2, HTO, HTSO4, and TI) coupled with a variety of tritium source, sink, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of HT and H2 through pipes, vessels, and heat exchangers were importantly considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including both high-temperature electrolysis and sulfur-iodine process. The TPAC has unlimited flexibility for the system configurations, and provides easy drag-and-drops for making models by adopting a graphical user interface. Verification of the code has been performed by comparisons with the analytical solutions and the experimental data based on the Peach Bottom reactor design. The preliminary results calculated with a former tritium analyses code, THYTAN which was developed in Japan and adopted by Japan Atomic Energy Agency were also compared with the TPAC solutions. This report contains descriptions of the basic tritium pathways, theory, simple user guide, verifications, sensitivity studies, sample cases, and code tutorials. Tritium behaviors in a very high temperature reactor/high temperature steam electrolysis system have been analyzed by the TPAC based on the reference indirect parallel configuration proposed by Oh et al. (2007). This analysis showed that only 0.4% of tritium released from the core is transferred to the product hydrogen

  18. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Quarterly progress report, April 1, 1977--June 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The objectives of this program are to evaluate candidate alloys of Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle helium Turbine (DCHT) applications, in terms of the effects of simulated reactor primary coolant (impure helium), high temperatures, and long time exposures on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in the report includes completion of alloy selection for the screening tests. The alloys selected for potential VHTR Nuclear Process Heat (NPH) applications and for potential DCHT applications are listed. The present status on the simulated reactor helium loop design and construction and the design and construction progress on the testing and analysis facilities and equipment are discussed.

  19. Measured gas and particle temperatures in VTT's entrained flow reactor

    DEFF Research Database (Denmark)

    Clausen, Sønnik; Sørensen, L.H.

    2006-01-01

    Particle and gas temperature measurements were carried out in experiments on VTTs entrained flow reactor with 5% and 10% oxygen using Fourier transform infrared emission spectroscopy (FTIR). Particle temperature measurements were performed on polish coal,bark, wood, straw particles, and bark...... and wood particles treated with additive. A two-color technique with subtraction of the background light was used to estimate particle temperatures during experiments. A transmission-emission technique was used tomeasure the gas temperature in the reactor tube. Gas temperature measurements were in good...

  20. HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

    1960-06-14

    A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

  1. Parametric Study on the Tensile Properties of Ni-Based Alloy for a VHTR

    Science.gov (United States)

    Kim, Dong-Jin; Jung, Su Jin; Mun, Byung Hak; Kim, Sung Woo; Lim, Yun Soo

    2015-01-01

    A very high-temperature reactor (VHTR) has been studied among generation IV nuclear power plants owing to its many advantages such as high-electric efficiency and massive hydrogen production. The material used for the heat exchanger should sustain structural integrity for its life even though the material is exposed to a harsh environment at 1223 K (950 °C) in an impure helium coolant. Therefore, an enhancement of the material performance at high temperature gives a margin in determining the operating temperature and life time. This work is an effort to find an optimum combination of alloying elements and processing parameters to improve the material performance. The tensile property and microstructure for nickel-based alloys fabricated in a laboratory were evaluated as a function of the heat treatment, cold working, and grain boundary strengthener using a tension test at 1223 K (950 °C), scanning electron microscopy, and transmission electron microscopy. Elongation to rupture was increased by additional heat treatment and cold working, followed by additional heat treatment in the temperature range from 1293 K to 1383 K (1020 °C to 1110 °C) implying that the intergranular carbide contributes to grain boundary strengthening. The temperature at which the grain boundary is improved by carbide decoration was higher for a cold-worked specimen, which was described by the difference in carbide stability and carbide formation kinetics between no cold-worked and cold-worked specimens. Zr and Hf played a scavenging effect of harmful elements causing an increase in ductility.

  2. Containment for the low temperature district nuclear-heating reactor

    Energy Technology Data Exchange (ETDEWEB)

    He Shuyan (Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China)); Dong Duo (Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China))

    1993-04-01

    An integral arrangement is adopted for the Low Temperature District Nuclear-Heating Reactor. The primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with the reactor core. The primary coolant flows in natural circulation through the reactor core and the primary heat exchangers. The primary coolant pipes penetrating the wall of the reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of the pressure boundary of the primary coolant. Therefore a small sized metallic containment closed to the wall of the reactor vessel can be used for the reactor. Design principles and functions of the containment are the same as for the containment of a PWR. But the adoption of a small sized containment brings about some benefits such as a short period of manufacturing, relatively low cost, and ease for sealing. A loss of primary coolant accident would not be happening during a rupture accident of the primary coolant pressure boundary inside the containment owing to its intrinsic safety. (orig.).

  3. Tritium Formation and Mitigation in High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  4. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  5. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  6. The Status of the US High-Temperature Gas Reactors

    Directory of Open Access Journals (Sweden)

    Andrew C. Kadak

    2016-03-01

    Full Text Available In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor (HTGR by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current status is that high temperature reactors have been relegated to completing research programs on advanced fuels, graphite and materials with no plans to build a demonstration plant as required by the US Congress in 2005. There are many reasons behind this diminution of HTGR development, including but not limited to insufficient government funding requirements for research, unrealistically high temperature requirements for the reactor, the delay in the need for a “hydrogen” economy, competition from light water small modular light water reactors, little utility interest in new technologies, very low natural gas prices in the US, and a challenging licensing process in the US for non-water reactors.

  7. Gas-cooled reactors: the importance of their development

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U/sub 3/O/sub 8/ before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production.

  8. Alcohol synthesis in a high-temperature slurry reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, G.W.; Marquez, M.A.; McCutchen, M.S. [North Carolina State Univ., Raleigh, NC (United States)

    1995-12-31

    The overall objective of this contract is to develop improved process and catalyst technology for producing higher alcohols from synthesis gas or its derivatives. Recent research has been focused on developing a slurry reactor that can operate at temperatures up to about 400{degrees}C and on evaluating the so-called {open_quotes}high pressure{close_quotes} methanol synthesis catalyst using this reactor. A laboratory stirred autoclave reactor has been developed that is capable of operating at temperatures up to 400{degrees}C and pressures of at least 170 atm. The overhead system on the reactor is designed so that the temperature of the gas leaving the system can be closely controlled. An external liquid-level detector is installed on the gas/liquid separator and a pump is used to return condensed slurry liquid from the separator to the reactor. In order to ensure that gas/liquid mass transfer does not influence the observed reaction rate, it was necessary to feed the synthesis gas below the level of the agitator. The performance of a commercial {open_quotes}high pressure {close_quotes} methanol synthesis catalyst, the so-called {open_quotes}zinc chromite{close_quotes} catalyst, has been characterized over a range of temperature from 275 to 400{degrees}C, a range of pressure from 70 to 170 atm., a range of H{sub 2}/CO ratios from 0.5 to 2.0 and a range of space velocities from 2500 to 10,000 sL/kg.(catalyst),hr. Towards the lower end of the temperature range, methanol was the only significant product.

  9. Tritium Formation and Mitigation in High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-10-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  10. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  11. Technologies for Upgrading Light Water Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  12. System and method for temperature control in an oxygen transport membrane based reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.

    2017-02-21

    A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  13. Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-09-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding

  14. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  15. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    OpenAIRE

    Cisneros, Anselmo Tomas

    2013-01-01

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids - flibe (33%7Li2F-67%BeF) - from molten salt reactors. This combination of fuel and coolant enables FHRs to operate i...

  16. Methane reforming in a temperature-controlled DBD reactor

    Science.gov (United States)

    Levko, Dmitry; Raja, Laxminarayan

    2015-09-01

    Methane and carbon dioxide are among the main products of human activity. Therefore, they are considered among greenhouse gases, which may cause the global warming. On the other hand, methane is widely used in everyday life as an energy source and in industry for the synthesis of different chemicals. In order to utilize greenhouse gases or to generate chemicals from methane, one needs first to dissociate it. Then, this gas converts into desired products such as methanol, gasoline, syn-gas etc. Nowadays, there are several methods for CH4 conversion. Steam reforming, partial oxidation, thermal and non-thermal plasmas are among them. During the last decades, the use of non-thermal plasma for methane reforming attracts more and more attention. This is caused by the possibility to control the process of methane conversion as well as the gas component content at the reactor outlet. In addition, the use of non-thermal plasma facilitates the control of reactor start up. The goal of the present work is the deep understanding of the plasma chemical processes accompanying the methane-air conversion in a temperature-controlled DBD reactor. To do this, we have developed the kinetic mechanism of CH4/N2/O2 conversion for the gas temperature range 300-800 K and applied it to the global model.

  17. Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae K., E-mail: taekyu8@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Arcilesi, David J., E-mail: arcilesi.1@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Kim, In H., E-mail: ihkim0730@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Sun, Xiaodong, E-mail: sun.200@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Christensen, Richard N., E-mail: rchristensen@uidaho.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Oh, Chang H. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Kim, Eung S., E-mail: kes7741@snu.ac.kr [Idaho National Laboratory, Idaho Falls, ID 83402 (United States)

    2016-04-15

    Highlights: • Uncertainty quantification and benchmark study are performed to validate an ANSYS FLUENT computer model for a depressurization process in a high-temperature gas-cooled reactor. • An ANSYS FLUENT computer model of a 1/8th scaled-down geometry of a VHTR hot exit plenum is presented, which is similar to the experimental test facility that has been constructed at The Ohio State University. • Using the computer model of the scaled-down geometry, the effects of the depressurization process and flow oscillations on the subsequent density-driven stratified flow phenomenology are examined computationally. • The effects of the scaled-down hot exit plenum internal structure temperature on the density-driven stratified flow phenomenology are investigated numerically. - Abstract: An air-ingress accident is considered to be one of the design basis accidents of a very high-temperature gas-cooled reactor (VHTR). The air-ingress accident is initiated, in its worst-case scenario, by a complete break of the hot duct in what is referred to as a double-ended guillotine break. This leads to an initial loss of the primary helium coolant via depressurization. Following the depressurization process, the air–helium mixture in the reactor cavity could enter the reactor core via the hot duct and hot exit plenum. In the event that air ingresses into the reactor vessel, the high-temperature graphite structures in the reactor core and hot plenum will chemically react with the air, which could lead to damage of in-core graphite structures and fuel, release of carbon monoxide and carbon dioxide, core heat up, failure of the structural integrity of the system, and eventually the release of radionuclides to the environment. Studies in the available literature focus on the phenomena of the air ingress accident that occur after the termination of the depressurization, such as density-driven stratified flow, molecular diffusion, and natural circulation. However, a recent study

  18. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  19. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)

    1998-12-31

    In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

  20. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Michael Swanson; Daniel Laudal

    2008-03-31

    . Higher-reactivity (low-rank) coals appear to perform better in a transport reactor than the less reactive bituminous coals. Factors that affect TRDU product gas quality appear to be coal type, temperature, and air/coal ratios. Testing with a higher-ash, high-moisture, low-rank coal from the Red Hills Mine of the Mississippi Lignite Mining Company has recently been completed. Testing with the lignite coal generated a fuel gas with acceptable heating value and a high carbon conversion, although some drying of the high-moisture lignite was required before coal-feeding problems were resolved. No ash deposition or bed material agglomeration issues were encountered with this fuel. In order to better understand the coal devolatilization and cracking chemistry occurring in the riser of the transport reactor, gas and solid sampling directly from the riser and the filter outlet has been accomplished. This was done using a baseline Powder River Basin subbituminous coal from the Peabody Energy North Antelope Rochelle Mine near Gillette, Wyoming.

  1. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  2. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  3. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program: Topical report I, selection of candidate alloys. Volume 3. Selection of surface coating/substrate systems for screening creep and structural stability studies

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-20

    Considering the high temperature, low O/sub 2/, high C environment of operation in the Very High Temperature Reactor (VHTR) Systems, the utilization of coatings is envisaged to hold potential for extending component lifetimes through the formation of stable and continuous oxide films with enhanced resistance to C diffusion. A survey of the current state of technology for high temperature coatings has been performed. The usefulness of these coatings on the Mo, Ni, and Fe base alloys is discussed. Specifically, no coating substitute was identified for TZM other than the well known W-3 (pack silicide) and Al/sub 2/O/sub 3/ forming coatings were recommended for the Fe and Ni base structural materials. Recommendations as to coating types and processng have been made based on the predicted VHTR component size, shape, base metal and operational environment. Four tests designed to evaluate the effects of selected combinations of coatings and substrate matrices are recommended for consideration.

  4. Optimization of Reactor Temperature and Catalyst Weight for Plastic Cracking to Fuels Using Response Surface Methodology

    OpenAIRE

    Istadi Istadi; S Suherman; Luqman Buchori

    2011-01-01

    The present study deals with effect of reactor temperature and catalyst weight on performance of plastic waste cracking to fuels over modified catalyst waste as well as their optimization. From optimization study, the most operating parameters affected the performance of the catalytic cracking process is reactor temperature followed by catalyst weight. Increasing the reactor temperature improves significantly the cracking performance due to the increasing catalyst activity. The optimal operat...

  5. Helium turbine power generation in high temperature gas reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Yasuo [Tokyo Inst. of Tech. (Japan)

    1995-03-01

    This paper presents studies on the helium turbine power generator and important components in the indirect cycle of high temperature helium cooled reactor with multi-purpose use of exhaust thermal energy from the turbine. The features of this paper are, firstly the reliable estimation of adiabatic efficiencies of turbine and compressor, secondly the introduction of heat transfer enhancement by use of the surface radiative heat flux from the thin metal plates installed in the hot helium and between the heat transfer coil rows of IHX and RHX, thirdly the use of turbine exhaust heat to produce fresh water from seawater for domestic, agricultural and marine fields, forthly a proposal of plutonium oxide fuel without a slight possibility of diversion of plutonium for nuclear weapon production and finally the investigation of GT-HTGR of large output such as 500 MWe. The study of performance of GT-HTGR reduces the result that for the reactor of 450 MWt the optimum thermal efficiency is about 43% when the turbine expansion ratio is 3.9 for the turbine efficiency of 0.92 and compressor efficiency of 0.88 and the helium temperature at the compressor inlet is 45degC. The produced amount of fresh water is about 8640 ton/day. It is made clear that about 90% of the reactor thermal output is totally used for the electric power generation in the turbine and for the multi-puposed utilization of the heat from the turbine exhaust gas and compressed helium cooling seawater. The GT-Large HTGR is realized by the separation of the pressure and temperature boundaries of the pressure vessel, the increase of burning density of the fuel by 1.4 times, the extention of the nuclear core diameter and length by 1.2 times, respectively, and the enhancement of the heat flux along the nuclear fuel compact surface by 1.5 times by providing riblets with the peak in the flow direction. (J.P.N.).

  6. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  7. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  8. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-25

    Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

  9. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, January 1, 1979-March 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-19

    This report presents the results of work performed from January 1, 1979 through March 31, 1979 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. Work covered in this report includes the activities associated with the creep-rupture testing of the test materials for the purpose of verifying the stresses selected for the screening creep test program, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment.

  10. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, October 1, 1978--December 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1979-03-09

    Results of work performed from October 1, 1978 through December 31, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program is presented. Objectives are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys, and selection of materials for future test facilities and more extensive qualification programs. The activities associated with the characterization of the materials for the screening test program, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are included. The status of the data management system is presented.

  11. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-07

    The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  12. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1979-June 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-25

    The results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  13. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  14. Analysis of supercritical CO{sub 2} cycle control strategies and dynamic response for Generation IV Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J. (Nuclear Engineering Division)

    2011-04-12

    The analysis of specific control strategies and dynamic behavior of the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle has been extended to the two reactor types selected for continued development under the Generation IV Nuclear Energy Systems Initiative; namely, the Very High Temperature Reactor (VHTR) and the Sodium-Cooled Fast Reactor (SFR). Direct application of the standard S-CO{sub 2} recompression cycle to the VHTR was found to be challenging because of the mismatch in the temperature drop of the He gaseous reactor coolant through the He-to-CO{sub 2} reactor heat exchanger (RHX) versus the temperature rise of the CO{sub 2} through the RHX. The reference VHTR features a large temperature drop of 450 C between the assumed core outlet and inlet temperatures of 850 and 400 C, respectively. This large temperature difference is an essential feature of the VHTR enabling a lower He flow rate reducing the required core velocities and pressure drop. In contrast, the standard recompression S-CO{sub 2} cycle wants to operate with a temperature rise through the RHX of about 150 C reflecting the temperature drop as the CO{sub 2} expands from 20 MPa to 7.4 MPa in the turbine and the fact that the cycle is highly recuperated such that the CO{sub 2} entering the RHX is effectively preheated. Because of this mismatch, direct application of the standard recompression cycle results in a relatively poor cycle efficiency of 44.9%. However, two approaches have been identified by which the S-CO{sub 2} cycle can be successfully adapted to the VHTR and the benefits of the S-CO{sub 2} cycle, especially a significant gain in cycle efficiency, can be realized. The first approach involves the use of three separate cascaded S-CO{sub 2} cycles. Each S-CO{sub 2} cycle is coupled to the VHTR through its own He-to-CO{sub 2} RHX in which the He temperature is reduced by 150 C. The three respective cycles have efficiencies of 54, 50, and 44%, respectively, resulting in a net cycle

  15. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  16. Systems and methods for enhancing isolation of high-temperature reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per F.

    2017-09-26

    A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

  17. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  18. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  19. Development of Barrier Layers for the Protection of Candidate Alloys in the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Levi, Carlos G. [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States); Jones, J. Wayne [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States); Pollock, Tresa M. [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States); Was, Gary S. [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States)

    2015-01-22

    The objective of this project was to develop concepts for barrier layers that enable leading candi- date Ni alloys to meet the longer term operating temperature and durability requirements of the VHTR. The concepts were based on alpha alumina as a primary surface barrier, underlay by one or more chemically distinct alloy layers that would promote and sustain the formation of the pro- tective scale. The surface layers must possess stable microstructures that provide resistance to oxidation, de-carburization and/or carburization, as well as durability against relevant forms of thermo-mechanical cycling. The system must also have a self-healing ability to allow endurance for long exposure times at temperatures up to 1000°C.

  20. CEA and AREVA R and D on V/HTR fuel fabrication with the CAPRI experimental manufacturing line

    Energy Technology Data Exchange (ETDEWEB)

    Charollais, Francois; Fonquernie, Sophie; Perrais, Christophe [DEN/CAD/DEC/SPUA, CEA, Centre de Cadarache, F-13108 Saint Paul lez Durance (France); Perez, Marc [DRT/GRE/DTEN/S3ME, CEA Grenoble, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France); Cellier, Francois [FRAMATOME ANP, 10, rue Juliette Recamier, F-69456 Lyon cedex (France); Vitali, Marie-Pierre [CERCA Romans, FRAMATOME ANP, 10, rue Juliette Recamier, F-69456 Lyon cedex (France)

    2006-07-01

    In the framework of the French V/HTR fuel development and qualification program, the Commissariat a l'Energie Atomique (CEA) and AREVA through its program called ANTARES (Areva New Technology for Advanced Reactor Energy Supply) conduct R and D projects covering the mastering of UO{sub 2} coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture: - The different stages of UO{sub 2} kernel fabrication GSP Sol-Gel process have been reviewed, reproduced and improved; - The experimental conditions for the chemical vapour deposition (CVD) of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out; - Former CERCA compacting process has been reviewed and updated. In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line). The major objectives of the CAPRI line are: - to recover and validate past knowledge; - to permit the optimisation of reference fabrication processes for kernels and coatings and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating); - to test alternative compact process options; - to fabricate and characterize fuel required for irradiation and qualification purpose; - to specify needs for the fabrication of representative V/HTR TRISO fuel meeting industrial standards. This paper presents the progress status of the R and D conducted on V/HTR fuel particle and compact manufacture by mid 2005. (authors)

  1. Neutron analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  2. Spectral emissivity of candidate alloys for very high temperature reactors in high temperature air environment

    Energy Technology Data Exchange (ETDEWEB)

    Cao, G., E-mail: gcao@wisc.edu; Weber, S.J.; Martin, S.O.; Sridharan, K.; Anderson, M.H.; Allen, T.R.

    2013-10-15

    Emissivity measurements for candidate alloys for very high temperature reactors were carried out in a custom-built experimental facility, capable of both efficient and reliable measurements of spectral emissivities of multiple samples at high temperatures. The alloys studied include 304 and 316 austenitic stainless steels, Alloy 617, and SA508 ferritic steel. The oxidation of alloys plays an important role in dictating emissivity values. The higher chromium content of 304 and 316 austenitic stainless steels, and Alloy 617 results in an oxide layer only of sub-micron thickness even at 700 °C and consequently the emissivity of these alloys remains low. In contrast, the low alloy SA508 ferritic steel which contains no chromium develops a thicker oxide layer, and consequently exhibits higher emissivity values.

  3. Spectral emissivity of candidate alloys for very high temperature reactors in high temperature air environment

    Science.gov (United States)

    Cao, G.; Weber, S. J.; Martin, S. O.; Sridharan, K.; Anderson, M. H.; Allen, T. R.

    2013-10-01

    Emissivity measurements for candidate alloys for very high temperature reactors were carried out in a custom-built experimental facility, capable of both efficient and reliable measurements of spectral emissivities of multiple samples at high temperatures. The alloys studied include 304 and 316 austenitic stainless steels, Alloy 617, and SA508 ferritic steel. The oxidation of alloys plays an important role in dictating emissivity values. The higher chromium content of 304 and 316 austenitic stainless steels, and Alloy 617 results in an oxide layer only of sub-micron thickness even at 700 °C and consequently the emissivity of these alloys remains low. In contrast, the low alloy SA508 ferritic steel which contains no chromium develops a thicker oxide layer, and consequently exhibits higher emissivity values.

  4. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  5. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    OpenAIRE

    Shutang Zhu; Ying Tang; Kun Xiao; Zuoyi Zhang

    2008-01-01

    This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR) technology with the supercritical (SC) steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR) reveal that the development of SCWR power plants still needs further research and develop...

  6. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    Directory of Open Access Journals (Sweden)

    Fic Adam

    2015-03-01

    Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  7. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    Science.gov (United States)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  8. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  9. Temperature uniformity mapping in a high pressure high temperature reactor using a temperature sensitive indicator

    NARCIS (Netherlands)

    Grauwet, T.; Plancken, van der I.; Vervoort, L.; Matser, A.M.; Hendrickx, M.; Loey, van A.

    2011-01-01

    Recently, the first prototype ovomucoid-based pressure–temperature–time indicator (pTTI) for high pressure high temperature (HPHT) processing was described. However, for temperature uniformity mapping of high pressure (HP) vessels under HPHT sterilization conditions, this prototype needs to be

  10. Production test PTA-002, increased graphite temperature limit -- B, C and D Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Russell, A.

    1965-12-17

    The fundamental objective of the graphite temperature limit is to prevent excessive oxidation of the graphite moderator blocks with carbon dioxide and water vapor in the reactor atmosphere. Laboratory tests have shown that 10% uniform oxidation of graphite results in a loss in strength of approximately 50%. Production Test IP-725 was conducted at F Reactor for a period of six months at graphite temperatures approximately 50 and 100 C higher than the present graphite temperature limit of 650 C. The results from the F Reactor test suggest that an increase in the graphite temperature limit from 650 C to 700 C is technically feasible from the standpoint of oxidation of the graphite moderator with CO{sub 2}. Any significant additional increase was shown to lead to excessively high oxidation rates and is therefore not considered feasible. The objective of this test, therefore, is to extend the higher temperature investigations to B, C, and D Reactors. For the duration of this test, the graphite temperature limit will be increased from 650 C and 700 C, corresponding to an increase in the graphite stringer temperature limit from 735 C to 790 C. The test is expected to last for approximately six months but may be terminated early on any or all the reactors.

  11. Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL; Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Holcomb, David Eugene [ORNL

    2011-01-01

    Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700 C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.

  12. Measurement of cryogenic moderator temperature effects in a small heterogeneous thermal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hoovler, G.S.; Ball, R.M.; Lewis, R.H.

    1994-12-31

    Past papers have described a critical experiment (CX) built at Sandia National Laboratories to investigate the neutronic behavior of the particle-bed reactor (PBK). Among the experiments previously reported were tests to measure the reactivity effect of uniform temperature variations between 20 and 80{degree}C. This paper describes additional experiments designed to examine the effects of cryogenic moderator temperatures on core reactivity and neutron spectrum. The general importance of temperature effects to the design of the PBR have been previously discussed. A unique feature of the PBR is that the moderator may be at cryogenic temperatures during reactor startup. Because temperature effects in small, heterogeneous thermal reactors can be significant and because we found no integral measurements with cryogenic moderators in such systems, an experiment with a cryogenic moderator was designed and performed in the CX as an extension to the isothermal measurements previously reported.

  13. High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

    Directory of Open Access Journals (Sweden)

    Dudek Michał

    2016-01-01

    Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  14. Development of ceramic membrane reactors for high temperature gas cleanup. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, D.L.; Abraham, I.C.; Blum, Y.; Gottschlich, D.E.; Hirschon, A.; Way, J.D.; Collins, J.

    1993-06-01

    The objective of this project was to develop high temperature, high pressure catalytic ceramic membrane reactors and to demonstrate the feasibility of using these membrane reactors to control gaseous contaminants (hydrogen sulfide and ammonia) in integrated gasification combined cycle (IGCC) systems. Our strategy was to first develop catalysts and membranes suitable for the IGCC application and then combine these two components as a complete membrane reactor system. We also developed a computer model of the membrane reactor and used it, along with experimental data, to perform an economic analysis of the IGCC application. Our results have demonstrated the concept of using a membrane reactor to remove trace contaminants from an IGCC process. Experiments showed that NH{sub 3} decomposition efficiencies of 95% can be achieved. Our economic evaluation predicts ammonia decomposition costs of less than 1% of the total cost of electricity; improved membranes would give even higher conversions and lower costs.

  15. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

  16. Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Dunfu

    2015-07-01

    The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

  17. Analysis of Air Temperatures around Reactor Vessel in EU-APR

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Seok; Lee, Keun Sung; Hwang, Do Hyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    EU-APR, modified and improved from its original design of APR1400, has been developed to comply with European Utility Requirements (EUR) and nuclear design requirements of the European countries. In EU-APR, the removable concrete shielding blocks is newly adopted to reduce the radioactive dose rate in order to allow personnel access to the containment building during power operation. During plants startup, hot shutdown and power operation, the reactor cavity HVAC system maintains temperature and humidity for the in-core instrument (ICI) chase, ex-core detector spaces, reactor cavity and the hot and cold leg penetration opening. CFD analysis has been performed in order to check if the air temperature in the reactor cavity and concrete in EU-APR are exceeded the temperature limits. Six calculations for EU-APR have been carried out with different opening areas of the relief damper plate and air flow rates sucked out through six ventilation pipes. For the temperature distribution in the reactor cavity, it is found that there are the regions above a temperature of 120℉(48.9℃). However, these regions are relatively small and are observed very near the reactor vessel insulation. Therefore, these high temperature regions do not directly influence on concrete. Furthermore, the heat emission used in the current calculations already includes 20% margin including the conservative margins used for the metal reflective heat losses and the margins used to estimate Gamma and neutron heat into the primary wall. Totally, 10% margin is included.

  18. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

  19. High temperature ceramic membrane reactors for coal liquid upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, T.T.

    1992-01-01

    In this project we will study a novel process concept, i.e., the use of ceramic membrane reactors in upgrading of coal model compounds and coal derived liquids. In general terms, the USC research team is responsible for constructing and operating the membrane reactor apparatus and for testing various inorganic membranes for the upgrading of coal derived asphaltenes and coal model compounds. The USC effort will involve the principal investigator of this project and two graduate research assistants. The ALCOA team is responsible for the preparation of the inorganic membranes, for construction and testing of the ceramic membrane modules, and for measurement of their transport properties. The ALCOA research effort will involve Dr. Paul K. T. Liu, who is the project manager of the ALCOA research team, an engineer and a technician. UNOCAL's contribution will be limited to overall technical assistance in catalyst preparation and the operation of the laboratory upgrading membrane reactor and for analytical back-up and expertise in oil analysis and materials characterization. UNOCAL is a no-cost contractor but will be involved in all aspects of the project, as deemed appropriate.

  20. Effect of post-digestion temperature on serial CSTR biogas reactor performance

    DEFF Research Database (Denmark)

    Boe, Kanokwan; Karakashev, Dimitar Borisov; Trably, Eric

    2009-01-01

    of 5.3 days. Three post-digestion temperatures (55 degrees C, 37 degrees C and 15 degrees C) were compared in terms of biogas production, process stability, microbial community and methanogenic activity, The results showed that the post-digesters operated at 55 degrees C, 37 degrees C and 15 degrees C...... gave extra biogas production of 11.7%, 8.4% and 1.2%, respectively. The post-digester operated at 55 degrees C had the highest biogas production and was the most stable in terms of low VFA concentrations. The specific methanogenic activity tests revealed that the main reactor and the post-digester......The effect of post-digestion temperature on a lab-scale serial continuous-flow stirred tank reactor (CSTR) system performance was investigated. The system consisted of a main reactor operated at 55 degrees C with hydraulic retention time (HRT) of 15 days followed by post-digestion reactors with HRT...

  1. Effect of temperature on selenium removal from wastewater by UASB reactors.

    Science.gov (United States)

    Dessì, Paolo; Jain, Rohan; Singh, Satyendra; Seder-Colomina, Marina; van Hullebusch, Eric D; Rene, Eldon R; Ahammad, Shaikh Ziauddin; Carucci, Alessandra; Lens, Piet N L

    2016-05-01

    The effect of temperature on selenium (Se) removal by upflow anaerobic sludge blanket (UASB) reactors treating selenate and nitrate containing wastewater was investigated by comparing the performance of a thermophilic (55 °C) versus a mesophilic (30 °C) UASB reactor. When only selenate (50 μM) was fed to the UASB reactors (pH 7.3; hydraulic retention time 8 h) with excess electron donor (lactate at 1.38 mM corresponding to an organic loading rate of 0.5 g COD L(-1) d(-1)), the thermophilic UASB reactor achieved a higher total Se removal efficiency (94.4 ± 2.4%) than the mesophilic UASB reactor (82.0 ± 3.8%). When 5000 μM nitrate was further added to the influent, total Se removal was again better under thermophilic (70.1 ± 6.6%) when compared to mesophilic (43.6 ± 8.8%) conditions. The higher total effluent Se concentration in the mesophilic UASB reactor was due to the higher concentrations of biogenic elemental Se nanoparticles (BioSeNPs). The shape of the BioSeNPs observed in both UASB reactors was different: nanospheres and nanorods, respectively, in the mesophilic and thermophilic UASB reactors. Microbial community analysis showed the presence of selenate respirers as well as denitrifying microorganisms. Copyright © 2016 Elsevier Ltd. All rights reserved.

  2. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  3. High Temperature Ultrasonic Transducers for In-Service Inspection of Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Griffin, Jeffrey W.; Posakony, Gerald J.; Harris, Robert V.; Baldwin, David L.; Jones, Anthony M.; Bond, Leonard J.

    2011-12-31

    In-service inspection of liquid metal (sodium) fast reactors requires the use of ultrasonic transducers capable of operating at high temperatures (>200°C), high gamma radiation fields, and the chemically reactive liquid sodium environment. In the early- to mid-1970s, the U.S. Atomic Energy Commission supported development of high-temperature, submersible single-element transducers, used for scanning and under-sodium imaging in the Fast Flux Test Facility and the Clinch River Breeder Reactor. Current work is building on this technology to develop the next generation of high-temperature linear ultrasonic transducer arrays for under-sodium viewing and in-service inspections.

  4. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  5. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis

    Science.gov (United States)

    Adam, Zachary R.

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 105-106 years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  6. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

    Science.gov (United States)

    Adam, Zachary R

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  7. One centimeter spatial resolution temperature measurements in a nuclear reactor using Rayleigh scatter in optical fiber

    Science.gov (United States)

    Sang, A. K.; Gifford, D. K.; Dickerson, B. D.; Fielder, B. F.; Froggatt, M. E.

    2007-07-01

    We present the use of swept wavelength interferometry for distributed fiber-optic temperature measurements in a Nuclear Reactor. The sensors consisted of 2 m segments of commercially available, single mode optical fibers. The interrogation technique is based on measuring the spectral shift of the intrinsic Rayleigh backscatter signal along the optical fiber and converting the spectral shift to temperature.

  8. Optimization of Reactor Temperature and Catalyst Weight for Plastic Cracking to Fuels Using Response Surface Methodology

    Directory of Open Access Journals (Sweden)

    Istadi Istadi

    2011-01-01

    Full Text Available The present study deals with effect of reactor temperature and catalyst weight on performance of plastic waste cracking to fuels over modified catalyst waste as well as their optimization. From optimization study, the most operating parameters affected the performance of the catalytic cracking process is reactor temperature followed by catalyst weight. Increasing the reactor temperature improves significantly the cracking performance due to the increasing catalyst activity. The optimal operating conditions of reactor temperature about 550 oC and catalyst weight about 1.25 gram were produced with respect to maximum liquid fuel product yield of 29.67 %. The liquid fuel product consists of gasoline range hydrocarbons (C4-C13 with favorable heating value (44,768 kJ/kg. ©2010 BCREC UNDIP. All rights reserved(Received: 10th July 2010, Revised: 18th September 2010, Accepted: 19th September 2010[How to Cite: I. Istadi, S. Suherman, L. Buchori. (2010. Optimization of Reactor Temperature and Catalyst Weight for Plastic Cracking to Fuels Using Response Surface Methodology. Bulletin of Chemical Reaction Engineering and Catalysis, 5(2: 103-111. doi:10.9767/bcrec.5.2.797.103-111][DOI: http://dx.doi.org/10.9767/bcrec.5.2.797.103-111 || or local:  http://ejournal.undip.ac.id/index.php/bcrec/article/view/797

  9. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  10. An investigation of moving bed biofilm reactor nitrification during long-term exposure to cold temperatures.

    Science.gov (United States)

    Hoang, Valerie; Delatolla, Robert; Laflamme, Edith; Gadbois, Alain

    2014-01-01

    Biological treatment is the most common and economical means of ammonia removal in wastewater; however, nitrification rates can become completely impeded at cold temperatures. Attached growth processes and, specifically, moving bed biofilm reactors (MBBRs) have shown promise with respect to low-temperature nitrification. In this study, two laboratory MBBRs were used to investigate MBBR nitrification rates at 20, 5, and 1 degree C. Furthermore, the solids detached by the MBBR reactors were investigated and Arrhenius temperature correction models used to predict nitrification rates after long-term low-temperature exposure was evaluated. The nitrification rate at 5 degrees C was 66 +/- 3.9% and 64 +/- 3.7% compared to the rate measured at 20 degrees C for reactors 1 and 2, respectively. The nitrification rates at 1 degree C over a 4-month exposure period compared to the rate at 20 degrees C were 18.7 +/- 5.5% and 15.7 +/- 4.7% for the two reactors. The quantity of solids detached from the MBBR biocarriers was low and the mass of biofilm per carrier did not vary significantly at 20 degrees C compared to that after long-term exposure at 1 degree C. Lastly, a temperature correction model based on exposure time to cold temperatures showed a strong correlation to the calculated ammonia removal rates relative to 20 degrees C following a gradual acclimatization period to cold temperatures.

  11. New steady-state microbial community compositions and process performances in biogas reactors induced by temperature disturbances.

    Science.gov (United States)

    Luo, Gang; De Francisci, Davide; Kougias, Panagiotis G; Laura, Treu; Zhu, Xinyu; Angelidaki, Irini

    2015-01-01

    The microbial community in a biogas reactor greatly influences the process performance. However, only the effects of deterministic factors (such as temperature and hydraulic retention time (HRT)) on the microbial community and performance have been investigated in biogas reactors. Little is known about the manner in which stochastic factors (for example, stochastic birth, death, colonization, and extinction) and disturbance affect the stable-state microbial community and reactor performances. In the present study, three replicate biogas reactors treating cattle manure were run to examine the role of stochastic factors and disturbance in shaping microbial communities. In the triplicate biogas reactors with the same inoculum and operational conditions, similar process performances and microbial community profiles were observed under steady-state conditions. This indicated that stochastic factors had a minor role in shaping the profile of the microbial community composition and activity in biogas reactors. On the contrary, temperature disturbance was found to play an important role in the microbial community composition as well as process performance for biogas reactors. Although three different temperature disturbances were applied to each biogas reactor, the increased methane yields (around 10% higher) and decreased volatile fatty acids (VFAs) concentrations at steady state were found in all three reactors after the temperature disturbances. After the temperature disturbance, the biogas reactors were brought back to the original operational conditions; however, new steady-state microbial community profiles were observed in all the biogas reactors. The present study demonstrated that temperature disturbance, but not stochastic factors, played an important role in shaping the profile of the microbial community composition and activity in biogas reactors. New steady-state microbial community profiles and reactor performances were observed in all the biogas reactors

  12. Hybrid sulfur cycle operation for high-temperature gas-cooled reactors

    Science.gov (United States)

    Gorensek, Maximilian B

    2015-02-17

    A hybrid sulfur (HyS) cycle process for the production of hydrogen is provided. The process uses a proton exchange membrane (PEM) SO.sub.2-depolarized electrolyzer (SDE) for the low-temperature, electrochemical reaction step and a bayonet reactor for the high-temperature decomposition step The process can be operated at lower temperature and pressure ranges while still providing an overall energy efficient cycle process.

  13. Analysis of Fluid Flow and Heat Transfer Model for the Pebble Bed High Temperature Gas Cooled Reactor

    OpenAIRE

    S. Yamoah; E.H.K. Akaho; Nana G.A. Ayensu; M. Asamoah

    2012-01-01

    The pebble bed type high temperature gas cooled nuclear reactor is a promising option for next generation reactor technology and has the potential to provide high efficiency and cost effective electricity generation. The reactor unit heat transfer poses a challenge due to the complexity associated with the thermalflow design. Therefore to reliably simulate the flow and heat transport of the pebble bed modular reactor necessitates a heat transfer model that deals with radiation as well as ther...

  14. Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

    Directory of Open Access Journals (Sweden)

    Hsoung-Wei Chou

    2015-01-01

    Full Text Available The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to be the severest accident for a boiling water reactor pressure vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic reactor pressure vessel has sufficient structural integrity until doubling of the present end-of-license operation.

  15. Dynamic behavior of the HTR-10 reactor: Dual temperature feedback model

    Directory of Open Access Journals (Sweden)

    Hosseini Seyed Ali

    2015-01-01

    Full Text Available The current work aims at presenting a simple model for PBM-type reactors' dynamic behavior analysis. The proposed model is based on point kinetics equations coupled with feedbacks from fuel and moderator temperatures. The temperature reactivity coefficients were obtained through MCNP code and via available experimental data. Parameters such as heat capacity and heat conductivity were carefully analyzed and the final system of equations was numerically solved. The obtained results, while in partial agreement with previously proposed models, suggest lower sensitivity to step reactivity insertion as compared to other reactor designs and inherent safety of the design.

  16. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  17. Feasibility study of high temperature reactor utilization in Czech Republic after 2025

    Energy Technology Data Exchange (ETDEWEB)

    Losa, Evžen, E-mail: evzen.losa@fjfi.cvut.cz [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Heřmanský, Bedřich; Kobylka, Dušan; Rataj, Jan; Sklenka, Ľubomír [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Souček, Václav; Kohout, Petr [AZIN CZ, s.r.o., Hanusova 3, 140 00 Praha 4 (Czech Republic)

    2014-05-01

    High temperature reactors (HTRs) were examined as an option to intended future broadening of the nuclear energy production in Czech Republic. The known qualities as the inherent safety, high thermal utilization and non-electrical applications have been assessed in years 2009–2011 during the survey funded by Czech Ministry of Industry and Trade. The survey of high temperature reactors with spherical fuel was initiated by reason of mature state of the art of this technology type in South Africa and in China, where in both countries pilot plants were planned. Unfortunately, the global financial crisis caused the decision of stopping the governmental support in South African programme was made. In China, however, the development still continues. Czech Republic has almost 60 years nuclear research history and the knowledge of operation of gas cooled and heavy water moderated reactor has been gained in the past. Nevertheless, the design of light water reactors was more developed in former Soviet Union, which provided Czech scientists by initial knowledge base; hence the research has been reoriented to this technology. But, the demands on future nuclear reactors application are still growing and the same or even higher living standard of next generations have to be taken into consideration. Therefore the systems, which can produce more energy and less waste, are getting into foreground of interest of Czech decision makers. The high temperature reactor technology seems to be the successful representative of the GEN IV reactor types, which will be operated commercially in the near future. The broad spectrum of utilization enables this system to be an option after 2030, when the electricity demand is planned to be covered from about 50% by nuclear in our country.

  18. Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J. [Iowa State Univ., Ames, IA (United States); Bowler, John R. [Iowa State Univ., Ames, IA (United States)

    2017-08-30

    The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-service inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO3-xPbTiO3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.

  19. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    2004-07-29

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  20. Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D. A.; Hobbins, R. R.; Lowry, P.; Gougar, H.

    2013-11-01

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  1. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  2. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  3. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  4. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  5. Low temperature oxidation of hydrocarbons using an electrochemical reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide

    at different reaction temperatures. The study of the effect of the infiltration of different electroactive materials on the electrode behavior has been carried on by the use of electrochemical impedance spectroscopy (EIS). Both the methods have been employed to understand the relationship between the catalytic...... the catalytic activity at open circuit voltage and the effect of polarization on propene oxidation rate at low temperature. The future development of this technology will see the infiltration of an active catalyst towards propene oxidation together with a NOx storage compound for the simultaneous oxidation...... high propene conversion at open circuit voltage together high rate enhancement ratio and faradaic efficiency values at low temperatures (300-350 °C). Although some stability problems affected the performance of multiple infiltrated Ce0.9Gd0.1O1.95 on LSM/CGO backbone, the strong activation of LSM upon...

  6. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  7. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  8. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    Energy Technology Data Exchange (ETDEWEB)

    Lee Nelson

    2011-09-01

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  9. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    NARCIS (Netherlands)

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable

  10. Power cycle assessment of nuclear high temperature gas-cooled reactors

    OpenAIRE

    Herranz, L.E.; Linares, J.I.; Moratilla, B.Y.

    2009-01-01

    Power cycle assessment of nuclear high temperature gas-cooled reactors correspondance: Corresponding author. Tel.: +34 91 346 62 36; fax: +34 91 346 62 33. (Herranz, L.E.) (Herranz, L.E.) Unit of Nuclear Safety Research (CIEMAT) Avda. Complutense--> , 22 - 28040 Madrid - Spain--> - (Herranz, L.E.) Unit of Nuclear Safety Research (CIEMAT) Avda. Complutense--> , 22 - 28040 Madrid - Spain--...

  11. Effect of reactor temperature on direct growth of carbon nanomaterials on stainless steel

    Science.gov (United States)

    Edzatty, A. N.; Syazwan, S. M.; Norzilah, A. H.; Jamaludin, S. B.

    2016-07-01

    Currently, carbon nanomaterials (CNMs) are widely used for various applications due to their extraordinary electrical, thermal and mechanical properties. In this work, CNMs were directly grown on the stainless steel (SS316) via chemical vapor deposition (CVD). Acetone was used as a carbon source and argon was used as carrier gas, to transport the acetone vapor into the reactor when the reaction occurred. Different reactor temperature such as 700, 750, 800, 850 and 900 °C were used to study their effect on CNMs growth. The growth time and argon flow rate were fixed at 30 minutes and 200 ml/min, respectively. Characterization of the morphology of the SS316 surface after CNMs growth using Scanning Electron Microscopy (SEM) showed that the diameter of grown-CNMs increased with the reactor temperature. Energy Dispersive X-ray (EDX) was used to analyze the chemical composition of the SS316 before and after CNMs growth, where the results showed that reduction of catalyst elements such as iron (Fe) and nickel (Ni) at high temperature (700 - 900 °C). Atomic Force Microscopy (AFM) analysis showed that the nano-sized hills were in the range from 21 to 80 nm. The best reactor temperature to produce CNMs was at 800 °C.

  12. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

    The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  13. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Grodzki Marcin

    2017-12-01

    Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  14. Recycling of polyethene and polypropene in a novel bench-scale rotating cone reactor by high temperature pyrolysis.

    NARCIS (Netherlands)

    Westerhout, R.W.J.; Westerhout, R.W.J.; Waanders, J.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

    1998-01-01

    The high-temperature pyrolysis of polyethene (PE), polypropene (PP), and mixtures of these polymers was studied in a novel bench-scale rotating cone reactor (RCR). Experiments showed that the effect of the sand or reactor temperature on the product spectrum obtained is large compared to the effect

  15. Current hurdles to the success of high temperature membrane reactors

    NARCIS (Netherlands)

    Saracco, G.; Versteeg, Geert; van Swaaij, Willibrordus Petrus Maria

    1994-01-01

    High-temperature catalytic processs performed using inorganic membranes have been in recent years a fast growing area of research, which seems to have not yet reached its peak. Chemical engineers, catalysts and materials scientists have addressed this topic from different viewpoint in a common

  16. THERMAL COUPLE FOR MEASURING TEMPERATURE IN A REACTOR

    Science.gov (United States)

    Kanne, W.

    1959-11-24

    A thermocouple device for measuring the temperature of a flowing fluid in a conduit within which is positioned a metallic rod is presented. A thermocouple junction is secured to the rod centrally, and thermal insulating support disks having a diameter greater than the rod are secured to the end portions of the rod and adapted to fit transversely in the conduit.

  17. Commercialization of modular high temperature gas-cooled reactors in the world

    Energy Technology Data Exchange (ETDEWEB)

    Hayakawa, Hitoshi; Okamoto, Futoshi; Ohhashi, Kazutaka [Fuji Electric Co. Ltd., Tokyo (Japan)

    2001-07-01

    The construction programs of the commercial high temperature gas-cooled reactors have been activated extraordinarily all over the world. This paper gives an overview of the three major programs, the South African PBMR project (US utility Exelon announced recently their plan to import PBMRs), the international GT-MHR project (by US DOE, General Atomics, MINATOM of Russian Federation, FRAMATOME ANP, Fuji Electric) and Chinese HTR-PM project. And the reasons why the utilities selected small modular HTGRs as next generation reactors, the superior characteristics of the small modular HTGRs for power generation plant and prospects of them are summarized and discussed. (author)

  18. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  19. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs.

  20. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  1. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  2. A high temperature drop-tube and packed-bed solar reactor for continuous biomass gasification

    Science.gov (United States)

    Bellouard, Quentin; Abanades, Stéphane; Rodat, Sylvain; Dupassieux, Nathalie

    2017-06-01

    Biomass gasification is an attractive process to produce high-value syngas. Utilization of concentrated solar energy as the heat source for driving reactions increases the energy conversion efficiency, saves biomass resource, and eliminates the needs for gas cleaning and separation. A high-temperature tubular solar reactor combining drop tube and packed bed concepts was used for continuous solar-driven gasification of biomass. This 1 kW reactor was experimentally tested with biomass feeding under real solar irradiation conditions at the focus of a 2 m-diameter parabolic solar concentrator. Experiments were conducted at temperatures ranging from 1000°C to 1400°C using wood composed of a mix of pine and spruce (bark included) as biomass feedstock. The aim of this study was to demonstrate the feasibility of syngas production in this reactor concept and to prove the reliability of continuous biomass gasification processing using solar energy. The study first consisted of a parametric study of the gasification conditions to obtain an optimal gas yield. The influence of temperature and oxidizing agent (H2O or CO2) on the product gas composition was investigated. The study then focused on solar gasification during continuous biomass particle injection for demonstrating the feasibility of a continuous process. Regarding the energy conversion efficiency of the lab scale reactor, energy upgrade factor of 1.21 and solar-to-fuel thermochemical efficiency up to 28% were achieved using wood heated up to 1400°C.

  3. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Swanson

    2005-08-30

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was

  4. Research Activities on Development of Piping Design Methodology of High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Nam-Su [Seoul National Univ. of Science and Technology, Seoul(Korea, Republic of); Won, Min-Gu [Sungkyukwan Univ., Suwon (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering and Construction Co. Inc., Gimcheon (Korea, Republic of); Lee, Hyeog-Yeon; Kim, Yoo-Gon [Korea Atomic Energy Research Institute, Daejeon(Korea, Republic of)

    2016-10-15

    A SFR is operated at high temperature and low pressure compared with commercial pressurized water reactor (PWR), and such an operating condition leads to time-dependent damages such as creep rupture, excessive creep deformation, creep-fatigue interaction and creep crack growth. Thus, high temperature design and structural integrity assessment methodology should be developed considering such failure mechanisms. In terms of design of mechanical components of SFR, ASME B and PV Code, Sec. III, Div. 5 and RCC-MRx provide high temperature design and assessment procedures for nuclear structural components operated at high temperature, and a Leak-Before-Break (LBB) assessment procedure for high temperature piping is also provided in RCC-MRx, A16. Three web-based evaluation programs based on the current high temperature codes were developed for structural components of high temperature reactors. Moreover, for the detailed LBB analyses of high temperature piping, new engineering methods for predicting creep C*-integral and creep COD rate based either on GE/EPRI or on reference stress concepts were proposed. Finally, the numerical methods based on Garofalo's model and RCC-MRx have been developed, and they have been implemented into ABAQUS. The predictions based on both models were compared with the experimental results, and it has been revealed that the predictions from Garafalo's model gave somewhat successful results to describe the deformation behavior of Gr. 91 at elevated temperatures.

  5. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  6. SuPer-Homogenization (SPH) Corrected Cross Section Generation for High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hiruta, Hikaru [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-03-01

    The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy of full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.

  7. Potential for use of high-temperature superconductors in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hull, J.R.

    1991-01-01

    The present rate of development of high-temperature superconductors (HTSs) is sufficiently rapid that there may be opportunities for their use in contemporary fusion devices such as the International Thermonuclear Experimental Reactor (ITER). The most likely 1application is for delivering power to the superconducting magnets, especially in substituting for the current leads between the temperatures of 4 K and 77K. A second possible application of HTSs is as a liquid-nitrogen-cooled power bus, connecting the power supplies to the magnets, thus reducing the ohmic heating losses over these relatively long cables. A third potential application of HTSs is as an inner high-field winding of the toroidal field coils that would operate at {approx}20 K. While the use of higher temperature magnets offers significant advantages to the reactor system, it is unlikely that tested conductors of this type will be available within the ITER time frame. 23 refs., 2 figs.

  8. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  9. Nitrogen Removal by Anammox Biofilm Column Reactor at Moderately Low Temperature

    Directory of Open Access Journals (Sweden)

    Tuty Emilia Agustina

    2017-10-01

    Full Text Available The anaerobic ammonium oxidation (anammox as a new biological approach for nitrogen removal has been considered to be more cost-effective compared with the combination of nitrification and denitrification process. However, the anammox bioreactors are mostly explored at high temperature (>300C in which temperature controlling system is fully required. This research was intended to develop and to apply anammox process for high nitrogen concentration removal at ambient temperature used for treating wastewater in tropical countries. An up-flow biofilm column reactor, which the upper part constructed with a porous polyester non-woven fabric material as a carrier to attach the anammox bacteria was operated without heating system. A maximum nitrogen removal rate (NRR of 1.05 kg-N m3 d-1 was reached in the operation days of 178 with a Total Nitrogen (TN removal efficiency of 74%. This showed the biofilm column anammox reactor was successfully applied to moderate high nitrogen removal from synthetic wastewater at moderately low temperature. Keywords: Anammox, biofilm column reactor, ambient temperature, nitrogen removal

  10. Evaluation of temperature distribution sensing method for fast reactor using optical fiber

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, Atsushi; Nakazawa, Masaharu [Tokyo Univ. (Japan); Ichige, Satoshi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-12-01

    Optical fiber sensors (OFSs) have many advantages like flexible configuration, intrinsic immunity for electromagnetic fields, and so on. For these reasons, it is very useful to apply OFSs to fast reactor plants for remote inspection and surveillance. However, under irradiation, because of radiation-induced transmission loss of optical fibers, OFSs have radiation-induced errors. Therefore, to apply OFSs to nuclear facilities, we have to estimate and correct the errors. In this report, Raman Distributed Temperature Sensor (RDTS; one of the OFSs) has been installed at the primary coolant loop of the experimental fast reactor JOYO of JNC (Japan Nuclear Cycle Development Institute). Two correction techniques (correction technique with two thermocouples and correction technique with loop arrangement) for radiation-induced errors have been developed and demonstrated. Because of the radiation-induced loss, measured temperature distributions had radiation-induced errors. However, during the continuous measurements with the total dose of more than 8 x 10{sup 3}[C/kg](3 x 10{sup 7}[R]), the radiation induced errors showed a saturation tendency. In case of the temperature distributions with fluorine doped fiber, with one of the correction techniques, the temperature errors reduced to 1{approx}2degC and the feasibility of the loss correction techniques was demonstrated. For these results, it can be said that RDTS can be applied as a temperature distribution monitor in harsh radiation environments like fast reactor plants. (author)

  11. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trond Bjornard; John Hockert

    2011-08-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work

  12. Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO.

  13. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  14. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

  15. Localization of the hot spots in a pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Leisheng; Lee, Wooram; Lee, Jaeyoung [Handong Global University, Pohang (Korea, Republic of)

    2016-05-15

    The pebble bed reactor (PBR) is a candidate reactor type for the very high temperature reactor (VHTR), which is one of the Generation-IV reactor types. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. The conclusions are made and may contribute to a better design of a PBR core and a closer inspection of the local hot spots to avoid destruction of pebbles from happening. Thermal field of a PBR core is investigated in this study. Specifically, experiments on measuring the pebbles' surface temperature are performed. It is found that the upper pebble has an overall higher temperature profile than the other pebbles and the stagnation zone under does not increase its surface's temperature. In addition, the temperature profile of the side pebble shows a concave form and it keeps decreasing from the contact point to the vertex in the lower pebble. Lastly, the maximum temperature difference among these points is 5.83 deg. C. These findings above are validated by CFX simulations under two different turbulence models (k-e, SST) and two contact areas (diameter of 6mm and 3.5mm). By contrasting the temperature variation trends of all simulation cases, it is concluded that SST turbulence model with 20% intensity shows a better agreement with the experiment result, nevertheless, slightly deviation is also found in terms of total temperature difference and the peak appears in position 17-19 in experiments.

  16. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  17. Effect of temperature on two-phase anaerobic reactors treating slaughterhouse wastewater

    Directory of Open Access Journals (Sweden)

    Simone Beux

    2007-11-01

    Full Text Available The effectiveness of the anaerobic treatment of effluent from a swine and bovine slaughterhouse was assessed in two sets of two-phase anaerobic digesters, operated with or without temperature control. Set A, consisting of an acidogenic reactor with recirculation and an upflow biological filter as the methanogenic phase, was operated at room temperature, while set B, consisting of an acidogenic reactor without recirculation and an upflow biological filter as the methanogenic phase, was maintained at 32°C. The methanogenic reactors showed COD (Chemical Demand of Oxygen removal above 60% for HRT (Hydraulic Retention Time values of 20, 15, 10, 8, 6, 4, and 2 days. When the HRT value in those reactors was changed to 1 day, the COD percentage removal decreased to 50%. The temperature variations did not have harmful effects on the performance of reactors in set A.Avaliou-se a eficiência do tratamento anaeróbio de efluente de matadouro de suínos e bovinos em dois conjuntos de biodigestores anaeróbios de duas fases, operados com e sem controle de temperatura. O conjunto A, formado por um reator acidogênico com recirculação e um filtro biológico de fluxo ascendente, foi operado a temperatura ambiente e o conjunto B, formado por um reator de fluxo ascendente e um filtro biológico de fluxo ascendente, foi mantido a 32°C. Os reatores metanogênicos apresentaram remoção de DQO acima de 60 % para os TRHs de 20, 15, 10, oito, seis, quatro e dois dias. Quando o TRH destes reatores foi mudado para um dia observou-se uma queda da porcentagem de remoção de DQO para 50 %. As variações de temperatura parecem não ter prejudicado o desempenho dos reatores do conjunto A.

  18. Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Zuoyi [Inst. of Nuclear Energy Technology Tsinghua Univ., Beijing, BJ (China); Scherer, Winfried

    1997-12-31

    This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H{sub 2}O density increase rate in the reactor core is smaller than 0.3 kg/(m{sup 3}s). The liquid water vaporization in the steam generator and H{sub 2}O transport from the steam generator to the reactor core reduces the impulse of the H{sub 2}O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

  19. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2016-09-15

    Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

  20. Effects of Surface Roughness, Oxidation, and Temperature on the Emissivity of Reactor Pressure Vessel Alloys

    Energy Technology Data Exchange (ETDEWEB)

    King, J. L. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Jo, H. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Tirawat, R. [National Renewable Energy Laboratory, Concentrating Solar Power Group, Golden, Colorado; Blomstrand, K. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Sridharan, K. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin

    2017-08-31

    Thermal radiation will be an important mode of heat transfer in future high-temperature reactors and in off-normal high-temperature scenarios in present reactors. In this work, spectral directional emissivities of two reactor pressure vessel (RPV) candidate materials were measured at room temperature after exposure to high-temperature air. In the case of SA508 steel, significant increases in emissivity were observed due to oxidation. In the case of Grade 91 steel, only very small increases were observed under the tested conditions. Effects of roughness were also investigated. To study the effects of roughening, unexposed samples of SA508 and Grade 91 steel were roughened via one of either grinding or shot-peening before being measured. Significant increases were observed only in samples having roughness exceeding the roughness expected of RPV surfaces. While the emissivity increases for SA508 from oxidation were indeed significant, the measured emissivity coefficients were below that of values commonly used in heat transfer models. Based on the observed experimental data, recommendations for emissivity inputs for heat transfer simulations are provided.

  1. Design of Helium Brayton Cycle for Small Modular High Temperature Gas cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Yoon Han; Lee, Je Kyoung; Lee, Jeong Ik [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2012-05-15

    The small modular reactor (SMR) is gaining a lot of interest recently. Not only it can achieve better passive safety, but also it can be potentially utilized for the diverse applications to respond to the increasing global energy demands. As a part of the SMR development effort, SM-HTGR (Small Modular-High Temperature Gas-cooled Reactor), a 20MWth reactor is under development by the Korean Atomic Energy Research Institute (KAERI) for the complete passive safety, desalination and industrial process heat application. The Helium Brayton cycle is considered as a promising candidate for the SM-HTGR power conversion. The advantages of Helium Brayton cycles are: 1) helium is an inert gas that does not interact with structure material. 2) helium is chemically stable that helium Brayton cycle can be utilized under the high temperature circumstance. 3) higher thermal efficiency is achievable under higher outlet temperature range. Moreover, high temperature advantage can be utilized (reinforced) by diverting part of the heat for industrial process heat. This paper will discuss the progress on the helium power conversion cycle operating condition optimization by studying the sensitivity of the maximum pressure, pressure ratio and the component cooling on the total cycle efficiency

  2. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Avincola, V.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission Joint Research Centre - Institute for Transuranium Elements (JRC-ITU) (Germany)

    2012-11-01

    The High Temperature Reactor (HTR) is an advanced reactor concept with particular safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRISO (tri-isotropic) coating designed to provide high fission product retention. Passive safety features of the HTR include a low power density in the core compared to other reactor designs; this ensures sufficient heat transport in a loss of coolant accident scenario. The temperature during such events would not exceed 1600 C, remaining well below the melting point of the fuel. An experimental assessment of the fuel behaviour under severe accident conditions is necessary to confirm the fission product retention of TRISO coated particles and to validate relevant computer codes. Though helium is used as coolant for the HTR system, additional corrosion effects come into play in case of an in-leakage affecting the primary circuit. The experimental scope of the present work focuses on two key aspects associated with the HTR fuel safety. Fission product retention at high temperatures (up to {proportional_to}1800 C) is analyzed with the so-called cold finger apparatus (KueFA: Kuehlfinger-Apparatur), while the performance of HTR fuel elements in case of air/steam ingress accidents is assessed with a high temperature corrosion apparatus (KORA: Korrosions-Apparatur). (orig.)

  3. Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

    Science.gov (United States)

    Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.; Crowhurst, Jonathan C.; Weisz, David G.; Zaug, Joseph M.; Dai, Zurong; Radousky, Harry B.; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L.; Cappelli, Mark A.; Rose, Timothy P.

    2017-09-01

    We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

  4. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  5. A study of temperature sensor location based on fractal analysis for cascade control schemes in tubular reactors

    DEFF Research Database (Denmark)

    Eduardo Ramirez-Castelan, Carlos; Moguel-Castañeda, Jazael; Puebla, Hector

    2016-01-01

    Temperature sensor location for cascade control schemes in tubular reactors is still an open research problem. Several studies have pointed out that most temperature sensitive zones along the length of the reactor are suitable to this end. In this work, we have studied the problem of sensor...... location in a cascade control configuration using fractal analysis of time series obtained by random forcing of the jacket rector. A benchmark dispersion axial model displaying different temperature profiles is used to illustrate our findings....

  6. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  7. High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.; Swearingen, Gary L.; Pugh, Ray

    2011-10-31

    Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simple prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC

  8. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  9. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  10. Design aspects of the Chinese modular high-temperature gas-cooled reactor HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Zuoyi [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Wu Zongxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Sun Yuliang [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Li Fu [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)]. E-mail: lifu@tsinghua.edu.cn

    2006-03-15

    The modular high-temperature gas-cooled reactor (MHTGR) has distinct advantages in terms of inherent safety, economics potential, high efficiency, potential usage for hydrogen production, etc. The Chinese design of the MHTGR, named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM), based on the technology and experience of the HTR-10, is currently in the conceptual phase. The HTR-PM demonstration plant is planned to be finished by 2012. The main philosophy of the HTR-PM project can be pinned down as: (1) safety, (2) standardization, (3) economy, and (4) proven technology. The work in the categories of marketing, organization, project and technology is done in predefined order. The biggest challenge for the HTR-PM is to ensure its economical viability while maintaining its inherent safety. A design of a 450 MWth annular pebble bed core connected with steam turbine is aimed for and presented in this paper.

  11. The modular high-temperature gas-cooled reactor: A cost/risk competitive nuclear option

    Energy Technology Data Exchange (ETDEWEB)

    Gotschall, H.L. (Gas-Cooled Reactor Associates, San Diego, CA (United States))

    1994-01-01

    The business risks of nuclear plant ownership are identified as a constraint on the expanded use of nuclear power. Such risks stem from the exacting demands placed on owner/operator organizations of current plants to demonstrate ongoing compliance with safety regulations and the resulting high costs for operation and maintenance. This paper describes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) design, competitive economics, and approach to reducing the business risks of nuclear plant ownership.

  12. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    Energy Technology Data Exchange (ETDEWEB)

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

  13. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  14. Sensitivity studies of modular high-temperature gas-cooled reactor postulated accidents

    Energy Technology Data Exchange (ETDEWEB)

    Ball, Syd [Nuclear Science and Technology Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6010 (United States)]. E-mail: sjb@ornl.gov

    2006-03-15

    The results of various accident scenario simulations for the two major modular high temperature gas-cooled reactor (HTGR) variants (prismatic and pebble bed cores) are presented. Sensitivity studies can help to quantify the uncertainty ranges of the predicted outcomes for variations in some of the more crucial system parameters, as well as for occurrences of equipment and/or operator failures or errors. In addition, sensitivity studies can guide further efforts in improving the design and determining where more (or less) R and D is appropriate. Both of the modular HTGR designs studied - the 400-MW(t) pebble bed modular reactor (PBMR, pebble) and the 600-MW(t) gas-turbine modular helium reactor (GT-MHR, prismatic) - show excellent accident prevention and mitigation capabilities because of their inherent passive safety features. The large thermal margins between operating and 'potential damage' temperatures, along with the typically very slow accident response times (approximate days to reach peak temperatures), tend to reduce concerns about uncertainties in the simulation models, the initiating events, and the equipment and operator responses.

  15. A temperature swing adsorptive reactor for the enhancement of catalytic dehydrogenation reaction

    Energy Technology Data Exchange (ETDEWEB)

    Oh, M.; Jang, E.J. [LG Engineering Co., Ltd. (Korea, Republic of); Moon, J.K. [Korea Atomic Energy Research Institute (Korea, Republic of)

    1998-02-01

    Based on temperature swing concept arising from forced periodic variations in feed conditions, a theoretical study of a novel catalytic reactor system is carried out, in which adsorption and reaction occur simultaneously. In order to evaluate the performance of a temperature swing adsorptive reaction(TSAR) process for various operating conditions, the rigorous dynamic model of the reactor is developed and dynamic simulation is performed. In the application of the concept suggested in this paper, the dehydrogenation process of methyl-cyclohexane(MCH) to produce toluene is employed and the conversion and energy consumption is calculated. The results are compared with the case of isothermal equilibrium reaction. For the whole range of investigation, it is examined TSAR process is superior to an isothermal equilibrium reactor in the sense of a conversion and energy saving. In the consideration of the results of this study, TSAR process can provide a means for the effective utilization of a waste heat source for reaction heat and for feed gas preheat; periodic catalyst regeneration; the generation of optimal temperature for the improvement of a conversion; and reaction enhancement for theoretically limited reactions.

  16. Kinetic study on the effect of temperature on biogas production using a lab scale batch reactor.

    Science.gov (United States)

    Deepanraj, B; Sivasubramanian, V; Jayaraj, S

    2015-11-01

    In the present study, biogas production from food waste through anaerobic digestion was carried out in a 2l laboratory-scale batch reactor operating at different temperatures with a hydraulic retention time of 30 days. The reactors were operated with a solid concentration of 7.5% of total solids and pH 7. The food wastes used in this experiment were subjected to characterization studies before and after digestion. Modified Gompertz model and Logistic model were used for kinetic study of biogas production. The kinetic parameters, biogas yield potential of the substrate (B), the maximum biogas production rate (Rb) and the duration of lag phase (λ), coefficient of determination (R(2)) and root mean square error (RMSE) were estimated in each case. The effect of temperature on biogas production was evaluated experimentally and compared with the results of kinetic study. The results demonstrated that the reactor with operating temperature of 50°C achieved maximum cumulative biogas production of 7556ml with better biodegradation efficiency. Copyright © 2015 Elsevier Inc. All rights reserved.

  17. Low-temperature anaerobic digestion of swine manure in a plug-flow reactor.

    Science.gov (United States)

    Massé, Daniel I; Gilbert, Yan; Saady, N M C; Liu, Charle

    2013-01-01

    A low-temperature (25 degrees C) anaerobic eight-compartment (PF01 to PF08) cascade reactor simulating a plug-flow reactor (PFR) treating pig manure was monitored for a year. The bioreactor was fed at an average loading rate of 2.4 +/- 0.2 g of total chemical oxygen demand (TCOD) per litre of reactor per day for a theoretical hydraulic retention time (HRT) of 67 +/- 7 d. An average of 79% of TCOD was removed from pig manure (converted into biogas and in sediments), whereas specific methane yields ranging from 397 to 482 NL CH4 kg(-1) VS (148.6 to 171.4 NL CH4 kg(-1) TCOD) were obtained. After 150 d, fluctuating performances of the process were observed, associated with solids accumulation in the upstream compartments, preventing the complete anaerobic digestion of swine manure in the compartments PF01 to PF04. Low-temperature anaerobic PFR represents an interesting alternative for the treatment of pig manure and recovery of green energy. Further investigations regarding a modified design, with better accumulating solids management, are needed to optimize the performance of this low-temperature PFR treating pig manure.

  18. Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-05-15

    A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized.

  19. Evaluation of the high temperature electrolysis of steam to produce hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Park, Wonseok; Chang, Jonghwa; Park, Jongkuen [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yuseong-gu, Daejeon (Korea)

    2007-07-15

    A very high temperature gas-cooled reactor (VHTR) can be effectively used for hydrogen production through several CO{sub 2}-free alternative technologies, such as the Sulfur-Iodine (SI) cycle, the high temperature electrolysis of steam (HTES), and others. In our current study, the electrochemical thermodynamic properties and the overall thermal efficiency of the VHTR-assisted hydrogen production system by using the HTES technology have been calculated as a function of the operating temperature in the range of 600-1000 {sup circle} C. On the other hand, the effect of not only the gas turbine efficiency but also the recovery of waste heat for the overall hydrogen production thermal efficiency has also been evaluated. The thermal efficiency defined by a high heat value of the produced hydrogen (HHV) divided by the total energy of the heat and the electricity required to produce the hydrogen was adopted in our evaluation scheme. As a result, a maximized overall thermal efficiency of about 48% can be anticipated at 1000 {sup circle} C. Compared with a thermal efficiency of 27% by a conventional alkaline solution electrolysis at lower temperatures, a hydrogen production by the VHTR-assisted HTES can save on the required energy by about 2 times. The sensitivity of the operating temperature for the overall thermal efficiency defined by {delta}{eta}{sub {delta}} {sub T}/{eta}{sub T} is about 14.3% in the range of 600 to 1000 {sup circle} C. From the aspect of a conservative gas turbine efficiency and a reasonable recovery of the waste heat, an overall feasible efficiency of 46% is anticipated at 850 {sup circle} C. (author)

  20. Partial oxidation of methane in a temperature-controlled dielectric barrier discharge reactor

    KAUST Repository

    Zhang, Xuming

    2015-01-01

    We studied the relative importance of the reduced field intensity and the background reaction temperature in the partial oxidation of methane in a temperature-controlled dielectric barrier discharge reactor. We obtained important mechanistic insight from studying high-temperature and low-pressure conditions with similar reduced field intensities. In the tested range of background temperatures (297 < T < 773 K), we found that the conversion of methane and oxygen depended on both the electron-induced chemistry and the thermo-chemistry, whereas the chemical pathways to the products were overall controlled by the thermo-chemistry at a given temperature. We also found that the thermo-chemistry enhanced the plasma-assisted partial oxidation process. Our findings expand our understanding of the plasma-assisted partial oxidation process and may be helpful in the design of cost-effective plasma reformers. © 2014 The Combustion Institute.

  1. Uncertainty quantification of calculated temperatures for advanced gas reactor fuel irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh Thi-Cam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hawkes, Grant Lynn [Idaho National Lab. (INL), Idaho Falls, ID (United States); Einerson, Jeffrey James [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technology Research and Development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR tests, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The temperature simulation data for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. To quantify the uncertainty of AGR calculated temperatures, this study identifies and analyzes ABAQUS model parameters of potential importance to the AGR predicted fuel temperatures. The selection of input parameters for uncertainty quantification of the AGR calculated temperatures is based on the ranking of their influences on variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of AGR calculated temperatures. Expert judgment is used as the basis to specify the uncertainty range for selected input parameters. The input uncertainties are dynamic accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperature and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish

  2. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  3. Effect of Thermal Degradation on High Temperature Ultrasonic Transducer Performance in Small Modular Reactors

    Science.gov (United States)

    Bilgunde, Prathamesh N.; Bond, Leonard J.

    Prototype ultrasonic NDT transducers for use in immersion in coolants for small modular reactors have shown low signal to noise ratio. The reasons for the limitations in performance at high temperature are under investigation, and include changes in component properties. This current work seeks to quantify the issue of thermal expansion and degradation of the piezoelectric material in a transducer using a finite element method. The computational model represents an experimental set up for an ultrasonic transducer in a pulse-echo mode immersed in a liquid sodium coolant. Effect on transmitted and received ultrasonic signal due to elevated temperature (∼200oC) has been analysed.

  4. Design Strategies for Optically-Accessible, High-Temperature, High-Pressure Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  5. Design strategies for optically-accessible, high-temperature, high-pressure reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  6. Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

    National Research Council Canada - National Science Library

    Zhang, Guo-Qing; Wang, Shuai; Zhang, Hai-Qing; Zhu, Xing-Wang; Peng, Chao; Cai, Jun; He, Zhao-Zhong; Chen, Kun

    2017-01-01

    The release of fission products from coated particle fuel to primary coolant, as well as the activation of coolant and impurities, were analysed for a fluoride salt-cooled high-temperature reactor (FHR...

  7. Elevated-Temperature Ferritic and Martensitic Steels and Their Application to Future Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, RL

    2005-01-31

    In the 1970s, high-chromium (9-12% Cr) ferritic/martensitic steels became candidates for elevated-temperature applications in the core of fast reactors. Steels developed for conventional power plants, such as Sandvik HT9, a nominally Fe-12Cr-1Mo-0.5W-0.5Ni-0.25V-0.2C steel (composition in wt %), were considered in the United States, Europe, and Japan. Now, a new generation of fission reactors is in the planning stage, and ferritic, bainitic, and martensitic steels are again candidates for in-core and out-of-core applications. Since the 1970s, advances have been made in developing steels with 2-12% Cr for conventional power plants that are significant improvements over steels originally considered. This paper will review the development of the new steels to illustrate the advantages they offer for the new reactor concepts. Elevated-temperature mechanical properties will be emphasized. Effects of alloying additions on long-time thermal exposure with and without stress (creep) will be examined. Information on neutron radiation effects will be discussed as it applies to ferritic and martensitic steels.

  8. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  9. Evaluation of wrapper tubes temperature of neutron fast reactors by Transcoeur-2

    Energy Technology Data Exchange (ETDEWEB)

    Valentin, B.; Brun, P. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Chaigne, G. [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France)

    1995-12-31

    This paper deals with the thermal loading estimation of wrapper tubes by TRANSCOEUR-2 code. This estimation needs the knowledge of two temperature fields: the first is the peripheral sub-channels temperature of each sub-assembly computed by the design field computed by the thermohydraulic code TRIO-Vf with boundary conditions coming from CADET. The modeling of each code is presented as the first application of TRANSCOEUR-2 is performed on the European Fast Reactor (EFR) Core Design 6/92 (CD 6/91) in the nominal power conditions. The results show a temperature variation between the bottom and the top of the sub-assemblies fuel columns of 110 Celsius grades in the center of the core and 95 celsius grades at its periphery. The wrapper tubes temperatures are higher in the center than in the external side of the core. (authors). 2 refs., 8 figs.

  10. Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

    Science.gov (United States)

    Bowles, K. J.

    1974-01-01

    An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

  11. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Charles [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Peterson, Per [Univ. of California, Berkeley, CA (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  12. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  13. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    Science.gov (United States)

    Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.

    2014-02-01

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

  14. Transient analysis of nuclear graphite oxidation for high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Wei, E-mail: wxu12@mails.tsinghua.edu.cn; Shi, Lei; Zheng, Yanhua

    2016-09-15

    Graphite is widely used as moderator, reflector and structural materials in the high temperature gas-cooled reactor pebble-bed modular (HTR-PM). In normal operating conditions or water/air ingress accident, the nuclear graphite in the reactor may be oxidized by air or steam. Oxidation behavior of nuclear graphite IG-110 which is used as the structural materials and reflector of HTR-PM is mainly researched in this paper. To investigate the penetration depth of oxygen in IG-110, this paper developed the one dimensional spherical oxidation model. In the oxidation model, the equations considered graphite porosity variation with the graphite weight loss. The effect of weight loss on the effective diffusion coefficient and the oxidation rate was also considered in this model. Based on this theoretical model, this paper obtained the relative concentration and local weight loss ratio profile in graphite. In addition, the local effective diffusion coefficient and oxidation rate in the graphite were also investigated.

  15. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  16. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  17. Effect of inlet temperature on the performance of a catalytic reactor. [air pollution control

    Science.gov (United States)

    Anderson, D. N.

    1978-01-01

    A 12 cm diameter by 15 cm long catalytic reactor was tested with No. 2 diesel fuel in a combustion test rig at inlet temperatures of 700, 800, 900, and 1000 K. Other test conditions included pressures of 3 and 6 x 10 to the 5th power Pa, reference velocities of 10, 15, and 20 m/s, and adiabatic combustion temperatures in the range 1100 to 1400 K. The combustion efficiency was calculated from measurements of carbon monoxide and unburned hydrocarbon emissions. Nitrogen oxide emissions and reactor pressure drop were also measured. At a reference velocity of 10 m/s, the CO and unburned hydrocarbons emissions, and, therefore, the combustion efficiency, were independent of inlet temperature. At an inlet temperature of 1000 K, they were independent of reference velocity. Nitrogen oxides emissions resulted from conversion of the small amount (135 ppm) of fuel-bound nitrogen in the fuel. Up to 90 percent conversion was observed with no apparent effect of any of the test variables. For typical gas turbine operating conditions, all three pollutants were below levels which would permit the most stringent proposed automotive emissions standards to be met.

  18. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  19. EUROPAIRS: The European project on coupling of High Temperature Reactors with industrial processes

    Energy Technology Data Exchange (ETDEWEB)

    Angulo, C., E-mail: carmen.angulo@gdfsuez.com [Tractebel Engineering S.A. (GDF SUEZ), Avenue Ariane 7, 1200 Brussels (Belgium); Bogusch, E. [AREVA NP GmbH, Paul-Gossen-Strasse 100, 91052 Erlangen (Germany); Bredimas, A. [LGI Consulting, 37 rue de la Grange aux Belles, 75010 Paris (France); Delannay, N. [Tractebel Engineering S.A. (GDF SUEZ), Avenue Ariane 7, 1200 Brussels (Belgium); Viala, C. [AREVA NP SAS, 10 rue Juliette Recamier, 69456 Lyon Cedex 06 (France); Ruer, J.; Muguerra, Ph.; Sibaud, E. [SAIPEM S.A., 1/7 Avenue San Fernando, 78884 Saint Quentin en Yvelines Cedex (France); Chauvet, V. [LGI Consulting, 37 rue de la Grange aux Belles, 75010 Paris (France); Hittner, D. [AREVA NP Inc., 3315 Old Forest Road, Lynchburg, VA 24501 (United States); Fuetterer, M.A. [European Commission, Joint Research Centre, 1755ZG Petten (Netherlands); Groot, S. de [Nuclear Research and Consultancy Group, 1755ZG Petten (Netherlands); Lensa, W. von; Verfondern, K. [Forschungszentrum Juelich GmbH, Leo-Brandt-Strasse,52425 Juelich (Germany); Moron, R. [Solvay SA, rue du Prince Albert 33, 1050 Brussels (Belgium); Baudrand, O. [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP 17, 92262 Fontenay-aux-Roses cedex (France); Griffay, G. [Arcelor Mittal Maizieres Research SA, rue Luigi Cherubini 1A5, 39200 Saint Denis (France); Baaten, A. [USG/Baaten Energy Consulting, Burgermeester-Ceulen-Straat 78, 6212CT Maastricht (Netherlands); Segurado-Gimenez, J. [Tractebel Engineering S.A. (GDF SUEZ), Avenue Ariane 7, 1200 Brussels (Belgium)

    2012-10-15

    Developers of High Temperature Reactors (HTR) worldwide acknowledge that the main asset for market breakthrough is its unique ability to address growing needs for industrial cogeneration of heat and power (CHP) owing to its high operating temperature and flexibility, adapted power level, modularity and robust safety features. A strong alliance between nuclear and process heat user industries is a necessity for developing such a nuclear system for the conventional process heat market, just as the electro-nuclear development required a close partnership with utilities. Initiating such an alliance is one of the objectives of the EUROPAIRS project ( (www.europairs.eu)) presently on-going in the frame of the Euratom 7th Framework Programme (FP7). Although small and of short duration (21 months), EUROPAIRS is of strategic importance: it generates the boundary conditions for rapid demonstration of collocating HTR with industrial processes as proposed by the European High Temperature Reactor Technology Network (HTR-TN). This paper presents the main goals, the organization and the working approach of EUROPAIRS. It also presents the status of the viability assessment studies for coupling HTR with industrial end-user systems as one of the main pillars of the project. The main goal of the viability assessment is to identify developments required to remove the last technological and licensing barriers for a viable coupling scheme. The study is expected to result in guidelines for directing the choice of an industrial scale prototype.

  20. Preliminary Core Analysis of High Temperature Engineering Test Reactor Using DeCART Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The 2-dimensional core analysis for the High Temperature Engineering Test Reactor (HTTR) has been performed. The HTTR is a graphite-moderated and helium gas cooled reactor with an outlet temperature of 950 .deg. C and thermal output of 30 MW. In this study, the DECART code is used with a 190-group KARMA library. The calculation results are compared with those of the McCARD with the ENDF-B/VII.0 library. From the analysis results, it is known that the DeCART code generally overestimates k{sub inf} with a moderator temperature variation. In addition, it can be seen that the DeCART code predicts less negative MTC than the McCARD code. However, the DeCART code gives a slightly more negative FTC value. From the depletion results, the error of the DeCART decreases over the burnup until 600 FPD. The DeCART code gives very similar trend within the error of 190 pcm, which is very small error when compared with other result.

  1. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  2. Predictive functional control of temperature in a pharmaceutical hybrid nonlinear batch reactor

    Directory of Open Access Journals (Sweden)

    Štampar Simon

    2013-01-01

    Full Text Available These days, in times of recession, we are forced by competitiveness and the optimization of production to lower the costs of the temperature control in pharmaceutical batch reactors and increase the quantity and quality of the produced pharmaceutical product (active pharmaceutical substances. Therefore, a control algorithm is needed which provides us rapid and precise temperature control. This paper deals with the development of a control algorithm, where two predictive functional controllers are connected in a cascade for heating and cooling the content of the hybrid batch reactor. The algorithm has to be designed to cope with the constraints and the mixed discrete and continuous nature of the process of heating and cooling. The main goal of the control law is to achieve rapid and exact tracking of the reference temperature, good disturbance rejection and, in particular, a small number of heating and cooling medium switchings. The simulation results of the proposed algorithm give us much better performance compared to a conventional cascade PI algorithm.

  3. NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ugorowski, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ito, Takashi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

  4. Assessment of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    OpenAIRE

    Zhang, Z.; Dong, Y; Scherer, W

    2005-01-01

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemente...

  5. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2010-09-01

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  6. Hydrogen/Oxygen Reactions at High Pressures and Intermediate Temperatures: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    of the mixture was varied from oxidizing to reducing conditions. Moreover, a series of experiments in an oxygen atmosphere instead of a nitrogen atmosphere has been done. A reaction mechanism based on a recent work by Burke et al. has been developed. In addition to modeling of the present experiments......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time....

  7. Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants

    Energy Technology Data Exchange (ETDEWEB)

    McDonald, C.F.; Vrable, D.L.; Van Hagan, T.H.; King, J.H.; Spring, A.H.

    1980-02-01

    Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilization of nonmetallic materials and advanced heat exchanger surface geometries and methods of construction, is also discussed.

  8. Cyclic purging for low-temperature solid fusion reactor blanket operation

    Energy Technology Data Exchange (ETDEWEB)

    Was, G.S.; Lidsky, L.M.

    1979-05-01

    A purging process was developed that will permit operation of fusion reactor blankets employing solid LiAlO/sub 2/ as the breeder material at fuel temperatures of <600/sup 0/C. The low fuel temperature would greatly reduce the problems of fuel sintering, densification, and volume expansion that occur at fuel temperatures in excess of 900/sup 0/C without degrading the plant thermal efficiency. The process consists of heating the blanket to a specified temperature for a given time at regular intervals to release tritium held up in the breeding material. As an example, a detailed purging cycle was developed for the breeder rod shim rod blanket that uses LiAlO/sub 2/ in the form of micronsize particles compacted into millimeter-size pellets and is designed for low-temperature operation. Tritium inventory, doubling time, purging time, purging temperature, purging frequency, and particle size are the parameters used to evaluate the process. Calculations indicate that breeder particle sizes ranging from 20 to 50 ..mu..m and purging temperatures ranging from 600 to 700/sup 0/C can result in purge times of <1 h with three or more weeks between purges, and a doubling time of 7 yr for a blanket inventory limit of 5 kg and a breeding ratio of <1.02.

  9. New steady-state microbial community compositions and process performances in biogas reactors induced by temperature disturbances

    DEFF Research Database (Denmark)

    Luo, Gang; De Francisci, Davide; Kougias, Panagiotis

    2015-01-01

    Background The microbial community in a biogas reactor greatly influences the process performance. However, only the effects of deterministic factors (such as temperature and hydraulic retention time (HRT)) on the microbial community and performance have been investigated in biogas reactors. Little...... is known about the manner in which stochastic factors (for example, stochastic birth, death, colonization, and extinction) and disturbance affect the stable-state microbial community and reactor performances. Results In the present study, three replicate biogas reactors treating cattle manure were run...... to examine the role of stochastic factors and disturbance in shaping microbial communities. In the triplicate biogas reactors with the same inoculum and operational conditions, similar process performances and microbial community profiles were observed under steady-state conditions. This indicated...

  10. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  11. Degradation of phenol in an upflow anaerobic sludge blanket (UASB) reactor at ambient temperature.

    Science.gov (United States)

    Ke, Shui-zhou; Shi, Zhou; Zhang, Tong; Fang, Herbert H P

    2004-01-01

    A synthetic wastewater containing phenol as sole substrate was treated in a 2.8 L upflow anaerobic sludge blanket (UASB) reactor at ambient temperature. The operation conditions and phenol removal efficiency were discussed, microbial population in the UASB sludge was identified based on DNA cloning, and pathway of anaerobic phenol degradation was proposed. Phenol in wastewater was degraded in an UASB reactor at loading rate up to 18 gCOD/(L x d), with a 1:1 recycle ratio, at 26 +/- 1 degrees C, pH 7.0-7.5. An UASB reactor was able to remove 99% of phenol up to 1226 mg/L in wastewater with 24 h of hydraulic retention time (HRT). For HRT below 24 h, phenol degradation efficiency decreased with HRT, from 95.4% at 16 h to 93.8% at 12 h. It further deteriorated to 88.5% when HRT reached 8 h. When the concentration of influent phenol of the reactor was 1260 mg/L (corresponding COD 3000 mg/L), with the HRT decreasing (from 40 h to 4 h, corresponding COD loading increasing), the biomass yields tended to increase from 0.265 to 3.08 g/(L x d). While at 12 h of HRT, the biomass yield was lower. When HRT was 12 h, the methane yield was 0.308 L/(gCOD removed), which was the highest. Throughout the study, phenol was the sole organic substrate. The effluent contained only residual phenol without any detectable intermediates, such as benzoate, 4-hydrobenzoate or volatile fatty acids (VFAs). Based on DNA cloning analysis, the sludge was composed of five groups of microorganisms. Desulfotomaculum and Clostridium were likely responsible for the conversion of phenol to benzoate, which was further degraded by Syntrophus to acetate and H2/CO2. Methanogens lastly converted acetate and H2/CO2 to methane. The role of epsilon-Proteobacteria was, however, unsure.

  12. NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stevenson, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tsai, Kevin [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating.

  13. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  14. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    Directory of Open Access Journals (Sweden)

    Zhu Guifeng

    2016-01-01

    Full Text Available Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2 salt Temperature Reactivity Coefficient (TRC. Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern and two kinds of reflector materials (SiC and graphite. This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9Be(n,2n reaction and low neutron absorption of 6Li (even at 1000 ppm in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

  15. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

  16. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M.; Boehmert, J.; Gilles, R. [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  17. Analysis of Possible Application of High-Temperature Nuclear Reactors to Contemporary Large-Output Steam Power Plants on Ships

    Directory of Open Access Journals (Sweden)

    Kowalczyk T.

    2016-04-01

    Full Text Available This paper is aimed at analysis of possible application of helium to cooling high-temperature nuclear reactor to be used for generating steam in contemporary ship steam-turbine power plants of a large output with taking into account in particular variable operational parameters. In the first part of the paper types of contemporary ship power plants are presented. Features of today applied PWR reactors and proposed HTR reactors are discussed. Next, issues of load variability of the ship nuclear power plants, features of the proposed thermal cycles and results of their thermodynamic calculations in variable operational conditions, are presented.

  18. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  19. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    Directory of Open Access Journals (Sweden)

    Shutang Zhu

    2008-01-01

    Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

  20. Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K., E-mail: natesan@anl.go [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Li Meimei; Chopra, O.K.; Majumdar, S. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2009-07-15

    Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

  1. Numerical analysis of a SO{sub 3} packed column decomposition reactor with alloy RA 330 structural material for nuclear hydrogen production using the sulfur- iodine process

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Hyuk [Korea Maritime University, Busan (Korea, Republic of); Tak, Nam Il; Shin, Young Joon; Kim, Chan Soo; Lee, Ki Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-12-15

    A directly heated SO{sub 3} decomposer for the sulfur-iodine and hybrid-sulfur processes has been introduced and analyzed using the computational fluid dynamics (CFD) code CFX 11. The use of a directly heated decomposition reactor in conjunction with a very high temperature reactor (VHTR) allows for higher decomposition reactor operating temperatures. However, the high temperatures and strongly corrosive operating conditions associated with SO{sub 3} decomposition present challenges for the structural materials of decomposition reactors. In order to resolve these problems, we have designed a directly heated SO{sub 3} decomposer using RA 330 alloy as a structural material and have performed a CFD analysis of the design based on the finite rate chemistry model. The CFD results show the maximum temperature of the structural material could be maintained sufficiently below 1073 K, which is considered the target temperature for RA 330. The CFD simulations also indicated good performance in terms of SO{sub 3} decomposition for the design parameters of the present study

  2. Cooling performance of a water-cooling panel system for modular high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Shoji; Suzuki, Kunihiko; Inagaki, Yoshiyuki; Sudo, Yukio [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1995-12-31

    Experiments on a water cooling panel system were performed to investigate its heat removal performance and the temperature distribution of components for a modular high-temperature gas-cooled reactor (MHTGR). The analytical code THANPACST2 was applied to analyze the experimental results to verify the validity of the analytical method and the model.

  3. High Temperature and Pressure Alkaline Electrochemical Reactor for Conversion of Power to Chemicals

    DEFF Research Database (Denmark)

    Chatzichristodoulou, Christodoulos

    2016-01-01

    Moving away from fossil fuels requires harvesting more and more intermittent renewable energy resources and establishing a sustainable system for the production of chemicals. This brings forward the need for efficient large scale energy storage technologies 1-3 and technologies for the conversion...... of renewable electricity to chemicals. Electrochemical reactors can play a crucial role in this endeavor, since they can efficiently and reversibly transform electricity to high-value chemicals, and thus serve as energy storage and recovery devices for balancing the grid, while offering a means...... for the sustainable production of chemicals 4-6. A novel type of alkaline electrochemical cell that can operate at elevated temperature and pressure has been developed that relies on corrosion resistant high temperature diaphragms, based on mesoporous ceramic membranes where aqueous KOH is immobilized by capillary...

  4. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  5. Gamma spectroscopy for analysis of high temperature reactor fuel element KueFA tests

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission - Joint Research Centre, Eggenstein-Leopoldshafen (Germany). Institute for Transuranium Elements (JRC-ITU); Allelein, H.J. [RWTH Aachen Univ. (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik

    2013-07-01

    The High Temperature Reactor (HTR) is characterized by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRISO (TRi-ISOtropic) coating designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1600-1650 C, remaining well below the melting point of the fuel. An experimental assessment of the fuel behaviour under accident conditions is necessary to investigate the quality of fission product retention of TRISO coated particles in a given fuel element and to validate relevant computer codes. The device used to perform these studies is the cold finger apparatus KueFA (KuehlFinger-Apparatur). (orig.)

  6. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  7. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  8. TEMPERATURE MONITORING OPTIONS AVAILABLE AT THE IDAHO NATIONAL LABORATORY ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Daw; J.L. Rempe; D.L. Knudson; T. Unruh; B.M. Chase; K.L Davis

    2012-03-01

    As part of the Advanced Test Reactor National Scientific User Facility (ATR NSUF) program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced sensors for irradiation testing. To meet recent customer requests, an array of temperature monitoring options is now available to ATR users. The method selected is determined by test requirements and budget. Melt wires are the simplest and least expensive option for monitoring temperature. INL has recently verified the melting temperature of a collection of materials with melt temperatures ranging from 100 to 1000 C with a differential scanning calorimeter installed at INL’s High Temperature Test Laboratory (HTTL). INL encapsulates these melt wires in quartz or metal tubes. In the case of quartz tubes, multiple wires can be encapsulated in a single 1.6 mm diameter tube. The second option available to ATR users is a silicon carbide temperature monitor. The benefit of this option is that a single small monitor (typically 1 mm x 1 mm x 10 mm or 1 mm diameter x 10 mm length) can be used to detect peak irradiation temperatures ranging from 200 to 800 C. Equipment has been installed at INL’s HTTL to complete post-irradiation resistivity measurements on SiC monitors, a technique that has been found to yield the most accurate temperatures from these monitors. For instrumented tests, thermocouples may be used. In addition to Type-K and Type-N thermocouples, a High Temperature Irradiation Resistant ThermoCouple (HTIR-TC) was developed at the HTTL that contains commercially-available doped molybdenum paired with a niobium alloy thermoelements. Long duration high temperature tests, in furnaces and in the ATR and other MTRs, demonstrate that the HTIR-TC is accurate up to 1800 C and insensitive to thermal neutron interactions. Thus, degradation observed at temperatures above 1100 C with Type K and N thermocouples and decalibration due to transmutation with tungsten

  9. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  10. Experimental study and modeling of a high-temperature solar chemical reactor for hydrogen production from methane cracking

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, Stephane; Flamant, Gilles [Processes, Materials, and Solar Energy Laboratory, CNRS (PROMES-CNRS, UPR 8521), 7 Rue du Four Solaire, 66120 Odeillo Font-Romeu (France)

    2007-07-15

    A high-temperature fluid-wall solar reactor was developed for the production of hydrogen from methane cracking. This laboratory-scale reactor features a graphite tubular cavity directly heated by concentrated solar energy, in which the reactive flowing gas dissociates to form hydrogen and carbon black. The solar reactor characterization was achieved with: (a) a thorough experimental study on the reactor performance versus operating conditions and (b) solar reactor modeling. The results showed that the conversion of CH{sub 4} and yield of H{sub 2} can exceed 97% and 90%, respectively, and these depend strongly on temperature and on fluid-wall heat transfer and reaction surface area. In addition to the experimental study, a 2D computational model coupling transport phenomena was developed to predict the mapping of reactor temperature and of species concentration, and the reaction extent at the outlet. The model was validated and kinetics of methane decomposition were identified from simulations and comparison to experimental results. (author)

  11. Operation and Control Simulation of a Modular High Temperature Gas Cooled Reactor Nuclear Power Plant

    Science.gov (United States)

    Li, Haipeng; Huang, Xiaojin; Zhang, Liangju

    2008-08-01

    Issues in the operation and control of the multi-modular nuclear power plant are complicated. The high temperature gas cooled reactor pebble-bed module (HTR-PM) plant with two-module will be built as a demonstration plant in China. To investigate the operation and control characteristics of the plant, a simplified dynamic model is developed and mathematically formulated based upon the fundamental conversation of mass, energy and momentum. The model is implemented in a personal computer to simulate the power increase process of the HTR-PM operation. The open loop operation with no controller is first simulated and the results show that the essential parameter steam temperature varies drastically with time, which is not allowable in the normal operation. According to the preliminary control strategy of the HTR-PM, a simple steam temperature controller is proposed. The controller is of Proportional-type with a time lag. The closed loop operation with a steam temperature controller is then implemented and the simulation results show that the steam temperature and also other parameters are all well controlled in the allowable range.

  12. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission, Joint Research Center, Karlsruhe (Germany). Inst. for Transuranium Elements; Avincola, V. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. fuer Angewandte Materialien (IAM-AWP); Allelein, H.J. [RWTH Aachen Univ. (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik

    2013-11-15

    The High Temperature Reactor (HTR) is characterized by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRi-ISOtropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1,600 to 1,650 C, remaining well below the melting point of the fuel. Two key aspects associated with the safety of HTR fuel are assessed in this paper: fission product retention at temperatures up to 1,800 C is analyzed with the Cold Finger Apparatus (KueFA) while the behaviour of HTR-relevant fuel materials in an oxidizing environment is studied with the Corrosion Apparatus KORA. The KueFA is used to observe the combined effects of Depressurization and LOss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in helium atmosphere for several hundred hours, mimicking accident temperatures up to 1,800 C and realistic temperature transients. Nongaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analyzing plate deposits by means of High Purity Germanium (HPGe) gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. To achieve a good quantification of the release, a careful calibration of the setup is necessary and a collimator needs to be used in some cases. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Another relevant HTR accident scenario is air ingress into the reactor vessel as a consequence of a DLOFC incident. In

  13. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    Science.gov (United States)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  14. A Study on the High Temperature Irradiation Test Possibility for the HANARO Outer Core Region

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Cho, M. S.; Choo, K. N.; Shin, Y. T.; Sohn, J. M.; Park, S. J.; Kim, B. G

    2008-01-15

    1. Information on the neutron flux levels and the gamma heat of the concerned test holes, which have been produced from a series of nuclear analysis and tests performed at KAERI since 1993, were collected and analyzed to develop the nuclear data for the concerned test holes of HANARO and to develop the new design concepts of a capsule for the high temperature irradiation devices. 2. From the literature survey and analysis about the system design characteristics of the new concepts of irradiation devices in the ATR and MIT reactor, U.S. and the JHR reactor, France, which are helpful in understanding the key issues for the on-going R and D programmes related to a SFR and a VHTR, the most important parameters for the design of high temperature irradiation devices are identified as the neutron spectrum, the heat generation density, the fuel and cladding temperature, and the coolant chemistry. 3. From the thermal analysis of a capsule by using a finite element program ANSYS, high temperature test possibility at the OR and IP holes of HANARO was investigated based on the data collected from a literature survey. The OR holes are recommended for the tests of the SFR and VHTR nuclear materials. The IP holes could be applicable for an intermediate temperature irradiation of the SWR and LMR materials. 4. A thermal analysis for the development of a capsule with a new configuration was also performed. The size of the center hole, which is located at the thermal media of a capsule, did not cause specimen temperature changes. The temperature differences are found to be less than 2%. The introduction of an additional gap in the thermal media was able to contribute to an increase in the specimen temperature by up to 27-90 %.

  15. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  16. Application of optical fibers for optical diagnostics in high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shikama, T.; Narui, M. [Oarai Branch, Institute for Materials Research, Tohoku University, Ibaraki-ken (Japan); Kakuta, T. [Tokai Research Establishment, JAERI, Ibaraki-ken (Japan); Ishihara, M.; Sagawa, T.; Arai, T. [Oarai Research Establishment, JAERI, Ibaraki-ken (Japan)

    1998-09-01

    Visibility of a core region of a high temperature gas cooled reactor (HTGR) is very poor in general with its solid graphite moderator. Realization of optical diagnostics will improve safety and maintenance of the HTGR considerably. The applicability of fused silica core optical fibers for optical diagnostics in a core of the High Temperature Testing Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been studied in the present research. Optical diagnostics are also expected to play crucial roles in advanced research planned in the HTTR. Optical transmission of the optical fibers was found not to degrade for several hundred hours at 1070K in air and helium environments in the wavelength range of 350-1800nm. In general. the optical fibers were found to be heat-resistant. To study radiation effects, the optical fibers were irradiated in Japan Materials Testing Reactor (JMTR). where the fast neutron(E>1MeV) flux was up to 1.5x10{sup 18}n/m{sup 2}s and the gamma-ray dose rate was up to about 5W/g for iron. The estimated fast neutron flux and the gamma-ray dose rate would be in the order of 10{sup 16}n/m{sup 2} and about 0.1W/g for iron, respectively in the HTTR. In general, optical transmission loss increased substantially with a small irradiation dose in the visible wave length range, although some developed fibers showed better radiation resistance. Good optical transmissivity was kept in the infrared region with absorption rate of less than a few dB/m. Radioluminescence and thermoluminescence from sapphire and silica could be observed with optical fibers under irradiation. Cherenkov radiation was observed in the wavelength range of 600-1800nm, whose intensity was temperature-independent. Black-body radiation was dominant in the wavelength longer than 1200nm at elevated temperatures. The results showed that the silica core optical fibers could be used as an image guide as well as monitors for radiation dosimetry and for monitoring core

  17. Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL

    2005-12-01

    The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

  18. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    Energy Technology Data Exchange (ETDEWEB)

    Darmann, Frank [Zenergy Power, Inc., Burlingame, CA (United States); Lombaerde, Robert [Zenergy Power, Inc., Burlingame, CA (United States); Moriconi, Franco [Zenergy Power, Inc., Burlingame, CA (United States); Nelson, Albert [Zenergy Power, Inc., Burlingame, CA (United States)

    2012-03-01

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with warm bore diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged spider design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP's product development program, the amount of HTS

  19. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  20. Temperature and time influence on the waste plastics pyrolysis in the fixed bed reactor

    Directory of Open Access Journals (Sweden)

    Papuga Saša V.

    2016-01-01

    Full Text Available Pyrolysis as a technique of chemical recycling of plastic materials is causing an increasing level of interest as an environmentally and economically acceptable option for the processing of waste materials. Studies of these processes are carried out under different experimental conditions, in different types of reactors and with different raw materials, which makes the comparison of different processes and the direct application of process parameters quite complex. This paper presents the results of investigation of the influence of temperature in the range of 450°C to 525°C, on the yield of the process of pyrolysis of waste plastics mixture, composed of 45% polypropylene, 35% low density polyethylene and 25% high density polyethylene. Also, this paper presents results of the investigation of the effect of the reaction, atintervals of 30-90 [min], on the yield of pyrolysis of the mentioned waste plastics mixture. Research was conducted in a fixed bed pilot reactor, which was developed for this purpose. The results of the research show that at a temperature of 500°C, complete conversion of raw materials was achieved, for a period of 45 [min], with a maximum yield of the pyrolysis oil of 32.80%, yield of the gaseous products of 65.75% and the solid remains of 1.46%. Afurther increase of temperature increases the yield of gaseous products, at the expense of reducing the yield of pyrolysis oil. Obtained pyrolysis oil has a high calorific value of 45.96 [MJ/kg], and in this regard has potential applications as an alternative fuel.

  1. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

  2. Continuous flowing micro-reactor for aqueous reaction at temperature higher than 100 °C

    OpenAIRE

    Xie, Fei; Wang, Baojun; Wang, Wei; Dong, Tian; Tong, Jianhua; Xia, Shanhong; Wu, Wengang; Li, Zhihong

    2013-01-01

    Some aqueous reactions in biological or chemical fields are accomplished at a high temperature. When the reaction temperature is higher than 100 °C, an autoclave reactor is usually required to elevate the boiling point of the water by creating a high-pressure environment in a closed system. This work presented an alternative continuous flowing microfluidic solution for aqueous reaction with a reaction temperature higher than 100 °C. The pressure regulating function was successfully fulfilled ...

  3. Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

    2012-11-05

    This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

  4. Nonlinear Adaptive Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Dong, Zhe

    2013-04-01

    After the Fukushima nuclear accident, much more attention has to be drawn on the safety issues. The improvement of safety has already become the focus of the developing trend of the nuclear energy systems. Due to the inherent safety feature and the potential economic competitiveness, the modular high temperature gas-cooled reactor (MHTGR) has been seen as the central part of the next generation of nuclear plant (NGNP). Power-level control is one of the key techniques that guarantee the safe, stable and efficient operation for nuclear reactors. Since the MHTGR dynamics has the features of strong nonlinearity and uncertainty, in order to improve the operation performance, it is meaningful to develop the nonlinear adaptive power-level control law for the MHTGR. Based on using the natural dynamic features beneficial to system stabilization, a novel nonlinear adaptive power-level control is given for the MHTGR in this paper. It is theoretically proved that this newly-built controller does not only provide globally asymptotic closed-loop stability but is also adaptive to the system uncertainty. This control law is then applied to the power-level regulation of the pebble-bed MHTGR of the HTR-PM power plant. Numerical simulation results show the feasibility of this control law and the relationship between the performance and controller parameters.

  5. High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bell, Zane W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, Dominic R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lance, Michael J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Warmack, Robert J. Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Mark J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

  6. An analysis of the thermodynamic cycles with high-temperature nuclear reactor for power generation and hydrogen co-production

    Directory of Open Access Journals (Sweden)

    Dudek Michał

    2017-01-01

    Full Text Available In the present paper, numerical analysis of the thermodynamic cycle with the high-temperature nuclear gas reactor (HTGR for electricity and hydrogen production have been done. The analysed system consists of two independent loops. The first loop is for HTGR and consists of a nuclear reactor, heat exchangers, and blower. The second loop (Rankine cycle consist of up-to four steam turbines, that operate in heat recovery system. The analysis of the system shows that it is possible to achieve significantly higher efficiency than could be offered by traditional nuclear reactor technology (PWR and BWR. It is shown that the thermal efficiency about 52.5% it is possible to achieve when reactor works at standard conditions and steam is superheated up to 530oC. For the cases when the steam has supercritical conditions the value of thermal efficiency is still very high and equal about 50%.

  7. Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Jianghai Li

    2017-01-01

    Full Text Available High-temperature gas-cooled reactors (HTGR can incorporate wireless sensor network (WSN technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs. This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA and support vector machine (SVM are developed.

  8. Perspectives on understanding and verifying the safety terrain of modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Donald E., E-mail: donald@carlsonperin.net [11221 Empire Lane, Rockville, MD 20852 (United States); Ball, Sydney J., E-mail: beckysyd@comcast.net [100 Greywood Place, Oak Ridge, TN 37830 (United States)

    2016-09-15

    The passive safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving loss of forced cooling, coolant depressurization, air ingress, moisture ingress, and reactivity events. In addition to discussing associated uncertainties and potential measures to address them, this paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs.

  9. Fabrication of cermet bearings for the control system of a high temperature lithium cooled nuclear reactor

    Science.gov (United States)

    Yacobucci, H. G.; Heestand, R. L.; Kizer, D. E.

    1973-01-01

    The techniques used to fabricate cermet bearings for the fueled control drums of a liquid metal cooled reference-design reactor concept are presented. The bearings were designed for operation in lithium for as long as 5 years at temperatures to 1205 C. Two sets of bearings were fabricated from a hafnium carbide - 8-wt. % molybdenum - 2-wt. % niobium carbide cermet, and two sets were fabricated from a hafnium nitride - 10-wt. % tungsten cermet. Procedures were developed for synthesizing the material in high purity inert-atmosphere glove boxes to minimize oxygen content in order to enhance corrosion resistance. Techniques were developed for pressing cylindrical billets to conserve materials and to reduce machining requirements. Finishing was accomplished by a combination of diamond grinding, electrodischarge machining, and diamond lapping. Samples were characterized in respect to composition, impurity level, lattice parameter, microstructure and density.

  10. SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100

    Energy Technology Data Exchange (ETDEWEB)

    Vitro Engineering Company

    1964-07-15

    This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.

  11. Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Sacit M [ORNL

    2011-02-01

    This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

  12. Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Zuoyi; Scherer, W.

    1996-01-01

    This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H{sub 2}O concentration increase in the reactor core is smaller than 0.3 kg/(m{sup 3}s). The liquid water vaporization in the steam generator and H{sub 2}O transport from the steam generator to the reactor core reduce the ingress velocity of the H{sub 2}O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H{sub 2}O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [Deutsch] Dieser Bericht analysiert schwere Wassereinbruch-Stoerfaelle im 200 MW modularen Kugelhaufen-Hochtemperaturreaktor (HTR

  13. Approaches to experimental validation of high-temperature gas-cooled reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Belov, S.E. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Borovkov, M.N., E-mail: borovkov@okbm.nnov.ru [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Golovko, V.F.; Dmitrieva, I.V.; Drumov, I.V.; Znamensky, D.S.; Kodochigov, N.G. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Baxi, C.B.; Shenoy, A.; Telengator, A. [General Atomics, 3550 General Atomics Court, CA (United States); Razvi, J., E-mail: Junaid.Razvi@ga.com [General Atomics, 3550 General Atomics Court, CA (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. Black-Right-Pointing-Pointer Vibroacoustic investigations. Black-Right-Pointing-Pointer Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. Black-Right-Pointing-Pointer Experimental investigations of the catcher bearings design. - Abstract: The special feature of high-temperature gas-cooled reactors (HTGRs) is stressed operating conditions for equipment due to high temperature of the primary circuit helium, up to 950 Degree-Sign C, as well as acoustic and hydrodynamic loads upon the gas path elements. Therefore, great significance is given to reproduction of real operation conditions in tests. Experimental investigation of full-size nuclear power plant (NPP) primary circuit components is not practically feasible because costly test facilities will have to be developed for the power of up to hundreds of megawatts. Under such conditions, the only possible process to validate designs under development is representative tests of smaller scale models and fragmentary models. At the same time, in order to take in to validated account the effect of various physical factors, it is necessary to ensure reproduction of both individual processes and integrated tests incorporating needed integrated investigations. Presented are approaches to experimental validation of thermohydraulic and vibroacoustic characteristics for main equipment components and primary circuit path elements under standard loading conditions, which take account of their operation in the HTGR. Within the framework of the of modular helium reactor project, including a turbo machine in the primary circuit, a new and difficult problem is creation of multiple-bearing flexible vertical rotor. Presented are approaches to analytical and experimental validation of the rotor electromagnetic bearings, catcher bearings, flexible rotor

  14. Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

    2002-11-01

    In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

  15. Conducting thermomechanical fatigue test in air at light water reactor relevant temperature intervals

    Energy Technology Data Exchange (ETDEWEB)

    Ramesh, Mageshwaran [Paul Scherrer Institute, Laboratory for Nuclear Materials, CH-5232 Villigen-PSI (Switzerland); Leber, Hans J., E-mail: hans.leber@psi.ch [Paul Scherrer Institute, Laboratory for Nuclear Materials, CH-5232 Villigen-PSI (Switzerland); Diener, Markus; Spolenak, Ralph [Laboratory for Nanometallurgy, Department of Materials, ETH Zuerich, Wolfgang-Pauli-Strasse 10, CH-8093 Zurich (Switzerland)

    2011-08-01

    In Light Water Reactors (LWR), many structural components are made of austenitic stainless steels (SS). These components are subject to extreme conditions, such as large temperature gradients and pressure loads during service. Hence, the fatigue and fracture behavior of austenitic SS under these conditions has evoked consistent interest over the years. Most studies dealing with this problem in the past, investigated the isothermal fatigue (IF) condition, which is not the case in the service, and less attention has been paid to thermomechanical fatigue (TMF). Moreover, the existing codes of practice and standards for TMF testing are mainly derived from the high temperature TMF tests (T{sub mean} > 400 deg. C). This work presents the development of a facility to perform TMF tests under LWR relevant temperature interval in air. The realized testing parameters and tolerances are compared with the recommendations of existing codes of practice and standards from high temperature tests. The effectiveness of the testing facility was verified with series of TMF and IF tests performed on specimens made out of a commercial austenitic SS TP347 pipe material. The results revealed that the existing tolerances in standards are quite strict for the application of lower temperature ranges TMF tests. It was found that the synchronous, in-phase (IP) TMF tested specimens possess a higher lifetime than those subjected to the asynchronous, out-of-phase (OP) TMF and IF at T{sub max} in the investigated strain range for austenitic SS. Nevertheless, the fatigue lifetime of all the test conditions was similar in the engineering scale.

  16. Preliminary Safeguards Assessment for the Pebble-Bed Fluoride High-Temperature Reactor (PB-FHR) Concept

    Energy Technology Data Exchange (ETDEWEB)

    Disser, Jay; Arthur, Edward; Lambert, Janine

    2016-09-01

    This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.

  17. Removal efficiency and methanogenic activity profiles in a pilot-scale UASB reactor treating settled sewage at moderate temperatures

    NARCIS (Netherlands)

    Seghezzo, L.; Guerra, R.G.; González, S.M.; Trupiano, A.P.; Figueroa, M.E.; Cuevas, C.M.; Zeeman, G.; Lettinga, G.

    2002-01-01

    The performance of a sewage treatment system consisting of a settler followed by an Upflow Anaerobic Sludge Bed (UASB) reactor is described. Mean ambient and sewage temperature were 16.5 and 21.6 degrees C, respectively. Total Chemical Oxygen Demand (CODt) concentration averaged 224.2 and 152.6

  18. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  19. An numerical analysis of high-temperature helium reactor power plant for co-production of hydrogen and electricity

    Science.gov (United States)

    Dudek, M.; Podsadna, J.; Jaszczur, M.

    2016-09-01

    In the present work, the feasibility of using a high temperature gas cooled nuclear reactor (HTR) for electricity generation and hydrogen production are analysed. The HTR is combined with a steam and a gas turbine, as well as with the system for heat delivery for medium temperature hydrogen production. Industrial-scale hydrogen production using copper-chlorine (Cu-Cl) thermochemical cycle is considered and compared with high temperature electrolysis. Presented cycle shows a very promising route for continuous, efficient, large-scale and environmentally benign hydrogen production without CO2 emissions. The results show that the integration of a high temperature helium reactor, with a combined cycle for electric power generation and hydrogen production, may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  20. Temperature dependence of the rate constant of hydrogen isotope interactions with a lithium capillary-porous system under reactor irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibayeva, Irina, E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy NNC RK, Kurchatov (Kazakhstan); Kulsartov, Timur; Gordienko, Yuri [Institute of Atomic Energy NNC RK, Kurchatov (Kazakhstan); Mukanova, Aliya [Al’ Farabi Kazakh National University, Almaty (Kazakhstan); Ponkratov, Yuri; Barsukov, Nikolay; Tulubaev, Evgeniy [Institute of Atomic Energy NNC RK, Kurchatov (Kazakhstan); Platacis, Erik [University of Latvia (IPUL), Riga (Latvia); Kenzhin, Ergazy [Shakarim Semey State University, Semey (Kazakhstan)

    2013-10-15

    Highlights: • The experiments with Li CPS sample were carried out at reactor IVG-1.M. • The gas absorption technique was used to study hydrogen isotope interaction with lithium CPS. • The temperature dependence of constants of interaction rate was obtained for various power rates of the reactor. • Determination of the activation energies, and pre-exponents of Arrhenius dependence. • The effect of increase of the rate constant under reaction irradiation. -- Abstract: Experiments with a sample of a lithium capillary-porous system (CPS) were performed at the reactor IVG-1.M of the Institute of Atomic Energy NNC RK to study the effects of neutron irradiation on the parameters of hydrogen isotope interactions with a lithium CPS. The absorption technique was used during the experiments, and this technique allowed the temperature dependences of the hydrogen isotope interaction rate constants with the lithium CPS to be obtained under various reactor powers. The obtained dependencies were used to determine the main interaction parameters: the activation energies and the pre-exponents of the Arrhenius dependence of the hydrogen interaction rate constants with lithium and the lithium CPS. An increase of the hydrogen isotope interaction rate with the lithium CPS was observed under reactor irradiation.

  1. Use and Storage of Test and Operations Data from the High Temperature Test Reactor Acquired by the US Government from the Japan Atomic Energy Agency

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2010-02-01

    This document describes the use and storage of data from the High Temperature Test Reactor (HTTR) acquired from the Japan Atomic Energy Agency (JAEA) by the U.S. Government for high temperature reactor research under the Next Generation Nuclear Plant (NGNP) Project.

  2. CONVECTION REACTOR

    Science.gov (United States)

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  3. Effect of tempering temperature on the microstructure and mechanical properties of a reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Li, C.W.; Han, L.Z.; Luo, X.M.; Liu, Q.D.; Gu, J.F., E-mail: gujf@sjtu.edu.cn

    2016-08-15

    The microstructure and mechanical properties of reactor pressure vessel (RPV) steel were investigated after tempering at different temperatures ranging from 580 to 700 °C for 5 h. With increasing tempering temperature, the impact toughness, which is qualified by Charpy V-notch total absorbed energy, initially increases from 142 to 252 J, and then decreases to 47 J, with a maximum value at 650 °C, while the ultimate tensile strength varies in exactly the opposite direction. Comparing the microstructure and fracture surfaces of different specimens, the variations in toughness and strength with the tempering temperature were generally attributed to the softening of the bainitic ferrite, the agminated Fe{sub 3}C carbides that resulted from decomposition of martensite/austenite (M/A) constituents, the precipitation of Mo{sub 2}C carbides, and the newly formed M/A constituents at the grain boundaries. Finally, the correlation between the impact toughness and the volume fraction of the M/A constituents was established, and the fracture mechanisms for the different tempering conditions are explained. - Highlights: • The dependence of the deterioration of impact toughness on tempering temperature has been analysed. • The instrumented Charpy V-notch impact test has been employed to study the fracture mechanism. • The influence of M/A constituents on different fracture mechanisms based on the hinge model has been demonstrated. • A correlation between the mechanical properties and the amount of M/A constituents has been established.

  4. Studies of low temperature, low flux radiation embrittlement of nuclear reactor structural materials. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Lucas, G.E.

    1998-09-02

    A large matrix of simple alloys and complex commercial type steels was irradiated over a range of fluxes at 60 C up to a fast fluence of about 3 {times} 10{sup 22} n/m{sup 2}. Combined with data in the literature, these results show a negligible effect of flux on irradiation hardening in the range of 2 {times} 10{sup 13} to 5 {times} 10{sup 18} n/m{sup 2}-s. This observation lends indirect support to the proposal that the accelerated embrittlement in the High Flux Isotope Reactor surveillance steels was due to an anomalously high level of damage from gamma rays. A weak dependence of hardening on a number of elements, including copper, nickel, phosphorus, molybdenum and manganese, can be described by a simple empirical chemistry factor. Particular combinations of elements resulted in hardening differences of up to about 60% in the complex commercial type steels and up to about 100% in simple model alloys. Direct effects of microstructure appear to be minimal. Hardening varies with the square root of fluence above a threshold around 4 {times} 10{sup 20} n/m{sup 2}. The results suggest that low temperature hardening is dominated by local intracascade processes leading to the formation of small defect-solute clusters/complexes. The observed hardening corresponds to nominal maximum end-of-life transition temperature shifts in support structure steels of about 120 C.

  5. Treating municipal solid waste leachate in a pilot scale upflow anaerobic sludge blanket reactor under tropical temperature

    Directory of Open Access Journals (Sweden)

    Abbas Alizadeh Shooshtari

    2012-01-01

    Full Text Available Aims: The objective of this study was to investigate an Upflow Anaerobic Sludge Blanket (UASB reactor efficiency in treating municipal landfill leachate, under tropical temperature. Materials and Methods: A 30-liter pilot-scale UASB reactor was used to treat the municipal solid waste leachate, under tropical temperature, for 230 days. The reactor was inoculated with 10 liters of anaerobic sludge from an anaerobic digester, in an agro industry′s wastewater treatment plant. The Volatile Suspended Solids (VSS of sludge were 65 g/L, with volatile suspended solids to suspended solids (VSS/SS ratio of 0.74. The reactor was operated in mesophilic (34 - 39°C temperature. Results: After reaching a stable operation, the reactor was exposed to raw leachate, with mean chemical oxygen demand (COD concentrations of 35 g/L. The leachate was diluted to 9 - 10 g/L at Organic Loading Rates (OLRs of 2, 6, 12, 15 g COD/L.d and decreased again to 12 g COD/L.d, resulting in 45, 76, 84, 68, and 79% removal efficiency and increased again to 87% removal efficiency for COD, at Hydraulic Retention Times (HRTs of 6, 1.6, 0.83, and 0.67 days, respectively, in the UASB. In the reactor used in this study, the heavy metals were removed by adsorption on biomass, and the maximum removal rate was 68% for Zinc (Zn. Conclusions: It was concluded that the optimum OLR for diluted leachate up to 10 g COD/l, was 12 g COD/L.d at an HRT of 0.67 day (16 hours.

  6. A kinetic study of gaseous potassium capture by coal minerals in a high temperature fixed-bed reactor

    DEFF Research Database (Denmark)

    Zheng, Yuanjing; Jensen, Peter Arendt; Jensen, Anker Degn

    2008-01-01

    The reactions between gaseous potassium chloride and coal minerals were investigated in a lab-scale high temperature fixed-bed reactor using single sorbent pellets. The applied coal minerals included kaolin, mullite, silica, alumina, bituminous coal ash, and lignite coal ash that were formed...... when heated at temperatures above 450°C. The amounts of potassium captured by metakaolin pellet decreases with increasing reaction temperature in the range of 900-1300°C and increases again with further increasing the temperature up to 1500°C. There is no reaction of pre-made mullite with KCl...

  7. Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

    Directory of Open Access Journals (Sweden)

    Shanfang Huang

    2017-01-01

    Full Text Available High thermal neutron fluxes are needed in some research reactors and for irradiation tests of materials. A High Flux Research Reactor (HFRR with an inverse flux trap-converter target structure is being developed by the Reactor Engineering Analysis Lab (REAL at Tsinghua University. This paper studies the safety of the HFRR core by full core flow and temperature calculations using the porous media approach. The thermal nonequilibrium model is used in the porous media energy equation to calculate coolant and fuel assembly temperatures separately. The calculation results show that the coolant temperature keeps increasing along the flow direction, while the fuel temperature increases first and decreases afterwards. As long as the inlet coolant mass flow rate is greater than 450 kg/s, the peak cladding temperatures in the fuel assemblies are lower than the local saturation temperatures and no boiling exists. The flow distribution in the core is homogeneous with a small flow rate variation less than 5% for different assemblies. A large recirculation zone is observed in the outlet region. Moreover, the porous media model is compared with the exact model and found to be much more efficient than a detailed simulation of all the core components.

  8. Experimental investigation of pebble flow dynamics using radioactive particle tracking technique in a scaled-down Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Khane, Vaibhav; Said, I.A.; Al-Dahhan, Muthanna H., E-mail: aldahhanm@mst.edu

    2016-06-15

    Highlights: • Pebble Flow fields at Pebble Bed Modular Reactor was investigated. • Radioactive Particle Tracking (RPT) technique has been used. • Plug flow type velocity profile is suggested at upper cylindrical region. - Abstract: The Pebble Bed Modular Reactor (PBMR) is a type of very-high-temperature reactor (VHTR) that is conceptually very similar to moving bed reactors used in the chemical and petrochemical industries. In a PBMR core, nuclear fuel is in the form of pebbles and moves slowly under the influence of gravity. In this work, an integrated experimental and computational study of granular flow in a scaled-down cold flow PBMR was performed. A continuous pebble re-circulation experimental set-up, mimicking the flow of pebbles in a PBMR was designed and developed. An experimental investigation of pebble flow dynamics in a scaled down test reactor was carried out using a non-invasive radioactive particle tracking (RPT) technique that used a cobalt-60 based tracer to mimic pebbles in terms of shape, size and density. A cross-correlation based position reconstruction algorithm and RPT calibration data were used to obtain results about Lagrangian trajectories, the velocity field, and residence time distributions. The RPT technique results a serve as a benchmark data for assessing contact force models used in the discrete element method (DEM) simulations.

  9. Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

    2016-12-15

    Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

  10. Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-08-15

    The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

  11. Feasibility of Thorium Fuel Cycles in a Very High Temperature Pebble-Bed Hybrid System

    Directory of Open Access Journals (Sweden)

    L.P. Rodriguez

    2015-08-01

    Full Text Available Nuclear energy presents key challenges to be successful as a sustainable energy source. Currently, the viability of the use thorium-based fuel cycles in an innovative nuclear energy generation system is being investigated in order to solve these key challenges. In this work, the feasibility of three thorium-based fuel cycles (232Th-233U, 232Th-239Pu, and 232Th-U in a hybrid system formed by a Very High Temperature Pebble-Bed Reactor (VHTR and two Pebble-Bed Accelerator Driven Systems (ADSs was evaluated using parameters related to the neutronic behavior such as nuclear fuel breeding, minor actinide stockpile, the energetic contribution of each fissile isotope, and the radiotoxicity of the long lived wastes. These parameters were used to compare the fuel cycles using the well-known MCNPX ver. 2.6e computational code. The results obtained confirm that the 232Th-233U fuel cycle is the best cycle for minimizing the production of plutonium isotopes and minor actinides. Moreover, the inclusion of the second stage in the ADSs demonstrated the possibility of extending the burnup cycle duration and reducing the radiotoxicity of the discharged fuel from the VHTR.

  12. A Two-Dimensional Multiphysics Coupling Model of a Middle and Low Temperature Solar Receiver/Reactor for Methanol Decomposition

    Directory of Open Access Journals (Sweden)

    Yanjuan Wang

    2017-10-01

    Full Text Available Abstract: In this paper, the endothermic methanol decomposition reaction is used to obtain syngas by transforming middle and low temperature solar energy into chemical energy. A two-dimensional multiphysics coupling model of a middle and low temperature of 150~300 °C solar receiver/reactor was developed, which couples momentum equation in porous catalyst bed, the governing mass conservation with chemical reaction, and energy conservation incorporating conduction/convection/radiation heat transfer. The complex thermochemical conversion process of the middle and low temperature solar receiver/reactor (MLTSRR system was analyzed. The numerical finite element method (FEM model was validated by comparing it with the experimental data and a good agreement was obtained, revealing that the numerical FEM model is reliable. The characteristics of chemical reaction, coupled heat transfer, the components of reaction products, and the temperature fields in the receiver/reactor were also revealed and discussed. The effects of the annulus vacuum space and the glass tube on the performance of the solar receiver/reactor were further studied. It was revealed that when the direct normal irradiation increases from 200 W/m2 to 800 W/m2, the theoretical efficiency of solar energy transformed into chemical energy can reach 0.14–0.75. When the methanol feeding rate is 13 kg/h, the solar flux increases from 500 W/m2 to 1000 W/m2, methanol conversion can fall by 6.8–8.9% with air in the annulus, and methanol conversion can decrease by 21.8–28.9% when the glass is removed from the receiver/reactor.

  13. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    Science.gov (United States)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-01-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  14. "A New Class od Functionally Graded Cearamic-Metal Composites for Next Generation Very High Temperature Reactors"

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Mohit Jain; Dr. Ganesh Skandan; Dr. Gordon E. Khose; Mrs. Judith Maro, Nuclear Reactor Laboratory, MIT

    2008-05-01

    Generation IV Very High Temperature power generating nuclear reactors will operate at temperatures greater than 900 oC. At these temperatures, the components operating in these reactors need to be fabricated from materials with excellent thermo-mechanical properties. Conventional pure or composite materials have fallen short in delivering the desired performance. New materials, or conventional materials with new microstructures, and associated processing technologies are needed to meet these materials challenges. Using the concept of functionally graded materials, we have fabricated a composite material which has taken advantages of the mechanical and thermal properties of ceramic and metals. Functionally-graded composite samples with various microstructures were fabricated. It was demonstrated that the composition and spatial variation in the composition of the composite can be controlled. Some of the samples were tested for irradiation resistance to neutrons. The samples did not degrade during initial neutron irradiation testing.

  15. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  16. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1980-September 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-12

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, 950 and 1050/sup 0/C. Initiation of controlled purity helium creep-rupture testing in the intensive screening test program is discussed. In addition, the results of 1000-hour exposures at 750 and 850/sup 0/C on several experimental alloys are discussed.

  17. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-11-14

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.

  18. Physically-Based Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Dong, Zhe

    2012-10-01

    Because of its strong inherent safety, the modular high temperature gas-cooled nuclear reactor (MHTGR) has been regarded as the central part of the next generation nuclear plants (NGNPs). Power-level control is one of the key techniques which provide safe, stable and efficient operation for the MHTGRs. The physically-based regulation theory is definitely a promising trend of modern control theory and provides a control design method that can suppress the unstable part of the system dynamics and remain the stable part. Usually, the control law designed by the physically-based control theory has a simple form and high performance. Stimulated by this, a novel nonlinear dynamic output feedback power-level control is established in this paper for the MHTGR based upon its own dynamic features. This newly-built control strategy guarantees the globally asymptotic stability and provides a satisfactory transient performance through properly adjusting the feedback gains. Simulation results not only verify the correctness of the theoretical results but also illustrate the high control performance.

  19. Utilization of heat of modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ide, A. [Fuji Electric, Tokyo (Japan). Nuclear Power Promotion Dept.; Takenaka, Y. [Kawasaki Heavy Industries, Tokyo (Japan). Nuclear Systems Div.; Maeda, S. [Ube Industries, Yamaguchi (Japan). Machinery Dept.

    1996-07-01

    The demand for energy is increasing worldwide along with increases in population and rises in the standard of living. If the needed energy is supplied only by fossil fuels, environmental problems will impose limits on human activities. Recognizing that more than 60% of the energy consumed in Japan is non-electrical energy, FAPIG organized the HTR-HUC Working Group to study methods of using heat from high temperature gas-cooled reactors (HTR) to mitigate environmental and energy resource problems, and to contribute to the steady supply and effective use of energy. The authors chose three types of model plants to study: (1) a cogeneration plant which can be built with existing technology; (2) a coal gasification plant which can accelerate the clean use of coal and contribute to a stable supply of energy and the preservation of the environment; and (3) a hydrogen production plant whose hydrogen will release people from their dependence on fossil energy. For each of the above plants, a system outline and basic plan as well as costs, resultant social effects, management methods of the operating company and technical issues are studied.

  20. Low temperature biological phosphorus removal and partial nitrification in a pilot sequencing batch reactor system.

    Science.gov (United States)

    Yuan, Qiuyan; Oleszkiewicz, Jan A

    2011-01-01

    Partial nitrification and biological phosphorus removal appear to hold promise of a cost-effective and sustainable biological nutrient removal process. Pilot sequencing batch reactors (SBRs) were operated under anaerobic/aerobic configuration for 8 months. It was found that biological phosphorus removal can be achieved in an SBR system, along with the partial nitrification process. Sufficient volatile fatty acids supply was the key for enhanced biological phosphorus removal. This experiment demonstrated that partial nitrification can be achieved even at low temperature with high dissolved oxygen (>3 mg/L) concentration. Shorter solid retention time (SRT) for nitrite oxidizing bacteria (NOB) than for ammonia oxidizing bacteria due to the nitrite substrate limitation at the beginning of the aeration cycle was the reason that caused NOB wash-out. Controlling SRT should be the strategy for an SBR operated in cold climate to achieve partial nitrification. It was also found that the aerobic phosphorus accumulating organisms' P-uptake was more sensitive to nitrite inhibition than the process of anaerobic P-release.

  1. Inquiry into disintegration control of irradiated low enrichment uranium for high temperature gas cooled reactors

    Science.gov (United States)

    Reitsamer, G.; Stolba, G.; Falta, G.; Strigl, A.; Zeger, J.; Maly, V.

    1984-07-01

    The PyC-coatings of irradiated high temperature reactor (HTR) fuel particles from AVR fuel elements were burnt off by air and oxygen at 1120K (comparable to the current HTR head-end). Experiments on the solubility of this low enrichment uranium fuel prove that 99.93% of the uranium and 99.84% of the plutonium can be dissolved by 7n HNO3. After additional treatment with 7n HNO3/0.01 n NaF, only 0.01% of the original amount of uranium and 0.01% of the original amount of plutonium remain undissolved. Neither the insoluble residues nor the very small amounts of solids formed on standing (before and after concentrating the solution up to 200 g U/1 and acidity of 3 n HNO3) show any enrichment of plutonium compared with the nitric acid solution. Results indicate that LWR-PUREX-technology can be used for reprocessing HTR-LEU-fuel.

  2. Time-Resolved Kinetic Chirped-Pulse Rotational Spectroscopy in a Room-Temperature Flow Reactor.

    Science.gov (United States)

    Zaleski, Daniel P; Harding, Lawrence B; Klippenstein, Stephen J; Ruscic, Branko; Prozument, Kirill

    2017-12-11

    Chirped-pulse Fourier transform millimeter-wave spectroscopy is a potentially powerful tool for studying chemical reaction dynamics and kinetics. Branching ratios of multiple reaction products and intermediates can be measured with unprecedented chemical specificity; molecular isomers, conformers, and vibrational states have distinct rotational spectra. Here we demonstrate chirped-pulse spectroscopy of vinyl cyanide photoproducts in a flow tube reactor at ambient temperature of 295 K and pressures of 1-10 μbar. This in situ and time-resolved experiment illustrates the utility of this novel approach to investigating chemical reaction dynamics and kinetics. Following 193 nm photodissociation of CH2CHCN, we observe rotational relaxation of energized HCN, HNC, and HCCCN photoproducts with 10 μs time resolution and sample the vibrational population distribution of HCCCN. The experimental branching ratio HCN/HCCCN is compared with a model based on RRKM theory using high-level ab initio calculations, which were in turn validated by comparisons to Active Thermochemical Tables enthalpies.

  3. Effect of fuel burnup and cross sections on modular HTGR (High-Temperature Gas-cooled Reactor) reactivity coefficients

    Science.gov (United States)

    Lefler, W.; Baxter, A.; Mathews, D.

    1987-12-01

    The temperature dependence of the reactivity coefficient in a prismatic block Modular High-Temperature Gas-Cooled Reactor (MHTGR) design is examined and found to be large and negative. Temperature coefficient results obtained with the ENDF/B-V data library were almost the same as results obtained with the earlier versions of the ENDF/B data library usually used at GA Technologies Inc., in spite of a significant eigenvalue increase with the ENDF/B-V data. The effects of fuel burnup and arbitrarily assumed cross section variations were examined and tabulated.

  4. Dynamic Time-Resolved Chirped-Pulse Rotational Spectroscopy of Vinyl Cyanide Photoproducts in a Room Temperature Flow Reactor

    Science.gov (United States)

    Zaleski, Daniel P.; Prozument, Kirill

    2017-06-01

    Chirped-pulsed (CP) Fourier transform rotational spectroscopy invented by Brooks Pate and coworkers a decade ago is an attractive tool for gas phase chemical dynamics and kinetics studies. A good reactor for such a purpose would have well-defined (and variable) temperature and pressure conditions to be amenable to accurate kinetic modeling. Furthermore, in low pressure samples with large enough number of molecular emitters, reaction dynamics can be observable directly, rather than mediated by supersonic expansion. In the present work, we are evaluating feasibility of in situ time-resolved CP spectroscopy in a room temperature flow tube reactor. Vinyl cyanide (CH_2CHCN), neat or mixed with inert gasses, flows through the reactor at pressures 1-50 μbar (0.76-38 mTorr) where it is photodissociated by a 193 nm laser. Millimeter-wave beam of the CP spectrometer co-propagates with the laser beam along the reactor tube and interacts with nascent photoproducts. Rotational transitions of HCN, HNC, and HCCCN are detected, with ≥10 μs time-steps for 500 ms following photolysis of CH_2CHCN. The post-photolysis evolution of the photoproducts' rotational line intensities is investigated for the effects of rotational and vibrational thermalization of energized photoproducts. Possible contributions from bimolecular and wall-mediated chemistry are evaluated as well.

  5. Removal efficiency and methanogenic activity profiles in a pilot-scale UASB reactor treating settled sewage at moderate temperatures.

    Science.gov (United States)

    Seghezzo, L; Guerra, R G; González, S M; Trupiano, A P; Figueroa, M E; Cuevas, C M; Zeeman, G; Lettinga, G

    2002-01-01

    The performance of a sewage treatment system consisting of a settler followed by an Upflow Anaerobic Sludge Bed (UASB) reactor is described. Mean ambient and sewage temperature were 16.5 and 21.6 degrees C, respectively. Total Chemical Oxygen Demand (CODt) concentration averaged 224.2 and 152.6 mg/L, for raw and settled sewage, respectively. The effluent concentration was 68.5 mgCODt/L. Total and suspended COD removal efficiencies of approximately 70 and 80%, respectively, have been observed in the system at a mean Hydraulic Retention Time (HRT) of 2 + 5 h. Maximum COD removal efficiency was achieved in the UASB reactor when upflow velocity (Vup) was 0.43 m/h (HRT = 6 h). Mean Specific Methanogenic Activity (SMA) and Volatile Suspended Solids (VSS) concentration in the granular sludge bed were 0.11 gCOD-CH4/gVSS.d and 30.0 gVSS/Lsludge, respectively. SMA was inversely related to VSS concentration, and both parameters varied along the sludge bed height. The Solids Retention Time (SRT) in the reactor was 450 days. Sludge characteristics have not been affected by changes of up to one month in Vup in the range 0.28-0.85 m/h (HRT 3-9 h). This system or two UASB reactors in series could be an alternative for sewage treatment under moderate temperature conditions.

  6. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2011-10-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  7. Theoretical and Experimental Evaluation of the Temperature Distribution in a Dry Type Air Core Smoothing Reactor of HVDC Station

    Directory of Open Access Journals (Sweden)

    Yu Wang

    2017-05-01

    Full Text Available The outdoor ultra-high voltage (UHV dry-type air-core smoothing reactors (DASR of High Voltage Direct Current systems are equipped with a rain cover and an acoustic enclosure. To study the convective heat transfer between the DASR and the surrounding air, this paper presents a coupled model of the temperature and fluid field based on the structural features and cooling manner. The resistive losses of encapsulations calculated by finite element method (FEM were used as heat sources in the thermal analysis. The steady fluid and thermal field of the 3-D reactor model were solved by the finite volume method (FVM, and the temperature distribution characteristics of the reactor were obtained. Subsequently, the axial and radial temperature distributions of encapsulation were investigated separately. Finally, an optical fiber temperature measurement scheme was used for an UHV DASR under natural convection conditions. Comparative analysis showed that the simulation results are in good agreement with the experimental data, which verifies the rationality and accuracy of the numerical calculation. These results can serve as a reference for the optimal design and maintenance of UHV DASRs.

  8. Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zheng Yanhua, E-mail: zhengyh@mail.tsinghua.edu.c [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Shi Lei; Wang Yan [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-10-15

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of 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 and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

  9. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  10. The Low Temperature Oxidation of 2,7-Dimethyloctane in a Pressurized Flow Reactor

    Science.gov (United States)

    Farid, Farinaz

    The complexity of real fuels has fostered the use of simple mixtures of hydrocarbons whose combustion behavior approximates that of real fuels in both experimental and computational studies to develop models of the combustion of the real fuel. These simple mixtures have been called surrogates. Lightly branched paraffins are an important class of constituents in gasoline, diesel and aviation turbine fuels and therefore are primary candidates for use as a component in a surrogate. Unfortunately, fundamental studies on combustion characteristics of high molecular weight mono- and di-methylated iso-paraffins are scarce. Therefore, this study was designed to investigate the low-temperature oxidation of 2,7-dimethyloctane (2,7-DMO) (C10H22), a lightly branched isomer of decane. Replicate 2,7-DMO oxidation experiments were conducted in a pressurized flow reactor (PFR) over the temperature range of 550 -- 850 K, at a pressure of 8 atm and an equivalence ratio of 0.3 in 4.21% oxygen / nitrogen. The reactivity was mapped by continuous monitoring of CO, CO 2, and O2 using a non-dispersive infrared (NDIR) carbon monoxide / carbon dioxide analyzer and an electrochemical oxygen sensor. For examining the underlying reaction chemistry, detailed speciation of samples was performed at selected temperatures using a gas chromatograph with a flame ionization detector coupled to a mass spectrometer. Comparable oxidation experiments for n-decane were carried out to examine the unique effects of branching on fuel reactivity and distribution of major stable intermediates. For both isomers, the onset of negative temperature coefficient (NTC) region was observed near 700 K, with the reactivity decreasing with increasing the temperature. The flow reactor study of n-decane oxidation confirmed that the isomerization reduces the amount of CO produced at peak reactivity. In addition to reaction inhibition, branching affected the distribution of C2-C 4 olefin intermediates. While the oxidation of

  11. Continuous flowing micro-reactor for aqueous reaction at temperature higher than 100 °C.

    Science.gov (United States)

    Xie, Fei; Wang, Baojun; Wang, Wei; Dong, Tian; Tong, Jianhua; Xia, Shanhong; Wu, Wengang; Li, Zhihong

    2013-01-01

    Some aqueous reactions in biological or chemical fields are accomplished at a high temperature. When the reaction temperature is higher than 100 °C, an autoclave reactor is usually required to elevate the boiling point of the water by creating a high-pressure environment in a closed system. This work presented an alternative continuous flowing microfluidic solution for aqueous reaction with a reaction temperature higher than 100 °C. The pressure regulating function was successfully fulfilled by a small microchannel based on a delicate hydrodynamic design. Combined with micro heater and temperature sensor that integrated in a single chip by utilizing silicon-based microfabrication techniques, this pressure regulating microchannel generated a high-pressure/high-temperature environment in the upstream reaction zone when the reagents continuously flow through the chip. As a preliminary demonstration, thermal digestion of aqueous total phosphorus sample was achieved in this continuous flowing micro-reactor at a working pressure of 990 kPa (under the working flow rate of 20 nl/s) along with a reaction temperature of 145 °C. This continuous flowing microfluidic solution for high-temperature reaction may find applications in various micro total analysis systems.

  12. Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

    Directory of Open Access Journals (Sweden)

    E. A. Kuleshova

    2017-01-01

    Full Text Available This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C. The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

  13. Analysis of the gas diffusion process during a hypothetical air ingress accident in a modular high temperature gas cooled reactor

    OpenAIRE

    Zhang, Z.; Gerwin, Helmut; Scherer, Winfried

    1993-01-01

    In order to simulate the diffusion process during a hypothetical air ingress accident in a modular high temperature gas cooled reactor, a one-dimensional coupled diffusion-convection model has been established. In this analysis it is shown first, that experiments performed at the Japan Atomic Energy Research Institute (JAERI) have been recalculated successfully, thus validating the new model. Applying this model to the NACOK facility, now under construction at the Institute for Safety Researc...

  14. Coherent neutrino scattering with low temperature bolometers at Chooz reactor complex

    Science.gov (United States)

    Billard, J.; Carr, R.; Dawson, J.; Figueroa-Feliciano, E.; Formaggio, J. A.; Gascon, J.; Heine, S. T.; De Jesus, M.; Johnston, J.; Lasserre, T.; Leder, A.; Palladino, K. J.; Sibille, V.; Vivier, M.; Winslow, L.

    2017-10-01

    We present the potential sensitivity of a future recoil detector for a first detection of the process of coherent elastic neutrino nucleus scattering (CEνNS). We use the Chooz reactor complex in France as our luminous source of reactor neutrinos. Leveraging the ability to cleanly separate the rate correlated with the reactor thermal power against (uncorrelated) backgrounds, we show that a 10 kg cryogenic bolometric array with 100 eV threshold should be able to extract a CEνNS signal within one year of running.

  15. An attempt to minimize the temperature gradient along a plug-flow methane/steam reforming reactor by adopting locally controlled heating zones

    Science.gov (United States)

    Mozdzierz, M.; Brus, G.; Sciazko, A.; Komatsu, Y.; Kimijima, S.; Szmyd, J. S.

    2014-08-01

    Plug flow reactors are very common in the chemical process industry, including methane/steam reforming applications. Their operation presents many challenges, such as a strong dependence of temperature and composition distribution on the inlet conditions. The strongly endothermic methane/steam reforming reaction might result in a temperature drop at the inlet of the reactor and consequently the occurrence of large temperature gradients. The strongly non-uniform temperature distribution due to endothermic chemical reaction can have tremendous consequences on the operation of the reactor, such as catalyst degradation, undesired side reactions and thermal stresses. To avoid such unfavorable conditions, thermal management of the reactor becomes an important issue. To carry out thermal management properly, detailed modeling and corresponding numerical analyses of the phenomena occurring inside the reforming system is required. This paper presents experimental and numerical studies on the methane/steam reforming process inside a plug-flow reactor. To optimize the reforming reactors, detailed data about the entire reforming process is required. In this study the kinetics of methane/steam reforming on the Ni/YSZ catalyst was experimentally investigated. Measurements including different thermal boundary conditions, the fuel flow rate and the steam- to-methane ratios were performed. The reforming rate equation derived from experimental data was used in the numerical model to predict gas composition and temperature distribution along the steam-reforming reactor. Finally, an attempt was made to control the temperature distribution by adopting locally controlled heating zones.

  16. The research of the hot spot temperature of the dry-type reactor winding based on the inversed-heat conduction model

    Science.gov (United States)

    Lai, Wenqing; Luo, Hanwu; Li, Wenpeng; Cao, Yongdong; Ye, Ligang; Guo, Kai; Ding, Renjie; Wang, Yongqiang

    2017-10-01

    There is an important significance for the design and the life time evaluation of the dry-type reactor based on the hot-spot temperature. At present the methods of obtaining the reactor's hot-spot temperature are as follows: direct measurement and numerical calculation. Direct measurement can obtain winding hot spot temperature through the optical fiber temperature sensor or thermocouple; the numerical calculation mostly uses the finite element or finite difference method. This paper establishes the inversed-heat transfer calculation model of high and low voltage winding using the high-precision infrared sensor to acquire the temperature of the high voltage winding. Through calculation the temperature distribution of the low voltage winding has been obtained and the hottest spot temperature of the winding is much closed to the result obtained by the IEEE model in a certain range. It provides a new method for the acquisition of the hottest spot temperature of the dry-type reactor.

  17. Effect of temperature increase from 55 to 65 degrees C on performance and microbial population dynamics of an anaerobic reactor treating cattle manure

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær; Ibrahim, Ashraf; Mladenovska, Zuzana

    2001-01-01

    The effect of a temperature increase from 55 to 65 degreesC on process performance and microbial population dynamics were investigated in thermophilic, lab-scale, continuously stirred tank reactors. The reactors had a working volume of 3 l and were fed with cattle manure at an organic loading rat...

  18. Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

    Science.gov (United States)

    Fratoni, Massimiliano

    This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be

  19. Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2012-07-01

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

  20. Separation Requirements for a Hydrogen Production Plant and High-Temperature Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Scott Beck; Bill Galyean

    2005-09-01

    This report provides the methods, models, and results of an evaluation for locating a hydrogen production facility near a nuclear power plant. In order to answer the risk-related questions for this combined nuclear and chemical facility, we utilized standard probabilistic safety assessment methodologies to answer three questions: what can happen, how likely is it, and what are the consequences? As part of answering these questions, we developed a model suitable to determine separation distances for hydrogen process structures and the nuclear plant structures. Our objective of the model-development and analysis is to answer key safety questions related to the placement of one or more hydrogen production plants in the vicinity of a high-temperature nuclear reactor. From a thermal-hydraulic standpoint we would like the two facilities to be quite close. However, safety and regulatory implications force the separation distance to be increased, perhaps substantially. Without answering these safety questions, the likelihood for obtaining a permit to construct and build such as facility in the U.S. would be questionable. The quantitative analysis performed for this report provides us with a scoping mechanism to determine key parameters related to the development of a nuclear-based hydrogen production facility. From our calculations, we estimate that when the separation distance is less than 100m, the core damage frequency is large enough (greater than 1E-6/yr) to become problematic in a risk-informed environment. However, a variety of design modifications, for example blast-deflection barriers, were explored to determine the impact of potential mitigating strategies. We found that these mitigating cases may significantly reduce risk and should be explored as the design for the hydrogen production facility evolves.

  1. Femtosecond laser induced breakdown spectroscopy of silver within surrogate high temperature gas reactor fuel coated particles

    CSIR Research Space (South Africa)

    Roberts, DE

    2010-11-01

    Full Text Available The detection of metallic silver on Chemical Vapour Deposited (CVD) grown silicon carbide and in Pebble Bed Modular Reactor (PBMR) supplied tri-structural isotropic (TRISO) coated particles (with 500 µm diameter zirconium oxide surrogate kernel) has...

  2. Water Gas Shift Reaction with A Single Stage Low Temperature Membrane Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ciora, Richard J [Media and Process Technology Inc., Pittsburgh, PA (United States); Liu, Paul KT [Media and Process Technology Inc., Pittsburgh, PA (United States)

    2013-12-31

    Palladium membrane and Palladium membrane reactor were developed under this project for hydrogen separation and purification for fuel cell applications. A full-scale membrane reactor was designed, constructed and evaluated for the reformate produced from a commercial scale methanol reformer. In addition, the Pd membrane and module developed from this project was successfully evaluated in the field for hydrogen purification for commercial fuel cell applications.

  3. The Optimization of Radiation Protection in the Design of the High Temperature Reactor-Pebble-Bed Module

    Directory of Open Access Journals (Sweden)

    Sida Sun

    2017-01-01

    Full Text Available The optimization of radiation protection is an important task in both the design and operation of a nuclear power plant. Although this topic has been considerably investigated for pressurized water reactors, there are very few public reports on it for pebble-bed reactors. This paper proposes a routine that jointly optimizes the system design and radiation protection of High Temperature Reactor-Pebble-Bed Module (HTR-PM towards the As Low As Reasonably Achievable (ALARA principle. A systematic framework is also established for the optimization of radiation protection for pebble-bed reactors. Typical calculations for the radiation protection of radioactivity-related systems are presented to quantitatively evaluate the efficiency of the optimization routine, which achieve 23.3%~90.6% reduction of either dose rate or shielding or both of them. The annual collective doses of different systems are reduced through iterative optimization of the dose rates, designs, maintenance procedures, and work durations and compared against the previous estimates. The comparison demonstrates that the annual collective dose of HTR-PM is reduced from 0.490 man-Sv/a before optimization to 0.445 man-Sv/a after optimization, which complies with the requirements of the Chinese regulatory guide and proves the effectiveness of the proposed routine and framework.

  4. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  5. Technology Development Roadmap for the Advanced High Temperature Reactor Secondary Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    P. Sabharwall; M. McCllar; A. Siahpush; D. Clark; M. Patterson; J. Collins

    2012-09-01

    This Technology Development Roadmap (TDRM) presents the path forward for deploying large-scale molten salt secondary heat exchangers (MS-SHX) and recognizing the benefits of using molten salt as the heat transport medium for advanced high temperature reactors (AHTR). This TDRM will aid in the development and selection of the required heat exchanger for: power production (the first anticipated process heat application), hydrogen production, steam methane reforming, methanol to gasoline production, or ammonia production. This TDRM (a) establishes the current state of molten salt SHX technology readiness, (b) defines a path forward that systematically and effectively tests this technology to overcome areas of uncertainty, (c) demonstrates the achievement of an appropriate level of maturity prior to construction and plant operation, and (d) identifies issues and prioritizes future work for maturing the state of SHX technology. This study discusses the results of a preliminary design analysis of the SHX and explains the evaluation and selection methodology. An important engineering challenge will be to prevent the molten salt from freezing during normal and off-normal operations because of its high melting temperature (390°C for KF ZrF4). The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The need for efficiency, compactness, and safety challenge the capabilities of existing heat exchanger technology. The description of potential heat exchanger configurations or designs (such as printed circuit, spiral or helical coiled, ceramic, plate and fin, and plate type) were covered in an earlier report (Sabharwall et al. 2011). Significant future work, much of which is suggested in this report, is needed before the benefits and full potential of the AHTR can be realized. The execution of this TDRM will focuses

  6. Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O2 cannot be ignored, especially for the FHR, in which environment the product, SiO2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

  7. Simulation of a low temperature water gas shift reactor using the heterogeneous model/application to a pem fuel cell

    Science.gov (United States)

    Giunta, Pablo; Amadeo, Norma; Laborde, Miguel

    In the last few years, a renewed interest in the water gas shift (WGS) reaction at low temperature has arisen due to its application to fuel cells. In this work, a simulation of a fixed bed reactor for this reaction, which forms part of a hydrogen production-purification train for a 10 kW PEM fuel cell using ethanol as the raw material, was carried out. A commercial Cu/Zn/Ba/Al 2O 3 catalyst was employed and a one-dimensional heterogeneous model was applied for the simulation. The catalyst deactivation due to thermal factors (sintering) was taken into account in the model. Isothermal and adiabatic regimes were analyzed as well. Results of the simulation indicate that the pellet can be considered isothermal but temperature gradients in the film cannot be disregarded. On the other hand, concentration gradients in the film can be ignored but CO profiles are established inside the pellet. Adiabatic operation can be recommended because of its simplicity of operation and construction. The reactor volume is strongly sensitive to the CO outlet concentration at CO levels lower than 6000 ppm. For a 10 kW PEM fuel cell, using adequate pellet size and taking into account the catalyst deactivation, a reactor volume of 0.64 l would be enough to obtain an outlet CO concentration of about 7160 ppm. This concentration value can be handled by the next purification stage, COPROX.

  8. Coupling of high temperature nuclear reactor with chemical plant by means of steam loop with heat pump

    Directory of Open Access Journals (Sweden)

    Kopeć Mariusz

    2017-01-01

    Full Text Available High temperature nuclear reactors (HTR can be used as an excellent, emission-free source of technological heat for various industrial applications. Their outlet helium temperature (700°-900°C allows not only for heat supply to all processes below 600°C (referred to as “steam class”, but also enables development of clean nuclear-assisted hydrogen production or coal liquefaction technologies with required temperatures up to 900°C (referred to as “chemical class”. This paper presents the results of analyses done for various configurations of the steam transport loop coupled with the high-temperature heat pump designed for “chemical class” applications. The advantages and disadvantages as well as the key issues are discussed in comparison with alternative solutions, trying to answer the question whether the system with the steam loop and the hightemperature heat pump is viable and economically justified.

  9. Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

    Science.gov (United States)

    Kasahara, Shigeki; Kitsunai, Yuji; Chimi, Yasuhiro; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-11-01

    This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. One of the temperature profiles was that the specimens experienced neutron irradiation in both reactors, under which the irradiation temperature transiently increased to 290 °C from room temperature with increasing reactor power during reactor startup periods. Another was that the specimens were pre-heated to about 150 °C prior to the irradiation to suppress the transient temperature increase. Tensile tests at 290 °C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Difference of the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. Although influence of neutron irradiation involving transient temperature increase to 290 °C from room temperature on the yield strength and the Vickers hardness is buried in the trend curves of existing data, the influence was also found certainly in increment of in yield strength, existence of modest yield drop, and loss of strain hardening capacity and ductility. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, seemed to have important implications regarding the interpretation of not irradiation hardening, but deformation of the austenitic stainless steel.

  10. OPTIMASI GEOMETRI TERAS REAKTOR DAN KOMPOSISI BAHAN BAKAR BERBENTUK BOLA PADA DESAIN HIGH TEMPERATURE FAST REACTOR (HTFR

    Directory of Open Access Journals (Sweden)

    Agustina Mega

    2015-04-01

    Full Text Available Telah dilakukan desain High Temperature Fast Reactor (HTFR tipe pebble dengan bahan bakar uranium plutonium nitrida berpendingin Pb-Bi. Parameter yang dianalisis adalah kritikalitas teras, koefisien reaktivitas temperatur bahan bakar, koefisien reaktivitas void pendingin dan kemampuan breeding reaktor. Perhitungan dilakukan dengan paket program SRAC2K3. Dari penelitian ini diharapkan diperoleh desain teras berumur lama dan memiliki fitur keselamatan melekat. Dari penelitian ini diperoleh desain reaktor dengan diameter 520 cm dan tinggi 480 cm. Bahan bakar berbentuk pebble dengan 63 % UN-37 % PuN pada zona core dan 95,5 % UN-4,5 % PuN pada zona blanket. Reaktor tidak kritis setelah kurang lebih 800 hari dan keff pada BoL 1,078223 dan keff setelah 800 hari adalah 0,986379. Dari penelitian ini diperoleh koefisien reaktivitas temperatur bahan bakar sebesar -2,190014E-05 pada saat BoL dan -1,390773E-05 setelah 800 hari serta koefisien reaktivitas void pendingin sebesar -2,160402E-04/% void pada saat BoL dan setelah 800 hari sebesar -2,942364E-03/% void. Reaktor merupakan jenis fast breeder ditandai dengan naiknya densitas plutonium 239. Kata kunci : desain, teras, HTFR, keselamatan, umur, koefisien reaktivitas.   Design of pebble bed type High Temperature Fast Reactor (HTFR with uranium plutonium nitride fuel and Pb-Bi cooled has been done. The parameters being analyzed were core criticality, fuel temperature coefficient, void coefficient and reactor breeding ability. Calculation was done by using SRAC2K3 computer code. This research is expected to obtaine the design with long life core and inherent safety features. This research obtained core design with a diameter of 520 cm and 480 cm core high. Shaped pebble fuel bed with the 63 % UN-37 % PUN on core zone and 95.5 % UN-4.5 % Pu on blanket zone and keff value is 1.078223 with approximately 800 day of core life. The fuel temperature coefficient is -2.190014E-05 at BOL and is 1.390773E-05 at EOL and

  11. Thermal decomposition of expanded polystyrene in a pebble bed reactor to get higher liquid fraction yield at low temperatures.

    Science.gov (United States)

    Chauhan, R S; Gopinath, S; Razdan, P; Delattre, C; Nirmala, G S; Natarajan, R

    2008-11-01

    Expanded polystyrene is one of the polymers produced in large quantities due to its versatile application in different fields. This polymer is one of the most intractable components in municipal solid waste. Disposal of polymeric material by pyrolysis or catalytic cracking yields valuable hydrocarbon fuels or monomers. Literature reports different types of reactors and arrangements that have uniform temperatures during pyrolysis and catalytic cracking. The present study focuses on reducing the temperature to maximize the quantity of styrene monomer in the liquid product. A bench scale reactor has been developed to recover the styrene monomer and other valuable chemicals. Experiments were carried under partial oxidation and vacuum conditions in the temperature range of 300-500 degrees C. In the pyrolysis optimization studies, the best atmospheric condition was determined to be vacuum, the pyrolysis temperature should be 500 degrees C, yield of liquid product obtained was 91.7% and yield of styrene obtained was 85.5%. In the characterization studies, distillation and IR spectroscopy experiments were carried out. The remaining of the liquid product comprises of benzene, ethyl benzene, and styrene dimers and trimers.

  12. Airlift-reactor design and test for aerobic environmental bioprocesses with extremely high solid contents at high temperatures.

    Science.gov (United States)

    Feitkenhauer, H; Maleski, R; Märkl, H

    2003-01-01

    Bioprocesses at high temperatures gained considerably in importance within the last years and several new applications for aerobic, extreme thermophilic environmental bioprocesses are emerging. However, this development is not yet matched by adequate bioreactor designs, especially if it comes to the treatment of solids. In this communication we propose the use of airlift reactors to bridge this gap. The design of an internal draught tube bioreactor (Area(Riser)/Area(Downcomer) . 1; Height/Diameter . 8) is described in detail. The influence of the temperature on gas hold-up, liquid velocity and mixing characteristics was investigated. It was shown that this reactor could hold up to 1 t quartz sand per m3 in suspension at moderate aeration rates. Despite the decreasing oxygen solubility, the oxygen transfer rate increased with rising temperature due to the improved mass transfer parameters. With rising solid content, the oxygen transfer rate increased and reached a maximum at a solid content of about 140 kg m(-3) before it decreased again. However, it is only slightly reduced at the highest solid contents. The results demonstrate that aerobic bioprocesses at high temperatures are not only feasible, but can be very efficient if carried out in proper bioreactors.

  13. Rapid formation and pollutant removal ability of aerobic granules in a sequencing batch airlift reactor at low temperature.

    Science.gov (United States)

    Jiang, Yu; Shang, Yu; Wang, Hongyu; Yang, Kai

    2016-12-01

    The start-up of an aerobic granular sludge (AGS) reactor at low temperature was more difficult than at ambient temperature.The rapid formation and characteristics of AGS in a sequencing batch airlift reactor at low temperature were investigated. The nutrient removal ability of the system was also evaluated. It was found that compact granules with clear boundary were formed within 10 days and steady state was achieved within 25 days. The settling properties of sludge were improved with the increasing secretion of extracellular polymeric substances and removal performances of pollutants were enhanced along with granulation. The average removal efficiencies of COD, NH4(+)-N, total nitrogen (TN), total phosphorus (TP) after aerobic granules maturing were over 90.9%, 94.7%, 75.4%, 80.2%, respectively. The rise of temperature had little impact on pollutant biodegradation while the variation of dissolved oxygen caused obvious changes in TN and TP removal rates. COD concentrations of effluents were below 30 mg l(-1) in most cycles of operation with a wide range of organic loading rates (0.6-3.0 kg COD m(-3) d(-1)). The rapid granulation and good performance of pollutant reduction by the system might provide an alternate for wastewater treatment in cold regions.

  14. Safeguards-by-Design: Guidance for High Temperature Gas Reactors (HTGRs) With Pebble Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Philip Casey Durst; Mark Schanfein

    2012-08-01

    The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on pebble fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information

  15. Research of the influence of intensification of heat transfer on distribution of temperature in the active core of the gas cooled nuclear reactor of the «GT-MHR» project

    Science.gov (United States)

    Kuzevanov, V. S.; Podgorny, S. K.

    2017-11-01

    The maximum wall temperature of a cooling channel of a nuclear reactor is one of the factors that affects directly of the safety and reliability of the nuclear reactor. In this paper suggested an equation, which allows calculating the maximum wall temperature of the cooling channel of the nuclear reactor with heat transfer enhancer installed, without enormous calculations.

  16. Design and development of a low-temperature reactor system for biorefining waste oil

    OpenAIRE

    Pedersen, Hans Olav

    2017-01-01

    The background for this master’s thesis is the focus on bioenergy and biofuels at NMBU. This has, among others, resulted in a prototype of a small-scale biorefinery, which uses methanol and waste cooking oil to produce biodiesel. The purpose of this thesis is to develop a reactor system that serves as a platform for reactors to operate on and a technological alternative for a periodically on-site clean of catalysts. The purpose of the catalysts wash is to extend their lifetime, in order to ma...

  17. Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

    2012-07-15

    Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

  18. The energy release and temperature field in the ultracold neutron source of the WWR-M reactor at the Petersburg Nuclear Physics Institute

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Kislitsin, B. V.; Onegin, M. S.; Lyamkin, V. A.; Prudnikov, D. V.; Ilatovskiy, V. A.; Orlov, S. P.; Kirsanov, G. A.; Fomin, A. K.; Filchenkova, D. V. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation)

    2016-12-15

    Results of calculations of energy releases and temperature fields in the ultracold neutron source under design at the WWR-M reactor are presented. It is shown that, with the reactor power of 18 MW, the power of energy release in the 40-L volume of the source with superfluid helium will amount to 28.5 W, while 356 W will be released in a liquid-deuterium premoderator. The lead shield between the reactor core and the source reduces the radiative heat release by an order of magnitude. A thermal power of 22 kW is released in it, which is removed by passage of water. The distribution of temperatures in all components of the vacuum structure is presented, and the temperature does not exceed 100°C at full reactor power. The calculations performed make it possible to go to design of the source.

  19. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

  20. Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium Environments

    Energy Technology Data Exchange (ETDEWEB)

    Crone, Wendy; Cao, Guoping; Sridhara, Kumar

    2013-05-31

    The helium coolant in high-temperature reactors inevitably contains low levels of impurities during steady-state operation, primarily consisting of small amounts of H{sub 2}, H{sub 2}O, CH{sub 4}, CO, CO{sub 2}, and N{sub 2} from a variety of sources in the reactor circuit. These impurities are problematic because they can cause significant long-term corrosion in the structural alloys used in the heat exchangers at elevated temperatures. Currently, the primary candidate materials for intermediate heat exchangers are Alloy 617, Haynes 230, Alloy 800H, and Hastelloy X. This project will evaluate the role of impurities in helium coolant on the stress-assisted grain boundary oxidation and creep crack growth in candidate alloys at elevated temperatures. The project team will: • Evaluate stress-assisted grain boundary oxidation and creep crack initiation and crack growth in the temperature range of 500-850°C in a prototypical helium environment. • Evaluate the effects of oxygen partial pressure on stress-assisted grain boundary oxidation and creep crack growth in impure helium at 500°C, 700°C, and 850°C respectively. • Characterize the microstructure of candidate alloys after long-term exposure to an impure helium environment in order to understand the correlation between stress-assisted grain boundary oxidation, creep crack growth, material composition, and impurities in the helium coolant. • Evaluate grain boundary engineering as a method to mitigate stress-assisted grain boundary oxidation and creep crack growth of candidate alloys in impure helium. The maximum primary helium coolant temperature in the high-temperature reactor is expected to be 850-1,000°C.Corrosion may involve oxidation, carburization, or decarburization mechanisms depending on the temperature, oxygen partial pressure, carbon activity, and alloy composition. These corrosion reactions can substantially affect long-term mechanical properties such as crack- growth rate and fracture

  1. Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database: 1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 G...

  2. Evaluation of the performance of high temperature conversion reactors for compound-specific oxygen stable isotope analysis.

    Science.gov (United States)

    Hitzfeld, Kristina L; Gehre, Matthias; Richnow, Hans-Hermann

    2017-05-01

    In this study conversion conditions for oxygen gas chromatography high temperature conversion (HTC) isotope ratio mass spectrometry (IRMS) are characterised using qualitative mass spectrometry (IonTrap). It is shown that physical and chemical properties of a given reactor design impact HTC and thus the ability to accurately measure oxygen isotope ratios. Commercially available and custom-built tube-in-tube reactors were used to elucidate (i) by-product formation (carbon dioxide, water, small organic molecules), (ii) 2nd sources of oxygen (leakage, metal oxides, ceramic material), and (iii) required reactor conditions (conditioning, reduction, stability). The suitability of the available HTC approach for compound-specific isotope analysis of oxygen in volatile organic molecules like methyl tert-butyl ether is assessed. Main problems impeding accurate analysis are non-quantitative HTC and significant carbon dioxide by-product formation. An evaluation strategy combining mass spectrometric analysis of HTC products and IRMS (18)O/(16)O monitoring for future method development is proposed.

  3. NUCLEAR REACTOR

    Science.gov (United States)

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  4. Quantitative Homogeneity and In-Contact Particles of High Temperature Reactors (htr) Compacts Determination via X-Ray Tomography

    Science.gov (United States)

    Lecomte, G.; Tisseur, D.; Létang, J. M.; Banchet, J.; Vitali, M. P.

    2008-02-01

    In AREVA Nuclear Power's High Temperature Reactor (HTR) design called ANTARES, fuel consists of compacts composed of few thousands millimetric quasi-spherical particles dispersed in a graphite matrix. Compact homogeneity, defined as the homogeneous particles spatial distribution in the matrix, as well as the possibility of obtaining particles in contact, need to be assessed since they condition the thermo-mechanical behavior of the nuclear fuel under irradiation. In this paper, image and data processing algorithms are developed to do so, based on X-Ray tomographic images.

  5. The preliminary design of bearings for the control system of a high-temperature lithium-cooled nuclear reactor

    Science.gov (United States)

    Yacobucci, H. G.; Waldron, W. D.; Walowit, J. A.

    1973-01-01

    The design of bearings for the control system of a fast reactor concept is presented. The bearings are required to operate at temperatures up to 2200 F in one of two fluids, lithium or argon. Basic bearing types are the same regardless of the fluid. Crowned cylindrical journals were selected for radially loaded bearings and modified spherical bearings were selected for bearings under combined thrust and radial loads. Graphite and aluminum oxide are the materials selected for the argon atmosphere bearings while cermet compositions (carbides or nitrides bonded with refractory metals) were selected for the lithium lubricated bearings. Mounting of components is by shrink fit or by axial clamping utilizing differential thermal expansion.

  6. High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Cole, T.E.; Sanders, J.P.; Kasten, P.R.

    1977-07-01

    Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies.

  7. Synthesis of Fused Pyrimidinone and Quinolone Derivatives in an Automated High-Temperature and High-Pressure Flow Reactor.

    Science.gov (United States)

    Tsoung, Jennifer; Bogdan, Andrew R; Kantor, Stanislaw; Wang, Ying; Charaschanya, Manwika; Djuric, Stevan W

    2017-01-20

    Fused pyrimidinone and quinolone derivatives that are of potential interest to pharmaceutical research were synthesized within minutes in up to 96% yield in an automated Phoenix high-temperature and high-pressure continuous flow reactor. Heterocyclic scaffolds that are either hard to synthesize or require multisteps are readily accessible using a common set of reaction conditions. The use of low-boiling solvents along with the high conversions of these reactions allowed for facile workup and isolation. The methods reported herein are highly amenable for fast and efficient heterocycle synthesis as well as compound scale-ups.

  8. Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-11-01

    Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

  9. Multi-scale modelling to relate beryllium surface temperature, deuterium concentration and erosion in fusion reactor environment

    Science.gov (United States)

    Safi, E.; Valles, G.; Lasa, A.; Nordlund, K.

    2017-05-01

    Beryllium (Be) has been chosen as the plasma-facing material for the main wall of ITER, the next generation fusion reactor. Identifying the key parameters that determine Be erosion under reactor relevant conditions is vital to predict the ITER plasma-facing component lifetime and viability. To date, a certain prediction of Be erosion, focusing on the effect of two such parameters, surface temperature and D surface content, has not been achieved. In this work, we develop the first multi-scale KMC-MD modeling approach for Be to provide a more accurate database for its erosion, as well as investigating parameters that affect erosion. First, we calculate the complex relationship between surface temperature and D concentration precisely by simulating the time evolution of the system using an object kinetic Monte Carlo (OKMC) technique. These simulations provide a D surface concentration profile for any surface temperature and incoming D energy. We then describe how this profile can be implemented as a starting configuration in molecular dynamics (MD) simulations. We finally use MD simulations to investigate the effect of temperature (300-800 K) and impact energy (10-200 eV) on the erosion of Be due to D plasma irradiations. The results reveal a strong dependency of the D surface content on temperature. Increasing the surface temperature leads to a lower D concentration at the surface, because of the tendency of D atoms to avoid being accommodated in a vacancy, and de-trapping from impurity sites diffuse fast toward bulk. At the next step, total and molecular Be erosion yields due to D irradiations are analyzed using MD simulations. The results show a strong dependency of erosion yields on surface temperature and incoming ion energy. The total Be erosion yield increases with temperature for impact energies up to 100 eV. However, increasing temperature and impact energy results in a lower fraction of Be atoms being sputtered as BeD molecules due to the lower D

  10. Computational model for a high temperature electrolyzer coupled to a HTTR for efficient nuclear hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, Daniel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy; Gamez, Abel; Brayner, Carlos, E-mail: danielgonro@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Garcia, Lazaro; Garcia, Carlos; Torre, Raciel de la, E-mail: lgarcia@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Sanchez, Danny [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)

    2015-07-01

    High temperature electrolysis process coupled to a very high temperature reactor (VHTR) is one of the most promising methods for hydrogen production using a nuclear reactor as the primary heat source. However there are not references in the scientific publications of a test facility that allow to evaluate the efficiency of the process and other physical parameters that has to be taken into consideration for its accurate application in the hydrogen economy as a massive production method. For this lack of experimental facilities, mathematical models are one of the most used tools to study this process and theirs flowsheets, in which the electrolyzer is the most important component because of its complexity and importance in the process. A computational fluid dynamic (CFD) model for the evaluation and optimization of the electrolyzer of a high temperature electrolysis hydrogen production process flowsheet was developed using ANSYS FLUENT®. Electrolyzer's operational and design parameters will be optimized in order to obtain the maximum hydrogen production and the higher efficiency in the module. This optimized model of the electrolyzer will be incorporated to a chemical process simulation (CPS) code to study the overall high temperature flowsheet coupled to a high temperature accelerator driven system (ADS) that offers advantages in the transmutation of the spent fuel. (author)

  11. Prediction of the moderator temperature field in a heavy water reactor based on a cellular neural network

    Directory of Open Access Journals (Sweden)

    S.O. Starkov

    2017-06-01

    Full Text Available Reactors with heavy water coolants and moderators have been used extensively in today's power industry. Monitoring of the moderator condition plays an important role in ensuring normal operation of a power plant. A cellular neural network, the architecture of which has been adapted for hardware implementation, is proposed for use in a system for prediction of the heavy water moderator temperature. A reactor model composed in accordance with the CANDU Darlington heavy water reactor design was used to form the training sample collection and to control correct operation of the neural network structure. The sample components for the adjustment and configuration of the network topology include key parameters that characterize the energy generation process in the core. The paper considers the feasibility of the temperature prediction only for the calandria's central cross-section. To solve this problem, the cellular neural network architecture has been designed, and major parts of the digital computational element and methods for their implementation based on an FPLD have also been developed. The method is described for organizing an optical coupling between individual neural modules within the network, which enables not only the restructuring of the topology in the training process, but also the assignment of priorities for the propagation of the information signals of neurons depending on the activity in a situation analysis at the neural network structure inlet. Asynchronous activation of cells was used based on an oscillating fractal network, the basis for which was a modified ring oscillator. The efficiency of training the proposed architecture using stochastic diffusion search algorithms is evaluated. A comparative analysis of the model behavior and the results of the neural network operation have shown that the use of the neural network approach is effective in safety systems of power plants.

  12. Start-up of UASB reactors treating municipal wastewater and effect of temperature/sludge age and hydraulic retention time (HRT on its performance

    Directory of Open Access Journals (Sweden)

    Hina Rizvi

    2015-11-01

    Full Text Available The upflow anaerobic sludge blanket reactors seeded with cow dung manure (UASBCD and activated sludge of a dairy wastewater treatment plant (UASBASDIT were used to treat raw domestic wastewater of medium strength. The UASBCD reactor required a period of 120 days to start up. In case of UASBASDIT reactor, sludge bed was stabilized in a period of 80 days. The performance of both reactors to treat wastewater was enhanced with an increase in the sludge age and temperature. Under psychrophilic temperature (17 °C and at early sludge age (60 days, chemical oxygen demand (COD and biochemical oxygen demand (BOD removal by the reactors were in the range of 57–62% and 61–66%, respectively. However, chemical oxygen demand (COD and biochemical oxygen demand (BOD removal efficiency of the reactors elevated to the range of 79–81% and 77–83%, respectively at sludge age of 150 days and temperature of 30 °C. In short, overall performance of both reactors was optimum at sludge age ranging from 120 to 150 days and temperature varying between 25 and 30 °C. At hydraulic retention time (HRT of 9 h the chemical oxygen demand (COD, total suspended solids (TSS and sulfate removal efficiency of UASBCD reactor reached 81%, 75% and 76%, respectively and 77%, 74% and 69%, respectively for UASBASDIT. The rate of removal of these parameters however, gradually declined with increasing hydraulic retention time. The UASB technology provides a low-cost system for the direct treatment of municipal wastewater and can be applied in small communities.

  13. High temperature fast reactor for hydrogen production in Brazil; Reator nuclear rapido de altissima temperatura para producao de hidrogenio no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados]. E-mail: jamil@ieav.cta.br

    2008-07-01

    The main nuclear reactors technology for the Generation IV, on development phase for utilization after 2030, is the fast reactor type with high temperature output to improve the efficiency of the thermo-electric conversion process and to enable applications of the generated heat in industrial process. Currently, water electrolysis and thermo chemical cycles using very high temperature are studied for large scale and long-term hydrogen production, in the future. With the possible oil scarcity and price rise, and the global warming, this application can play an important role in the changes of the world energy matrix. In this context, it is proposed a fast reactor with very high output temperature, {approx} 1000 deg C. This reactor will have a closed fuel cycle; it will be cooled by lead and loaded with nitride fuel. This reactor may be used for hydrogen, heat and electricity production in Brazil. It is discussed a development strategy of the necessary technologies and some important problems are commented. The proposed concept presents characteristics that meet the requirements of the Generation IV reactor class. (author)

  14. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  15. Creep Crack Growth Behavior of Alloys 617 and 800H in Air and Impure Helium Environments at High Temperatures

    Science.gov (United States)

    Grierson, D. S.; Cao, G.; Brooks, P.; Pezzi, P.; Glaudell, A.; Kuettel, D.; Fischer, G.; Allen, T.; Sridharan, K.; Crone, W. C.

    2017-03-01

    The environmental degradation of intermediate heat exchanger (IHX) materials in impure helium has been identified as an area with major ramifications on the design of very high-temperature reactors (VHTR). It has been reported that in some helium environments, non-ductile failure is a significant failure mode for Alloy 617 with long-term elevated-temperature service. Non-ductile failure of intermediate exchangers can result in catastrophic consequences; unfortunately, the knowledge of creep crack initiation and creep crack growth (CCG) in candidate alloys is limited. Current codes and code cases for the candidate alloys do not provide specific guidelines for effects of impure helium on the high-temperature behavior. The work reported here explores creep crack growth characterization of Alloy 617 and Alloy 800H at elevated temperatures in air and in impure helium environments, providing information on the reliability of these alloys in VHTR for long-term service. Alloy 617 was found to exhibit superior CCG resistance compared to Alloy 800H. For Alloy 617 tested at 973 K (700 °C), a notable increase in the resistance to crack growth was measured in air compared to that measured in the helium environment; CCG results for Alloy 800H suggest that air and helium environments produce similar behavior. Testing of grain boundary-engineered (GBE) Alloy 617 samples revealed that, although the technique produces superior mechanical properties in many respects, the GBE samples exhibited inferior resistance to creep crack growth compared to the other Alloy 617 samples tested under similar conditions. Grain size is noted as a confounding factor in creep crack growth resistance.

  16. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  17. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  18. Evaluation of two processes of hydrogen production starting from energy generated by high temperature nuclear reactors; Evaluacion de dos procesos de produccion de hidrogeno a partir de energia generada por reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J., E-mail: jvalle@upmh.edu.mx [Universidad Politecnica Metropolitana de Hidalgo, Boulevard Acceso a Tolcayuca 1009, Ex-Hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2013-10-15

    In this work an evaluation to two processes of hydrogen production using energy generated starting from high temperature nuclear reactors (HTGR's) was realized. The evaluated processes are the electrolysis of high temperature and the thermo-chemistry cycle Iodine-Sulfur. The electrolysis of high temperature, contrary to the conventional electrolysis, allows reaching efficiencies of up to 60% because when increasing the temperature of the water, giving thermal energy, diminishes the electric power demand required to separate the molecule of the water. However, to obtain these efficiencies is necessary to have water vapor overheated to more than 850 grades C, temperatures that can be reached by the HTGR. On the other hand the thermo-chemistry cycle Iodine-Sulfur, developed by General Atomics in the 1970 decade, requires two thermal levels basically, the great of them to 850 grades C for decomposition of H{sub 2}SO{sub 4} and another minor to 360 grades C approximately for decomposition of H I, a high temperature nuclear reactor can give the thermal energy required for the process whose products would be only hydrogen and oxygen. In this work these two processes are described, complete models are developed and analyzed thermodynamically that allow to couple each hydrogen generation process to a reactor HTGR that will be implemented later on for their dynamic simulation. The obtained results are presented in form of comparative data table of each process, and with them the obtained net efficiencies. (author)

  19. Application of techniques of dynamic reliability to the assessment of safety a high-temperature Nuclear reactor; Aplicacion de Tecnicas de Fiabilidad Dinamica a la Evaluacion de Seguridad de un Reactor Nuclear de Alta Temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Flores, A.; Gallego Diaz, E.

    2011-07-01

    The main objective of this work is to describe the application of the methodology of integrated analysis of safety to safety assessment of High Temperature Engineering Test Reactor (HTTR), as a demonstration of its ability to be applied to technologies other than light-water reactors, which was initially conceived. The practical application of the method in the case of the HTTR has required the development of a basic model of the HTTR, called DD-HTTR5+, that it allows to represent in a way joint dynamics of the plant and its characteristics of reliability, as well as existing interactions.

  20. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Directory of Open Access Journals (Sweden)

    Jeong Soon Park

    2016-04-01

    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  1. Intermediate temperature embrittlement of one new Ni-26W-6Cr based superalloy for molten salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Li [Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Science, Beijing 100049 (China); Ye, Xiangxi [University of Chinese Academy of Science, Beijing 100049 (China); Cui, Chuanyong [Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Huang, Hefei; Leng, Bin [Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Li, Zhijun, E-mail: lizhijun@sinap.ac.cn [Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Zhou, Xingtai [Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-06-21

    Ni-26W-6Cr based superalloy is considered a potential structure material for the molten salt reactors due to its high strength and good compatibility with the fluoride salt. In the present work, the temperature dependence of the tensile behavior of the alloy was studied by tensile tests in the temperature range of 25–850 °C. This alloy exhibited a good ductility at RT and 450 °C, a ductility minimum from 650 to 750 °C and an intermediate ductility at 850 °C. TEM and EBSD characterization was performed on specimens tested at three typical temperature points (RT, 650 °C and 850 °C) to determine the deformation and fracture mechanisms accounting for the intermediate temperature embrittlement. At RT, the grain boundaries can accommodate enough dislocations to provide compatibility of the sliding between adjacent grains, then M{sub 6}C carbides act as crack origins and cause the fracture. In case of 650 °C, the grain boundaries cannot withstand the local stress even if only a small number of dislocation pile-ups exist. The premature cracks at grain boundaries impede the development of plastic deformation from single slips to multiple ones and cause the low ductility. If tested at 850 °C, the fracture process is retarded by the dynamic recovery and local dynamic recrystallization at crack tips.

  2. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is

  3. Specifics of high-temperature sodium coolant purification technology in fast reactors for hydrogen production and other innovative applications

    Directory of Open Access Journals (Sweden)

    F.A. Kozlov

    2017-03-01

    Full Text Available In creating a large-scale atomic-hydrogen power industry, the resolution of technological issues associated with high temperatures in reactor plants (900°C and large hydrogen concentrations intended as long-term resources takes on particular importance. The paper considers technological aspects of removing impurities from high-temperature sodium used as a coolant in the high-temperature fast reactor (BN-HT 600MW (th. intended for the production of hydrogen as well as other innovative applications. The authors examine the behavior of impurities in the BN-HT circuits associated with the mass transfer intensification at high temperatures (Arrhenius law in different operating modes. Special attention is given to sodium purification from hydrogen, tritium and corrosion products in the BN-HT. Sodium purification from hydrogen and tritium by their evacuation through vanadium or niobium membranes will make it possible to develop compact highly-efficient sodium purification systems. It has been shown that sodium purification from tritium to concentrations providing the maximum permissible concentration of the produced hydrogen (3.6Bq/l according to NRB-99/2009 specifies more stringent requirements to the hydrogen removal system, i.e., the permeability index of the secondary tritium removal system should exceed 140kg/s. Provided that a BN-HN-type reactor meets these conditions, the bulk of tritium (98% will be accumulated in the compact sodium purification system of the secondary circuit, 0.6% (∼ 4·104Bq/s, will be released into the environment and 1.3% will enter the product (hydrogen. The intensity of corrosion products (CPs coming into sodium is determined by the corrosion rate of structural materials: at a high temperature level, a significant amount of corrosion products flows into sodium. The performed calculations showed that, for the primary BN-HT circuit, the amount of corrosion products formed at the oxygen concentration in sodium of 1mln

  4. An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-02-01

    Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

  5. Adaptation of a High Frequency Ultrasonic Transducer to the Measurement of Water Temperature in a Nuclear Reactor

    Science.gov (United States)

    Zaz, G.; Calzavara, Y.; Le Clézio, E.; Despaux, G.

    Most high flux reactors possess for research purposes fuel elements composed of plates. Their relative distance is a crucial parameter, particularly concerning the irradiation history. For the High Flux Reactor (RHF) of the Institute Laue-Langevin (ILL), the measurement of this distance with a microscopic resolution becomes extremely challenging. To address this issue, a specific ultrasonic transducer, presented in a first paper, has been designed and manufactured to be inserted into the 1.8 mm width channel existing between curved fuel plates. It was set on a blade yielding a total device thickness of 1 mm. To achieve the expected resolution, the system is excited with frequencies up to 70 MHz and integrated into a set of high frequency acquisition instruments. Thanks to a specific signal processing, this device allows the distance measurement through the evaluation of the ultrasonic wave time of fight. One of the crucial points is then the evaluation of the local water temperature inside the water channel. To obtain a precise estimation of this parameter, the ultrasonic sensor is used as a thermometer thanks to the analysis of the spectral components of the acoustic signal propagating inside the sensor multilayered structure. The feasibility of distance measurement was proved during the December 2013 experiment in the RHF fuel element of the ILL. Some of the results will be presented as well as some experimental constraints identified to improve the accuracy of the measurement in future works.

  6. Synthetic fuel production via carbon neutral cycles with high temperature nuclear reactors as a power source

    Energy Technology Data Exchange (ETDEWEB)

    Konarek, E.; Coulas, B.; Sarvinis, J. [Hatch Ltd., Mississauga, Ontario (Canada)

    2016-06-15

    This paper analyzes a number of carbon neutral cycles, which could be used to produce synthetic hydrocarbon fuels. Synthetic hydrocarbons are produced via the synthesis of Carbon Monoxide and Hydrogen. The . cycles considered will either utilize Gasification processes, or carbon capture as a source of feed material. In addition the cycles will be coupled to a small modular Nuclear Reactor (SMR) as a power and heat source. The goal of this analysis is to reduce or eliminate the need to transport diesel and other fossil fuels to remote regions and to provide a carbon neutral, locally produced hydrocarbon fuel for remote communities. The technical advantages as well as the economic case are discussed for each of the cycles presented. (author)

  7. Computational Fluid Dynamic Analysis for the Proposed VHTR Lower Plenum Standard Problem

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, R. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pointer, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2008-09-01

    One of the reference designs for the Next Generation Nuclear Plant (NGNP) is a prismatic reactor configuration. A major concern in the operation of this design is the nature of the flow of hot gas as it exits the core into the lower plenum and then out the vessel exit duct. The two primary areas of concern are the impingement of hot jets onto the walls of the lower plenum and the degree of thermal mixing of the helium coolant as it flows out of the reactor vessel. Computational fluid dynamic (CFD) simulations have been proposed by general consensus to estimate the desired local fluid dynamic and temperature information. Successful CFD simulations require careful planning and extensive calculations to help ensure that they accurately represent the flow physics. An important aspect of the application of CFD is to validate the results against detailed validation data, for scaled problems that involve similar physics. A section of the lower plenum of the reference prismatic NGNP was employed to create a scaled model to use to obtain experimental data for use as validation data to compare with a CFD analysis. Extensive flow data were obtained from the scaled model while installed in the INL’s Matched Index of Refraction (MIR) test facility.

  8. Rate enhancement in microfabricated chemical reactors under fast forced temperature oscillations

    DEFF Research Database (Denmark)

    Hansen, Heine Anton; Olsen, Jakob L.; Jensen, Søren

    2006-01-01

    Oxidation of CO under fast forced temperature oscillations shows increased reaction rate compared to steady state. A maximum increase of 40% is observed relative to steady state. The reaction rate is investigated for varying mean temperature, amplitude and frequency. As function of mean temperatu...

  9. Synthesis of barium titanate powders by low-temperature aqueous synthesis using a new segmented flow tubular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bowen, P.; Donnet, M. [Swiss Federal Inst. of Technology, Lausanne (Switzerland). Materials Science and Eng. Dept.; Testino, A.; Viviani, M.; Buscaglia, M.T.; Buscaglia, V. [National Research Council, Genoa (Italy). Inst. for Physical Chemistry of Materials; Nanni, P. [Chemical and Process Engineering Dept. - Univ. of Genoa (Italy)

    2002-07-01

    Barium titanate powders have been synthesised using a low temperature aqueous synthesis (LTAS) method in both batch and Segmented Flow Tubular Reactors. For similar stoichiometries the SFTR powder shows a much lower degree of aggregation with 80% particles <100 nm for the SFTR and only 57% <100 nm for the batch case. Specific surface areas for the SFTR powder are also higher at around 50 m{sup 2}/g compared to 38 m{sup 2}/g for the batch powders indicative of primary particle sizes of 24 nm (SFTR) and 31 nm (batch). After freeze drying, redispersion and slip casting the SFTR powder gives improved sintered densities (95.5 from 93.25%) the sintered microstructures are very similar. (orig.)

  10. Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

    Science.gov (United States)

    Koyanagi, Takaaki; Katoh, Yutai

    2017-10-01

    Silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230-340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

  11. Temperature increases from 55 to 75 C in a two-phase biogas reactor result in fundamental alterations within the bacterial and archaeal community structure

    Energy Technology Data Exchange (ETDEWEB)

    Rademacher, Antje [Leibniz-Institut fuer Agrartechnik Potsdam-Bornim e.V. (ATB), Potsdam (Germany). Abt. Bioverfahrenstechnik; Technische Univ. Berlin (Germany). Inst. fuer Technischen Umweltschutz; Nolte, Christine; Schoenberg, Mandy; Klocke, Michael [Leibniz-Institut fuer Agrartechnik Potsdam-Bornim e.V. (ATB), Potsdam (Germany). Abt. Bioverfahrenstechnik

    2012-10-15

    Agricultural biogas plants were operated in most cases below their optimal performance. An increase in the fermentation temperature and a spatial separation of hydrolysis/acetogenesis and methanogenesis are known strategies in improving and stabilizing biogas production. In this study, the dynamic variability of the bacterial and archaeal community was monitored within a two-phase leach bed biogas reactor supplied with rye silage and straw during a stepwise temperature increase from 55 to 75 C within the leach bed reactor (LBR), using TRFLP analyses. To identify the terminal restriction fragments that were obtained, bacterial and archaeal 16S rRNA gene libraries were constructed. Above 65 C, the bacterial community structure changed from being Clostridiales-dominated toward being dominated by members of the Bacteroidales, Clostridiales, and Thermotogales orders. Simultaneously, several changes occurred, including a decrease in the total cell count, degradation rate, and biogas yield along with alterations in the intermediate production. A bioaugmentation with compost at 70 C led to slight improvements in the reactor performance; these did not persist at 75 C. However, the archaeal community within the downstream anaerobic filter reactor (AF), operated constantly at 55 C, altered by the temperature increase in the LBR. At an LBR temperature of 55 C, members of the Methanobacteriales order were prevalent in the AF, whereas at higher LBR temperatures Methanosarcinales prevailed. Altogether, the best performance of this two-phase reactor was achieved at an LBR temperature of below 65 C, which indicates that this temperature range has a favorable effect on the microbial community responsible for the production of biogas. (orig.)

  12. Studies of low temperature, low flux radiation embrittlement of nuclear reactor structural materials. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Lucas, G.E.

    1993-06-01

    There are several existing research programs which have components pertinent to the issue of low flux/low temperature embrittlement; in particular, examination of the Shippingport shield tank which has been exposed to low flux and relatively low temperature is being performed by ANL, and evaluation of low temperature embrittlement in A508 and A533B steels in support of the HTGR is currently being performed by ORNL. However, these programs are not specifically directed at the broader issue of low flux/low temperature embrittlement in a range of structural steels. Hence, the authors coordinated their effort with these programs so that their investigations were complementary to existing programs, and they focused on a set of materials which expand the data base developed in these programs. In particular, the authors have investigated embrittlement phenomena in steels that are similar to those used in support structure.

  13. Biaxial Thermal Creep of Alloy 617 and Alloy 230 for VHTR Applications

    Energy Technology Data Exchange (ETDEWEB)

    Mo, Kun; Lv, Wei; Tung, Hsiao-Ming; Yun, Di; Miao, Yinbin; Lan, Kuan-Che; Stubbins, James F.

    2016-05-18

    In this study, we employed pressurized creep tubes to investigate the biaxial thermal creep behavior of Inconel 617 (alloy 617) and Haynes 230 (alloy 230). Both alloys are considered to he the primary candidate structural materials for very high-temperature reactors (VITITRs) due to their exceptional high-temperature mechanical properties. The current creep experiments were conducted at 900 degrees C for the effective stress range of 15-35 MPa. For both alloys, complete creep strain development with primary, secondary, and tertiary regimes was observed in all the studied conditions. Tertiary creep was found to he dominant over the entire creep lives of both alloys. With increasing applied creep stress, the fraction of the secondary creep regime decreases. The nucleation, diffusion, and coarsening of creep voids and carbides on grain boundaries were found to be the main reasons for the limited secondary regime and were also found to be the major causes of creep fracture. The creep curves computed using the adjusted creep equation of the form epsilon= cosh 1(1 rt) + P-sigma ntm agree well with the experimental results for both alloys at die temperatures of 850-950 degrees C.

  14. Finite element based stress analysis of graphite component in high temperature gas cooled reactor core using linear and nonlinear irradiation creep models

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurindranath

    2015-10-15

    Highlights: • High temperature gas cooled reactor. • Finite element based stress analysis. • H-451 graphite. • Irradiation creep model. • Graphite reflector stress analysis. - Abstract: Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  15. Simulation in CFD of a Pebble Bed: Advanced high temperature reactor core using OpenFOAM

    Energy Technology Data Exchange (ETDEWEB)

    Dahl, Pamela M.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    Numerical simulations of a Pebble Bed nuclear reactor core are presented using the multi-physics tool-kit OpenFOAM. The HTR-PM is modeled using the porous media approach, accounting both for viscous and inertial effects through the Darcy and Forchheimer model. Initially, cylindrical 2D and 3D simulations are compared, in order to evaluate their differences and decide if the 2D simulations carry enough of the sought information, considering the savings in computational costs. The porous medium is considered to be isotropic, with the whole length of the packed bed occupied homogeneously with the spherical fuel elements. Steady-state simulations for normal equilibrium operation are performed, using a semi sine function of the power density along the vertical axis as the source term for the energy balance equation.Total pressure drop is calculated and compared with that obtained from literature for a similar case. At a second stage, transient simulations are performed, where relevant parameters are calculated and compared to those of the literature. (author)

  16. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  17. The high-dose and high-temperature monitors of reactor irradiation based on insulators

    Directory of Open Access Journals (Sweden)

    V.A. Stepanov

    2015-10-01

    It has been experimentally shown that the use of Al2O3 single crystals and BN ceramics provides means of monitoring the temperature of irradiation from 370 to 1900 K. The temperature is derived from measurements of the optical absorption or X-ray diffraction line shifts after post-radiation annealing of the monitors. We discuss the applicability of (a the optical absorption and F-center luminescence spectroscopies of irradiated Al2O3 single crystals for gamma dose evaluation and (b the isotopic analysis of irradiated BN ceramics for neutron dose evaluation.

  18. Analysis of Creep Crack Growth Behavior of Alloy 617 for Use in a VHTR System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woo-Gon; Kim, Min-Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Jae-Young; Ekaputra, I. M. W.; Kim, Seon-Jin [Pukyong National University, Busan (Korea, Republic of)

    2015-05-15

    Alloy 617 is a major candidate material for the IHX component. The design of the component, which will operate well into the creep range, will require a good understanding of creep crack growth deformation. Efforts are now being undertaken in the Gen-IV program to provide data needed for the design and licensing of the nuclear plants, and with this goal in mind, to meet the needs of the conceptual designers of the VHTR system, 'Gen-IV Materials Handbook' is being established through an international collaboration program of GIF (Gen-IV Forum) countries. To logically obtain the B and q values in the CCGR equation, three methods in terms of LSFM, MVM, and PDM were adopted. The PDM was most useful. Both the B and q coefficients followed a lognormal distribution. Using a lognormal distribution in the PDM, a number of random variables were generated by Monte Carlo Simulation, and the CCGR lines could be successfully predicted from the viewpoint of reliability.

  19. Efficiency calculations and optimization analysis of a solar reactor for the high temperature step of the zinc/zinc-oxide thermochemical redox cycle

    Energy Technology Data Exchange (ETDEWEB)

    Haussener, S.

    2007-03-15

    A solar reactor for the first step of the zinc/zinc-oxide thermochemical redox cycle is analysed and dimensioned in terms of maximization of efficiency and reaction conversion. Zinc-oxide particles carried in an inert carrier gas, in our case argon, enter the reactor in absorber tubes and are heated by concentrated solar radiation mainly due to radiative heat transfer. The particles dissociate and, in case of complete conversion, a gas mixture of argon, zinc and oxygen leaves the reactor. The aim of this study is to find an optimal design of the reactor regarding efficiency, materials and economics. The number of absorber tubes and their dimensions, the cavity dimension and its material as well as the operating conditions should be determined. Therefore 2D and 3D simulations of an 8 kW reactor are implemented. The gases are modeled as ideal gases with temperature-dependent properties. Absorption and scattering of the particle gas mixture are calculated by Mie-theory. Radiative heat transfer is included in the simulation and implemented with the aid of the discrete ordinates (DO) method. The mixture is modeled as ideal mixture and the reaction with an Arrhenius-type ansatz. Temperature distribution, reaction efficiency (heat used for zinc-oxide reaction divided by input) and tube efficiency (heat going into absorber tubes divided by input) as well as reaction conversion are analyzed to find the most promising reactor design. The results show that the most significant factors for efficiencies, conversion and absorber fluid temperature are concentration of the solar incoming radiation, zinc-oxide mass flow, the number of tubes and their dimension. Higher concentration leads to solely positive effects. Zinc-oxide mass flow variations indicate the existence of an optimal flow rate for each reactor design which maximizes efficiencies and conversion. Higher zinc-oxide mass flow leads, on one hand, to higher tube efficiency but on the other hand to lower temperatures in

  20. Current hurdles to the success of high-temperature membrane reactors

    NARCIS (Netherlands)

    Saracco, G.; Versteeg, G.F.; Swaaij, W.P.M. van

    1994-01-01

    High-temperature catalytic processes performed using inorganic membranes have been in recent years a fast growing area of research, which seems to have not yet reached its peak. Chemical engineers, catalysts and materials scientists have addressed this topic from different viewpoints in a common

  1. Output feedback dissipation control for the power-level of modular high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Z. [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China)

    2011-07-01

    Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR) is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis. (author)

  2. Output Feedback Dissipation Control for the Power-Level of Modular High-Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2011-11-01

    Full Text Available Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis.

  3. Perancangan dan Simulasi MRAC untui Proses Pengendalian Temperatur pada Continuous Stirred Tank Reactor (CSTR

    Directory of Open Access Journals (Sweden)

    Amelia Sylvia

    2014-03-01

    Full Text Available Temperatur merupakan salah satu variabel proses dasar yang dikendalikan untuk menjaga suhu cairan di dalam reaktor. Model Reference Adaptive Controller (MRAC dengan MIT rule dipilih untuk mencapai spesifikasi respon yang diinginkan pada CSTR. Beban yang bervariasi berupa debit aliran likuid yang masuk ke dalam reaktor dapat menyebabkan perubahan parameter yang mempengaruhi perubahan temperatur output produk pada CSTR. Sebuah simulasi dilakukan dengan menggunakan MATLAB dan hasilnya dianalisa. Respon plant dapat melakukan adaptasi parameter – parameter kontrolernya cukup baik pada nilai gain adaptasi dengan rentang 0.00000010000 sampai 0.00000000001. Waktu yang dibutuhkan untuk mengatasi beban yang bervariasi berupa debit aliran yang masuk ke dalam reaktor dengan nilai yang maksimal (1.5 m^3/min menghasilkan respon plant lebih cepat 42 detik dari pada debit aliran masuk dengan nilai yang nominal (1 m^3/min 63 detik dan minimal (0.5 m^3/min 75 detik.

  4. Electron-induced dry reforming of methane in a temperature-controlled dielectric barrier discharge reactor

    KAUST Repository

    Zhang, Xuming

    2013-09-23

    Dry reforming of methane has the potential to reduce the greenhouse gases methane and carbon dioxide and to generate hydrogen-rich syngas. In reforming methane, plasma-assisted reforming processes may have advantages over catalytic processes because they are free from coking and their response time for mobile applications is quick. Although plasma-assisted reforming techniques have seen recent developments, systematic studies that clarify the roles that electron-induced chemistry and thermo-chemistry play are needed for a full understanding of the mechanisms of plasma-assisted reformation. Here, we developed a temperature-controlled coaxial dielectric barrier discharge (DBD) apparatus to investigate the relative importance of electron-induced chemistry and thermo-chemistry in dry reforming of methane. In the tested background temperature range 297-773 K, electron-induced chemistry, as characterized by the physical properties of micro-discharges, was found to govern the conversions of CH4 and CO2, while thermo-chemistry influenced the product selectivities because they were found to depend on the background temperature. Comparisons with results from arc-jet reformation indicated that thermo-chemistry is an efficient conversion method. Our findings may improve designs of plasma-assisted reformers by using relatively hotter plasma sources. However, detailed chemical kinetic studies are needed. © 2013 IOP Publishing Ltd.

  5. Characterization of a novel micro-pressure swirl reactor for removal of chemical oxygen demand and total nitrogen from domestic wastewater at low temperature.

    Science.gov (United States)

    Ren, Qingkai; Yu, Yang; Zhu, Suiyi; Bian, Dejun; Huo, Mingxin; Zhou, Dandan; Huo, Hongliang

    2017-06-01

     A novel micro-pressure swirl reactor (MPSR) was designed and applied to treat domestic wastewater at low temperature by acclimating microbial biomass with steadily decreasing temperature from 15 to 3 °C. Chemical oxygen demand (COD) was constantly removed by 85% and maintained below 50 mg L-1 in the effluent during the process. When the air flow was controlled at 0.2 m3 h-1, a swirl circulation was formed in the reactor, which created a dissolved oxygen (DO) gradient with a low DO zone in the center and a high DO zone in the periphery for denitrification and nitrification. 81% of total nitrogen was removed by this reactor, in which ammonium was reduced by over 90%. However, denitrification was less effective because of the presence of low levels of oxygen. The progressively decreasing temperature favored acclimation of psychrophilic bacteria in the reactor, which replaced mesophilic bacteria in the process of treatment.

  6. Review of activities of Research Association of High Temperature Gas Cooled Reactor Plant (RAHP)

    Energy Technology Data Exchange (ETDEWEB)

    An, Shigehiro [Research Association of High Temperature Gas Cooled Reactor Plant RAHP, University of Tokyo, Tokyo (Japan); Tsuchie, Yasuo [Research Association of High Temperature Gas Cooled Reactor Plant RAHP, The Japan Atomic Power Co. JAPC, Tokyo (Japan)

    1998-09-01

    The Research Association of the HTGR Plant (RAHP) is the sole research association in the private or industrial sector of Japan with respect to HTGR Plants. It was established in 1985, composing of professors, representatives of electric power companies, and fabricators of nuclear plant and fuels Activities in these years were to analyze world trends of R and D, to identify techno-economical issues to be cleared, to set-up fundamental development strategies, and to put the results of the studies into actions towards commercialization of the HTGR. Conclusions obtained through the activities so far are: (1) From the view point of effective use of energy and reduction of environmental impacts on a global scale, development of nuclear power is essential, in particular of the HTGR, because of its very highly inherent safety and feasibility of high temperature heat uses. The role of the HTGR is inter-complementary with those of LWR and FBR; (2) Future subjects on the HTGR are technical demonstration of its unique characteristics, economic prospects, public acceptance (PA) and industrial acceptance, R and D through international cooperation and share in role, and successful realization of demonstration plant(s). RAHP is to start a survey on HTGR from nuclear fuel cycle point of view to have a better outlook on future needs of high temperature heat uses. 6 refs.

  7. Calculations of multidimensional test reactor neutron diffusion problems with KORAT 3D code and its combination with two-velocity two-temperature thermohydraulic RATEG code

    Energy Technology Data Exchange (ETDEWEB)

    Voronova, O.A.; Grebennikov, A.N.; Zvenigorodskaya, O.A. [Russian Federal Nuclear Center, Arzamas (Russian Federation)] [and others

    1996-09-01

    The paper briefly presents the method for numerical computation of 3-D group neutron diffusion equation implemented in KORAT 3D code within SATURN complex and oriented to 3-D stationary and nonstationary calculations in nuclear reactor physics including those on up-to-date distributed-memory multiprocessors. The computational results are given obtained with the above mentioned code for multidimensional test reactor problems as well as the results of numerical efficiency studies for the code parallelization on multiprocessors Cray T3D and SP-2. Then the setup is given for 3-D dynamic problem simulating a hypothetical mode of WPBER reactor. The computational results are presented for the above problem which includes the neutron-nuclear processes, thermohydraulic processes and the feedback. The computations used two program complexes: KORAT 3D + multichannel two-velocity two-temperature thermohydraulics code RATEG and READY complex. (author)

  8. On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Holston, Anna-MariaAlvarez; Stjarnsater, Johan [Studsvik Nuclear AB, Nykoping (Sweden)

    2017-06-15

    Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below 300°C. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor (K{sub IH}) to initiate DHC as a function of temperature in Zry-4 for temperatures between 227°C and 315°C. The experimental technique used in this study was the pin-loading testing technique. To determine the K{sub IH}, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around 300°C, there was a sharp increase in K{sub IH} indicating the upper temperature limit for DHC. The value for K{sub IH} at 227°C was determined to be 2.6 ± 0.3 MPa √m.

  9. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    Science.gov (United States)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  10. Integration of High Temperature Gas-cooled Reactor Technology with Oil Sands Processes

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2011-10-01

    This paper summarizes an evaluation of siting an HTGR plant in a remote area supplying steam, electricity and high temperature gas for recovery and upgrading of unconventional crude oil from oil sands. The area selected for this evaluation is the Alberta Canada oil sands. This is a very fertile and active area for bitumen recovery and upgrading with significant quantities piped to refineries in Canada and the U.S Additionally data on the energy consumption and other factors that are required to complete the evaluation of HTGR application is readily available in the public domain. There is also interest by the Alberta oil sands producers (OSP) in identifying alternative energy sources for their operations. It should be noted, however, that the results of this evaluation could be applied to any similar oil sands area.

  11. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

  12. Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ballinger, Ronald G. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Wang, Chun Yun [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kadak, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todreas, Neil [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mirick, Bradley [Concepts, Northern Engineering and Research, Woburn, MA (United States); Demetri, Eli [Concepts, Northern Engineering and Research, Woburn, MA (United States); Koronowski, Martin [Concepts, Northern Engineering and Research, Woburn, MA (United States)

    2004-08-30

    power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

  13. The ARCHER project (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D)

    Energy Technology Data Exchange (ETDEWEB)

    Knol, S., E-mail: knol@nrg.eu [Nuclear Research and consultancy Group (NRG), PO Box 25, NL-1755 ZG Petten (Netherlands); Fütterer, M.A. [Joint Research Centre, Institute for Energy, Petten (Netherlands); Roelofs, F. [Nuclear Research and consultancy Group (NRG), PO Box 25, NL-1755 ZG Petten (Netherlands); Kohtz, N. [TÜV Rheinland, Köln (Germany); Laurie, M. [Joint Research Centre, Institute for Transuranium elements, Karlsruhe (Germany); Buckthorpe, D. [UMAN, University of Manchester, Manchester (United Kingdom); Scheuermann, W. [IKE, Stuttgart University, Stuttgart (Germany)

    2016-09-15

    The European HTR R&D project ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D) builds on a solid HTR technology foundation in Europe, established through former national UK and German HTR programs and in European framework programs. ARCHER runs from 2011 to 2015 and targets selected HTR R&D subjects that would specifically support demonstration, with a focus on experimental effort. In line with the R&D and deployment strategy of the European Sustainable Nuclear Energy Technology Platform (SNETP) ARCHER contributes to maintaining, strengthening and expanding the HTR knowledge base in Europe to lay the foundations for demonstration of nuclear cogeneration with HTR systems. The project consortium encompasses conventional and nuclear industry, utilities, Technical Support Organizations, R&D organizations and academia. ARCHER shares results with international partners in the Generation IV International Forum and collaborates directly with related projects in the US, China, Japan, the Republic of Korea and South Africa. The ARCHER project has finished, and the paper comprises an overview of the achievements of the project.

  14. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  15. Biogas production in an anaerobic sequencing batch reactor by using tequila vinasses: effect of pH and temperature.

    Science.gov (United States)

    Arreola-Vargas, J; Jaramillo-Gante, N E; Celis, L B; Corona-González, R I; González-Álvarez, V; Méndez-Acosta, H O

    2016-01-01

    In recent years, anaerobic digestion has been recognized as a suitable alternative for tequila vinasses treatment due to its high energy recovery and chemical oxygen demand (COD) removal efficiency. However, key factors such as the lack of suitable monitoring schemes and the presence of load disturbances, which may induce unstable operating conditions in continuous systems, have limited its application at full scale. Therefore, the aim of this work was to evaluate the anaerobic sequencing batch reactor (AnSBR) configuration in order to provide a low cost and easy operation alternative for the treatment of these complex effluents. In particular, the AnSBR was evaluated under different pH-temperature combinations: 7 and 32 °C; 7 and 38 °C; 8 and 32 °C and 8 and 38 °C. Results showed that the AnSBR configuration was able to achieve high COD removal efficiencies (around 85%) for all the tested conditions, while the highest methane yield was obtained at pH 7 and 38 °C (0.29 L/g COD added). Furthermore, high robustness was found in all the AnSBR experiments. Therefore, the full-scale application of the AnSBR technology for the treatment of tequila vinasses is quite encouraging, in particular for small and medium size tequila industries that operate under seasonal conditions.

  16. Life cycle assessment of hydrogen production from S-I thermochemical process coupled to a high temperature gas reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giraldi, M. R.; Francois, J. L.; Castro-Uriegas, D. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac No. 8532, Col. Progreso, C.P. 62550, Jiutepec, Morelos (Mexico)

    2012-07-01

    The purpose of this paper is to quantify the greenhouse gas (GHG) emissions associated to the hydrogen produced by the sulfur-iodine thermochemical process, coupled to a high temperature nuclear reactor, and to compare the results with other life cycle analysis (LCA) studies on hydrogen production technologies, both conventional and emerging. The LCA tool was used to quantify the impacts associated with climate change. The product system was defined by the following steps: (i) extraction and manufacturing of raw materials (upstream flows), (U) external energy supplied to the system, (iii) nuclear power plant, and (iv) hydrogen production plant. Particular attention was focused to those processes where there was limited information from literature about inventory data, as the TRISO fuel manufacture, and the production of iodine. The results show that the electric power, supplied to the hydrogen plant, is a sensitive parameter for GHG emissions. When the nuclear power plant supplied the electrical power, low GHG emissions were obtained. These results improve those reported by conventional hydrogen production methods, such as steam reforming. (authors)

  17. A Sample Calculation of Tritium Production and Distribution at VHTR by using TRITGO Code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ik Kyu; Kim, D. H.; Lee, W. J

    2007-03-15

    TRITGO code was developed for estimating the tritium production and distribution of high temperature gas cooled reactor(HTGR), especially GTMHR350 by General Atomics. In this study, the tritium production and distribution of NHDD was analyzed by using TRITGO Code. The TRITGO code was improved by a simple method to calculate the tritium amount in IS Loop. The improved TRITGO input for the sample calculation was prepared based on GTMHR600 because the NHDD has been designed referring GTMHR600. The GTMHR350 input with related to the tritium distribution was directly used. The calculated tritium activity among the hydrogen produced in IS-Loop is 0.56 Bq/g- H2. This is a very satisfying result considering that the limited tritium activity of Japanese Regulation Guide is 5.6 Bq/g-H2. The basic system to analyze the tritium production and the distribution by using TRITGO was successfully constructed. However, there exists some uncertainties in tritium distribution models, the suggested method for IS-Loop, and the current input was not for NHDD but for GTMHR600. The qualitative analysis for the distribution model and the IS-Loop model and the quantitative analysis for the input should be done in the future.

  18. Characteristics of DO, organic matter, and ammonium profile for practical-scale DHS reactor under various organic load and temperature conditions.

    Science.gov (United States)

    Nomoto, Naoki; Ali, Muntjeer; Jayaswal, Komal; Iguchi, Akinori; Hatamoto, Masashi; Okubo, Tsutomu; Takahashi, Masanobu; Kubota, Kengo; Tagawa, Tadashi; Uemura, Shigeki; Yamaguchi, Takashi; Harada, Hideki

    2017-04-20

    Profile analysis of the down-flow hanging sponge (DHS) reactor was conducted under various temperature and organic load conditions to understand the organic removal and nitrification process for sewage treatment. Under high organic load conditions (3.21-7.89 kg-COD m-3 day-1), dissolved oxygen (DO) on the upper layer of the reactor was affected by organic matter concentration and water temperature, and sometimes reaches around zero. Almost half of the CODCr was removed by the first layer, which could be attributed to the adsorption of organic matter on sponge media. After the first layer, organic removal proceeded along the first-order reaction equation from the second to the fourth layers. The ammoniacal nitrogen removal ratio decreased under high organic matter concentration (above 100 mg L-1) and low DO (less than 1 mg L-1) condition. Ammoniacal nitrogen removal proceeded via a zero-order reaction equation along the reactor height. In addition, the profile results of DO, CODCr, and NH3-N were different in the horizontal direction. Thus, it is thought the concentration of these items and microbial activities were not in a uniform state even in the same sponge layer of the DHS reactor.

  19. Fast Pyrolysis Behavior of Banagrass as a Function of Temperature and Volatiles Residence Time in a Fluidized Bed Reactor

    Science.gov (United States)

    Morgan, Trevor James; Turn, Scott Q.; George, Anthe

    2015-01-01

    A reactor was designed and commissioned to study the fast pyrolysis behavior of banagrass as a function of temperature and volatiles residence time. Four temperatures between 400 and 600°C were examined as well as four residence times between ~1.0 and 10 seconds. Pyrolysis product distributions of bio-oil, char and permanent gases were determined at each reaction condition. The elemental composition of the bio-oils and chars was also assessed. The greatest bio-oil yield was recorded when working at 450°C with a volatiles residence time of 1.4 s, ~37 wt% relative to the dry ash free feedstock (excluding pyrolysis water). The amounts of char (organic fraction) and permanent gases under these conditions are ~4 wt% and 8 wt% respectively. The bio-oil yield stated above is for 'dry' bio-oil after rotary evaporation to remove solvent, which results in volatiles and pyrolysis water being removed from the bio-oil. The material removed during drying accounts for the remainder of the pyrolysis products. The 'dry' bio-oil produced under these conditions contains ~56 wt% carbon which is ~40 wt% of the carbon present in the feedstock. The oxygen content of the 450°C, 1.4 s 'dry' bio-oil is ~38 wt%, which accounts for ~33 wt% of the oxygen in the feedstock. At higher temperature or longer residence time less bio-oil and char is recovered and more gas and light volatiles are produced. Increasing the temperature has a more significant effect on product yields and composition than increasing the volatiles residence time. At 600°C and a volatiles residence time of 1.2 seconds the bio-oil yield is ~21 wt% of the daf feedstock, with a carbon content of 64 wt% of the bio-oil. The bio-oil yield from banagrass is significantly lower than from woody biomass or grasses such as switchgrass or miscanthus, but is similar to barley straw. The reason for the low bio-oil yield from banagrass is thought to be related to its high ash content (8.5 wt% dry basis) and high concentration of alkali

  20. Fast Pyrolysis Behavior of Banagrass as a Function of Temperature and Volatiles Residence Time in a Fluidized Bed Reactor.

    Directory of Open Access Journals (Sweden)

    Trevor James Morgan

    Full Text Available A reactor was designed and commissioned to study the fast pyrolysis behavior of banagrass as a function of temperature and volatiles residence time. Four temperatures between 400 and 600°C were examined as well as four residence times between ~1.0 and 10 seconds. Pyrolysis product distributions of bio-oil, char and permanent gases were determined at each reaction condition. The elemental composition of the bio-oils and chars was also assessed. The greatest bio-oil yield was recorded when working at 450°C with a volatiles residence time of 1.4 s, ~37 wt% relative to the dry ash free feedstock (excluding pyrolysis water. The amounts of char (organic fraction and permanent gases under these conditions are ~4 wt% and 8 wt% respectively. The bio-oil yield stated above is for 'dry' bio-oil after rotary evaporation to remove solvent, which results in volatiles and pyrolysis water being removed from the bio-oil. The material removed during drying accounts for the remainder of the pyrolysis products. The 'dry' bio-oil produced under these conditions contains ~56 wt% carbon which is ~40 wt% of the carbon present in the feedstock. The oxygen content of the 450°C, 1.4 s 'dry' bio-oil is ~38 wt%, which accounts for ~33 wt% of the oxygen in the feedstock. At higher temperature or longer residence time less bio-oil and char is recovered and more gas and light volatiles are produced. Increasing the temperature has a more significant effect on product yields and composition than increasing the volatiles residence time. At 600°C and a volatiles residence time of 1.2 seconds the bio-oil yield is ~21 wt% of the daf feedstock, with a carbon content of 64 wt% of the bio-oil. The bio-oil yield from banagrass is significantly lower than from woody biomass or grasses such as switchgrass or miscanthus, but is similar to barley straw. The reason for the low bio-oil yield from banagrass is thought to be related to its high ash content (8.5 wt% dry basis and high

  1. Reactor Design for CO2 Photo-Hydrogenation toward Solar Fuels under Ambient Temperature and Pressure

    Directory of Open Access Journals (Sweden)

    Chun-Ying Chen

    2017-02-01

    Full Text Available Photo-hydrogenation of carbon dioxide (CO2 is a green and promising technology and has received much attention recently. This technique could convert solar energy under ambient temperature and pressure into desirable and sustainable solar fuels, such as methanol (CH3OH, methane (CH4, and formic acid (HCOOH. It is worthwhile to mention that this direction can not only potentially depress atmospheric CO2, but also weaken dependence on fossil fuel. Herein, 1 wt % Pt/CuAlGaO4 photocatalyst was successfully synthesized and fully characterized by ultraviolet-visible light (UV-vis spectroscopy, X-ray diffraction (XRD, Field emission scanning electron microscopy using energy dispersive spectroscopy analysis (FE-SEM/EDS, transmission electron microscopy (TEM, X-ray photoelectron spectroscopy (XPS, and Brunauer-Emmett-Teller (BET, respectively. Three kinds of experimental photo-hydrogenation of CO2 in the gas phase, liquid phase, and gas-liquid phase, correspondingly, were conducted under different H2 partial pressures. The remarkable result has been observed in the gas-liquid phase. Additionally, increasing the partial pressure of H2 would enhance the yield of product. However, when an extra amount of H2 is supplied, it might compete with CO2 for occupying the active sites, resulting in a negative effect on CO2 photo-hydrogenation. For liquid and gas-liquid phases, CH3OH is the major product. Maximum total hydrocarbons 8.302 µmol·g−1 is achieved in the gas-liquid phase.

  2. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor - FY-05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2005-09-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 9000C or operational fuel temperatures above 12500C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR’s higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Laboratory (INL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world’s computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertainty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  3. Creep-Fatigue Damage Investigation and Modeling of Alloy 617 at High Temperatures

    Science.gov (United States)

    Tahir, Fraaz

    The Very High Temperature Reactor (VHTR) is one of six conceptual designs proposed for Generation IV nuclear reactors. Alloy 617, a solid solution strengthened Ni-base superalloy, is currently the primary candidate material for the tubing of the Intermediate Heat Exchanger (IHX) in the VHTR design. Steady-state operation of the nuclear power plant at elevated temperatures leads to creep deformation, whereas loading transients including startup and shutdown generate fatigue. A detailed understanding of the creep-fatigue interaction in Alloy 617 is necessary before it can be considered as a material for nuclear construction in ASME Boiler and Pressure Vessel Code. Current design codes for components undergoing creep-fatigue interaction at elevated temperatures require creep-fatigue testing data covering the entire range from fatigue-dominant to creep-dominant loading. Classical strain-controlled tests, which produce stress relaxation during the hold period, show a saturation in cycle life with increasing hold periods due to the rapid stress-relaxation of Alloy 617 at high temperatures. Therefore, applying longer hold time in these tests cannot generate creep-dominated failure. In this study, uniaxial isothermal creep-fatigue tests with non-traditional loading waveforms were designed and performed at 850 and 950°C, with an objective of generating test data in the creep-dominant regime. The new loading waveforms are hybrid strain-controlled and force-controlled testing which avoid stress relaxation during the creep hold. The experimental data showed varying proportions of creep and fatigue damage, and provided evidence for the inadequacy of the widely-used time fraction rule for estimating creep damage under creep-fatigue conditions. Micro-scale damage features in failed test specimens, such as fatigue cracks and creep voids, were quantified using a Scanning Electron Microscope (SEM) to find a correlation between creep and fatigue damage. Quantitative statistical

  4. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials; Analyse des capacites des reacteurs a haute temperature sous l'aspect de l'utilisation des matieres fissiles

    Energy Technology Data Exchange (ETDEWEB)

    Damian, F

    2001-07-01

    The high temperature reactors fuel is made of micro-particles dispersed in a graphite matrix. This configuration makes it possible to reach high burnup, higher than 700 GWj/t. Thanks to the decoupling between the thermal and the neutronic behaviors in the core many types of fuels can be used. These characteristics give to HTR reactor very good capacities to burn fissile materials. This work was done in the frame of the evaluation of HTR capacities to enhance the value of the plutonium stocks. These stocks are currently composed of the irradiated fuels discharged from classical PWR or the dismantling of the nuclear weapons and represent a significant energy potential. These studies concluded that high cycles length can be reached whatever the plutonium quality is (from 50 % to 94 % of fissile plutonium). In addition, it was demonstrated that the moderator temperature coefficient becomes locally positive for highly burn fuel while the core global moderator temperature coefficient remained negative in the operation range of the reactor. A significant share of this work was first devoted to the setting of a modeling of the fuel element but also of the reactor's core with the codes of system SAPHYR. The whole of modeling was validated by reference calculations. This work of code assessment is justified by a preliminary work that showed that the classical calculation scheme used for PWR could not be transposed directly to HTR core. (author)

  5. Molecular Tagging Velocimetry Development for In-situ Measurement in High-Temperature Test Facility

    Science.gov (United States)

    Andre, Matthieu A.; Bardet, Philippe M.; Burns, Ross A.; Danehy, Paul M.

    2015-01-01

    The High Temperature Test Facility, HTTF, at Oregon State University (OSU) is an integral-effect test facility designed to model the behavior of a Very High Temperature Gas Reactor (VHTR) during a Depressurized Conduction Cooldown (DCC) event. It also has the ability to conduct limited investigations into the progression of a Pressurized Conduction Cooldown (PCC) event in addition to phenomena occurring during normal operations. Both of these phenomena will be studied with in-situ velocity field measurements. Experimental measurements of velocity are critical to provide proper boundary conditions to validate CFD codes, as well as developing correlations for system level codes, such as RELAP5 (http://www4vip.inl.gov/relap5/). Such data will be the first acquired in the HTTF and will introduce a diagnostic with numerous other applications to the field of nuclear thermal hydraulics. A laser-based optical diagnostic under development at The George Washington University (GWU) is presented; the technique is demonstrated with velocity data obtained in ambient temperature air, and adaptation to high-pressure, high-temperature flow is discussed.

  6. Computer Program of SIE ASME-NH Code for Structural Integrity Evaluation of Next Generation Reactors Subjecting to Elevated Temperature Operations

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2006-03-15

    In this report, the SIE ASME (Structural Integrity Evaluations by ASME-NH), which has a computerized implementation of ASME Pressure Vessels and Piping Code Section III Subsection NH rules, is developed to apply to the next generation reactor design subjecting to the elevated temperature operations over 500 .deg. C and over 30 years design lifetime, and the user's manual for this program is described in detail.

  7. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750–800°C Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    John Collins

    2009-08-01

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750–800°C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  8. Development of reactor graphite

    Science.gov (United States)

    Haag, G.; Mindermann, D.; Wilhelmi, G.; Persicke, H.; Ulsamer, W.

    1990-04-01

    The German graphite development programme for High Temperature Reactors has been based on the assumption that reactor graphite for core components with lifetime fluences of up to 4 × 10 22 neutrons per cm 2 (EDN) at 400°C can be manufactured from regular pitch coke. The use of secondary coke and vibrational moulding techniques have allowed production of materials with very small anisotropy, high strength, and high purity which are the most important properties of reactor graphite. A variety of graphite grades has been tested in fast neutron irradiation experiments. The results show that suitable graphites for modern High Temperature Reactors with spherical fuel elements are available.

  9. Pressurizing new reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neill, J.S.

    1956-01-30

    The Technical Division was asked recently to consider designs for new reactors that would add 8000 MW capacity to the Savannah River Plant. One modification of the existing SRP design that would enable a higher power rating, and therefore require fewer new reactors, is an increase in the maximum pressure in the D{sub 2}O system. The existing reactors at SRP are designed for a maximum pressure in the gas plenum of only 5 psig. Higher pressures enable higher D{sub 2} temperatures and higher sheath temperatures without local boiling or burnout. The requirements in reactor cooling facilities at any given power level would therefore be reduced by pressurizing.

  10. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    Science.gov (United States)

    1985-02-01

    implies, this reactor uses gas as the primary coolant . The coolant has a higher exit temperature when leaving the core than the PWR water 6 AFWL-TN-84...nuclear reactors, coolants must be used to ensure material components are not subject to failure due to the temperature exceeding melting points...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept

  11. Effect of process temperature on morphology of CNTs grown in a vertically fluidized bed reactor with Fe2O3/Al2O3 catalyst

    Science.gov (United States)

    Shukrullah, Shazia; Mohamed, Norani Muti; Shaharun, Maizatul Shima

    2015-07-01

    Carbon nanotubes (CNTs) are one of the most researched materials due to their exceptional mechanical and electrical properties. Among the various techniques, catalytic chemical vapor deposition in a fluidized bed reactor is the most promising technique for bulk production of CNTs. To meet the demand of good quality along with the bulk production of CNTs, the effect of reaction temperature on the micro structures, morphology, diameter, quality and quantity of CNTs was investigated in these studies. CNTs were synthesized at process temperature ranging from 700-850°C by catalytic decomposition of C2H4 on Fe2O3/Al2O3 catalyst a vertical fluidized bed reactor. The microstructures of the grown CNTs at different reaction temperatures were investigated by using scanning electron microscope. The results of this study depicted a positive correlation between the average diameter of CNTs and reaction temperature. Narrow diameters (35˜40 nm) of CNTs with fewer defects were found at the low and mild temperatures, in particular 800°C. At this temperature, a dynamic equilibrium between the rate of C2H4 decomposition and CNTs quantity was found due to maximum carbon diffusion over catalyst. The CNTs produced with Fe2O3/Al2O3 catalyst wer e also exhibiting high quality with relatively small mean outer diameter and fewer surface defects.

  12. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  13. Removal of residual dissolved methane gas in an upflow anaerobic sludge blanket reactor treating low-strength wastewater at low temperature with degassing membrane.

    Science.gov (United States)

    Bandara, Wasala M K R T W; Satoh, Hisashi; Sasakawa, Manabu; Nakahara, Yoshihito; Takahashi, Masahiro; Okabe, Satoshi

    2011-05-01

    In this study, we investigated the efficiency of dissolved methane (D-CH(4)) collection by degasification from the effluent of a bench-scale upflow anaerobic sludge blanket (UASB) reactor treating synthetic wastewater. A hollow-fiber degassing membrane module was used for degasification. This module was connected to the liquid outlet of the UASB reactor. After chemical oxygen demand (COD) removal efficiency of the UASB reactor became stable, D-CH(4) discharged from the UASB reactor was collected. Under 35 °C and a hydraulic retention time (HRT) of 10 h, average D-CH(4) concentration could be reduced from 63 mg COD L(-1) to 15 mg COD L(-1); this, in turn, resulted in an increase in total methane (CH(4)) recovery efficiency from 89% to 97%. Furthermore, we investigated the effects of temperature and HRT of the UASB reactor on degasification efficiency. Average D-CH(4) concentration was as high as 104 mg COD L(-1) at 15 °C because of the higher solubility of CH(4) gas in liquid; the average D-CH(4) concentration was reduced to 14 mg COD L(-1) by degasification. Accordingly, total CH(4) recovery efficiency increased from 71% to 97% at 15 °C as a result of degasification. Moreover, degasification tended to cause an increase in particulate COD removal efficiency. The UASB reactor was operated at the same COD loading rate, but different wastewater feed rates and HRTs. Although average D-CH(4) concentration in the UASB reactor was almost unchanged (ca. 70 mg COD L(-1)) regardless of the HRT value, the CH(4) discharge rate from the UASB reactor increased because of an increase in the wastewater feed rate. Because the D-CH(4) concentration could be reduced down to 12 ± 1 mg COD L(-1) by degasification at an HRT of 6.7 h, the CH(4) recovery rate was 1.5 times higher under degasification than under normal operation. Copyright © 2011 Elsevier Ltd. All rights reserved.

  14. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi [Japan Atomic Power Company, Tokyo (Japan)

    2001-07-01

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  15. Study on the Effects of Liquid Thermal Media on the Irradiation Capsule of High-Temperature Materials

    Directory of Open Access Journals (Sweden)

    Man Soon Cho

    2015-01-01

    Full Text Available Irradiation tests of materials at HANARO have usually been conducted using a standard capsule at temperatures of about 300°C for irradiation of materials used at PWR. Thus, the standard capsule uses aluminum as the specimen holder, which acts to dissipate the thermal energy. Future nuclear systems such as a VHTR and SFR require the irradiation tests at a relatively high temperature. As an alternative to aluminum which has been used as the thermal media in a standard material capsule, the characteristics of liquid metals such as NaK and LBE are reviewed. The temperatures of the capsule are affected by the variation of parameters such as the gap and wall thickness of the container. In particular, the external gap is most important in determining the temperature of the specimen. LBE raises the temperature of the specimen higher than NaK at the same configuration of the capsule. Thus, LBE can lessen the gap of the parts and reduce the vibration for a stable long-term test in reactor.

  16. Effect of the fuel element bundle statistical characteristics on the evaluation of temperature in the sodium-cooled fast-neutron reactor core

    Directory of Open Access Journals (Sweden)

    B.B. Tikhomirov

    2015-09-01

    Full Text Available Different fuel element bundle models used to calculate the coolant and fuel cladding temperatures inside fuel assemblies have been analyzed as applied to sodium-cooled fast-neutron reactors. The drawbacks of the existing models have been identified. A bundle model based on an experimental study into the actual arrangement of the fuel elements within the AF shroud has been proposed. The model's capabilities and advantages, as compared to conservative models, have been shown with regard for the need to raise the reliability of the fuel cladding working temperature estimation.

  17. Simultaneous measurements of the temperature, size and velocity distribution of coal particles in a coal dust pressure reactor; Simultane Erfassung der Temperatur, Groesse und Geschwindigkeit von Kohlepartikeln in einem Kohlenstaubdruckreaktor

    Energy Technology Data Exchange (ETDEWEB)

    Hackert, G.; Scherello, A.; Walter, M.; Wirtz, S.; Kremer, H.

    1997-12-31

    So far, methods of measurement in the field of particle combustion were commonly limited to measurements of particle clouds or wall temperatures. In order to investigate degassing and combustion processes of coal particles in a high-temperature, high-pressure downpipe reactor (KOALA), LEAT designed and constructed a new measuring system (TOSCA) by which individual coal particles can be made directly visible and their individual velocity and size can be measured quantitatively as well as, for the first time, their surface temperature. (orig.)

  18. Role of temperature in the radiation stability of yttria stabilized zirconia under swift heavy ion irradiation: A study from the perspective of nuclear reactor applications

    Science.gov (United States)

    Kalita, Parswajit; Ghosh, Santanu; Sattonnay, Gaël; Singh, Udai B.; Grover, Vinita; Shukla, Rakesh; Amirthapandian, S.; Meena, Ramcharan; Tyagi, A. K.; Avasthi, Devesh K.

    2017-07-01

    The search for materials that can withstand the harsh radiation environments of the nuclear industry has become an urgent challenge in the face of ever-increasing demands for nuclear energy. To this end, polycrystalline yttria stabilized zirconia (YSZ) pellets were irradiated with 80 MeV Ag6+ ions to investigate their radiation tolerance against fission fragments. To better simulate a nuclear reactor environment, the irradiations were carried out at the typical nuclear reactor temperature (850 °C). For comparison, irradiations were also performed at room temperature. Grazing incidence X-ray diffraction and Raman spectroscopy measurements reveal degradation in crystallinity for the room temperature irradiated samples. No bulk structural amorphization was however observed, whereas defect clusters were formed as indicated by transmission electron microscopy and supported by thermal spike simulation results. A significant reduction of the irradiation induced defects/damage, i.e., improvement in the radiation tolerance, was seen under irradiation at 850 °C. This is attributed to the fact that the rapid thermal quenching of the localized hot molten zones (arising from spike in the lattice temperature upon irradiation) is confined to 850 °C (i.e., attributed to the resistance inflicted on the rapid thermal quenching of the localized hot molten zones by the high temperature of the environment) thereby resulting in the reduction of the defects/damage produced. Our results present strong evidence for the applicability of YSZ as an inert matrix fuel in nuclear reactors, where competitive effects of radiation damage and dynamic thermal healing mechanisms may lead to a strong reduction in the damage production and thus sustain its physical integrity.

  19. Design and energy analysis of a electrolytic hydrogen production process by means of a high temperature nuclear reactor; Diseno y analisis energetico de un proceso de produccion de hidrogeno electrolitico por medio de un reactor nuclear de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI Campus Morelos, Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2008-07-01

    In this work an energy analysis to a process of production of hydrogen by means of electrolysis of high temperature is realized. This electrolysis type, unlike conventional electrolysis allows us to reach efficiencies of up to 60% because when increasing the temperature of the water, providing to its thermal energy, diminishes the demand of electrical energy required to separate the molecule of the water. Nevertheless, to obtain these efficiencies it is needed to have superheated aqueous vapor to but of 850 centigrade degrees, temperatures that can be reached about high temperature reactor; HTGR. In the present work it is mentioned to introduction way the importance of the hydrogen like energy vector and the advantages of obtaining it by means of nuclear energy. The electrolysis process of high temperature is described and a design is realized of this from its coupling to a nuclear power plant PBMR. The technological advances on which it counts the PBMR; efficiencies of 48% for optimized plants, their modular design and the thermodynamic cycle recuperative Brayton where upon operate; make the short term ideal candidate for the production of hydrogen. The thermodynamic analysis of optimized plant PBMR appears in another work, here the results of the balance of mass and energy involved in the process appear of hydrogen generation and the complete analysis of this. The result is a complete model of generation of hydrogen by electrolysis of high temperature coupled to an optimized plant PBMR that will be implemented for its dynamic simulation later. (Author)

  20. A development strategy for the business plan of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is)

    Energy Technology Data Exchange (ETDEWEB)

    Minatsuki, Isao, E-mail: isao_minatsuki@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Otani, Tomomi; Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Mizokami, Yorikata; Oyama, Sunao; Tsukamoto, Hiroki [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-Chome, Hyogo-ku, Kobe (Japan)

    2014-05-01

    A business plan and a new concept of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is) has been investigated toward a commercialization in near future by Mitsubishi Heavy Industries cooperated with Japan Atomic Energy Agency (JAEA) in Japan. The potential market of small sized reactor is expected to increase from the points of view of smaller investment, industrial use of the nuclear heat and IPP (Independent Power Producer). Especially minimization of construction unit cost including R and D and plant construction period are important issues in order to realize a business plan for them. The study includes four pertinent subject areas of (1) a market analysis, (2) a conceptual design, (3) improvement of safety design and (4) plant dynamics. In summary, the MHR-50/100 is designed to target a short construction period, competitive cost, and an inherent safety feature while applying only the verified technology of the High Temperature Engineering Test Reactor (HTTR) of JAEA or conventional technologies.

  1. Biohydrogen production from Tequila vinasses in an anaerobic sequencing batch reactor: effect of initial substrate concentration, temperature and hydraulic retention time.

    Science.gov (United States)

    Buitrón, Germán; Carvajal, Carolina

    2010-12-01

    The effect of the temperature (25 and 35 degrees C), the hydraulic retention time, HRT, (12 and 24 h) and initial substrate concentration on hydrogen production from Tequila vinasse was studied using a sequencing batch reactor. When 25 degrees C and 12-h HRT were applied, only insignificant biogas quantities were produced; however, using 24 h of HRT and temperatures of 25 and 35 degrees C, biogas containing hydrogen was produced. A maximum volumetric hydrogen production rate of 50.5 mL H(2) L(-1) h(-1) (48 mmol H(2) L(reactor)(-1) d(-1)) and an average hydrogen content in the biogas of 29.2+/-8.8% were obtained when the reactor was fed with 3 g COD L(-1), at 35 degrees C and 12-h HRT. Methane formation was observed when the longer HRT was applied. Results demonstrated the feasibility to produce hydrogen from this waste without a previous pre-treatment. 2010 Elsevier Ltd. All rights reserved.

  2. Treatment of a chocolate industry wastewater in a pilot-scale low-temperature UASB reactor operated at short hydraulic and sludge retention time.

    Science.gov (United States)

    Esparza-Soto, M; Arzate-Archundia, O; Solís-Morelos, C; Fall, C

    2013-01-01

    The aim of this work was to evaluate the performance of a 244-L pilot-scale upflow anaerobic sludge blanket (UASB) reactor during the treatment of chocolate-processing industry wastewater under low-temperature conditions (18 ± 0.6 °C) for approximately 250 d. The applied organic loading rate (OLR) was varied between 4 and 7 kg/m(3)/d by varying the influent soluble chemical oxygen demand (CODsol), while keeping the hydraulic retention time constant (6.4 ± 0.3 h). The CODsol removal efficiency was low (59-78%). The measured biogas production increased from 240 ± 54 to 431 ± 61 L/d during the experiments. A significant linear correlation between the measured biogas production and removed OLR indicated that 81.69 L of biogas were produced per kg/m(3) of CODsol removed. Low average reactor volatile suspended solids (VSS) (2,700-4,800 mg/L) and high effluent VSS (177-313 mg/L) were derived in a short sludge retention time (SRT) (4.9 d). The calculated SRT was shorter than those reported in the literature, but did not affect the reactor's performance. Average sludge yield was 0.20 kg-VSS/kg-CODsol. The low-temperature anaerobic treatment was a good option for the pre-treatment of chocolate-processing industry wastewater.

  3. High-temperature Elastic Structural Analysis on a Small-Scale PHE prototype

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Hong, S. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, H. Y. [AD solution, Daejeon (Korea, Republic of)

    2012-05-15

    Hydrogen is considered a promising future energy solution because it is clean, abundant, and storable, and has high-energy density. One of the major challenges in establishing a hydrogen economy is how to produce massive quantities of hydrogen in a clean, safe, and economical way. Among the various hydrogen production methods, nuclear hydrogen production is garnering worldwide attention since it can produce hydrogen, a promising energy carrier, without an environmental burden. The PHE (Process Heat Exchanger) is a component that utilizes the nuclear heat from the nuclear reactor to provide hydrogen. A PHE is used in several processes such as nuclear steam reforming, nuclear methanol, nuclear steel, nuclear oil refinery, and nuclear steam. The PHE of the SO{sub 3} decomposer, which generates the process gas such as H{sub 2}O, O{sub 2}, SO{sub 2}, and SO{sub 3} at a very high temperature, is a key component in the nuclear hydrogen program in Korea. Recently, KAERI (Korea Atomic Energy Research Institute) established a small-scale gas loop for the performance test of VHTR components and manufactured a small-scale PHE prototype made of Hastelloy-X. A performance test on the PHE prototype is under way in the small-scale gas loop at KAERI. In this study, in an effort to evaluate the high-temperature structural integrity of the PHE prototype under the test condition of the gas loop, a high-temperature elastic structural analysis was performed

  4. Selection of support structure materials for irradiation experiments in the HFIR (High Flux Isotope Reactor) at temperatures up to 500 degrees C

    Energy Technology Data Exchange (ETDEWEB)

    Farrell, K.; Longest, A.W.

    1990-01-01

    The key factor in the design of capsules for irradiation of test specimens in the High Flux Isotope Reactor at preselected temperatures up to 500{degree}C utilizing nuclear heating is a narrow gas-filled gap which surrounds the specimens and controls the transfer of heat from the specimens through the wall of a containment tube to the reactor cooling water. Maintenance of this gap to close tolerances is dependent on the characteristics of the materials used to support the specimens and isolate them from the water. These support structure materials must have low nuclear heating rates, high thermal conductivities, and good dimensional stabilities under irradiation. These conditions are satisfied by certain aluminum alloys. One of these alloys, a powder metallurgy product containing a fine dispersion of aluminum oxide, is no longer manufactured. A new alloys of this type, with the trade name DISPAL, is determined to be a suitable substitute. 23 refs., 13 figs., 3 tabs.

  5. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  6. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    Energy Technology Data Exchange (ETDEWEB)

    Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

    2007-05-02

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  7. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  8. State of Art Report for the Bypass and Cross Flows in Prismatic Modular High-Temperature Gas-Cooled Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Kim, Min Hwan

    2010-01-15

    The horizontal and vertical gaps between the adjacent fuel blocks occurs significantly in the prismatic modular high-temperature gas-cooled reactor core due to the thermal expansion, irradiation expansion/shrinking, and fuel block column bowing by pressure and gravity during the plant operation, in addition to the initial manufacture/design tolerance of fuel/reflector blocks. The coolant leakage of helium gas occurs through the gaps. These bypass and cross flows highly impact on the effective core cooling flow rate. This report describes the technical state of the bypass and cross flow study, based on ten reference papers, reviewing and summarizing four classified contents in the viewpoints of 'The reactor core fluctuation data for the first identification of the importance of the bypass and cross flow', 'The cross flow test and evaluation data', 'The flow test and evaluation data of the seal mechanism to prevent the leakage flow', and 'The reactor core thermal-fluid analysis and evaluation dat000.

  9. Reactor modeling to simulate catalytic partial oxidation and steam reforming of methane. Comparison of temperature profiles and strategies for hot spot minimization

    Energy Technology Data Exchange (ETDEWEB)

    Barrio, V.L.; Cambra, J.F.; Arias, P.L.; Gueemez, M.B. [School of Engineering (UPV/EHU), c/Alameda Urquijo s/n, 48013 Bilbao (Spain); Schaub, G.; Rohde, M. [Engler-Bunte-Institut, Universitaet Karlsruhe, Engler-Bunte-Ring 1, 76131 Karlsruhe (Germany); Rabe, S.; Vogel, F. [Paul Scherrer Institute, Laboratory for Energy and Materials Cycles, CH-5232 Villigen PSI (Switzerland)

    2007-07-15

    Catalytic partial oxidation (CPO) reactions of methane in the presence of steam (low temperature CPO, LTCPO) over a noble metal catalyst were investigated. A quasi-homogeneous one-dimensional model was developed in order to model a lab-scale fixed-bed reactor to produce syngas. These model calculations can contribute to the optimization of the process with respect to the formation of important hot spots ({delta} T{>=}130 ). These useful data can enable the model development in order to study the influence of the space velocity, product composition and other variables. Furthermore, a feed splitting study was performed. In the heat balance an overall heat transport term was included to account for small heat losses/gains along the reactor. The agreement between simulations and the degree of detail in the model is appropriate for the amount and kind of experimental data available. Thus, this model can also assist in a pilot reactor design, materials and further scale-up. (author)

  10. An in situ spatially resolved analytical technique to simultaneously probe gas phase reactions and temperature within the packed bed of a plug flow reactor.

    Science.gov (United States)

    Touitou, Jamal; Burch, Robbie; Hardacre, Christopher; McManus, Colin; Morgan, Kevin; Sá, Jacinto; Goguet, Alexandre

    2013-05-21

    This paper reports the detailed description and validation of a fully automated, computer controlled analytical method to spatially probe the gas composition and thermal characteristics in packed bed systems. As an exemplar, we have examined a heterogeneously catalysed gas phase reaction within the bed of a powdered oxide supported metal catalyst. The design of the gas sampling and the temperature recording systems are disclosed. A stationary capillary with holes drilled in its wall and a moveable reactor coupled with a mass spectrometer are used to enable sampling and analysis. This method has been designed to limit the invasiveness of the probe on the reactor by using the smallest combination of thermocouple and capillary which can be employed practically. An 80 μm (O.D.) thermocouple has been inserted in a 250 μm (O.D.) capillary. The thermocouple is aligned with the sampling holes to enable both the gas composition and temperature profiles to be simultaneously measured at equivalent spatially resolved positions. This analysis technique has been validated by studying CO oxidation over a 1% Pt/Al2O3 catalyst and the spatial resolution profiles of chemical species concentrations and temperature as a function of the axial position within the catalyst bed are reported.

  11. BLAST: a digital computer program for the dynamic simulation of the high temperature gas cooled reactor reheater-steam generator module

    Energy Technology Data Exchange (ETDEWEB)

    Hedrick, R.A.; Cleveland, J.C.

    1976-06-24

    BLAST simulates the high temperature gas cooled reactor reheater-steam generator module with a multi-node, fixed boundary, homogenous flow model. The time dependent conservation of energy, mass, and momentum equations are solved by an implicit integration technique. The code contains equation of state formulations for both helium and water as well as heat transfer and friction factor correlations. Normal operational transients and more severe transients such as those resulting in low and/or reverse flow can be simulated. The code calculates helium and water temperature, pressure, flow rate, and tube bulk and wall temperatures at various points within the reheater-steam generator module during the transients. BLAST predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in BLAST. BLAST is written in FORTRAN IV for the IBM 360/91 computer at the Oak Ridge National Laboratory.

  12. Influence of temperature and pretreatments on the anaerobic digestion of wastewater grown microalgae in a laboratory-scale accumulating-volume reactor.

    Science.gov (United States)

    Kinnunen, Viljami; Craggs, Rupert; Rintala, Jukka

    2014-06-15

    This laboratory-scale study investigated the performance of a low-cost anaerobic digester for microalgae. Low (∼2%) solids content wastewater-grown microalgal biomass (MB) was digested in an unmixed, accumulating-volume reactor (AVR) with solid and liquid separation that enabled a long solids retention time. AVRs (2 or 20 L) were operated at 20 °C, 37 °C or ambient temperature (8-21 °C), and the influence of two pretreatments - low-temperature thermal (50-57 °C) and freeze-thaw - on algal digestion were studied. The highest methane yield from untreated MB was in the 37 °C AVR with 225 L CH4 kg volatile solids (VS)(-1), compared with 180 L CH4 kg VS(-1)added in a conventional, 37 °C completely stirred tank reactor (CSTR), and 101 L CH4 kg VS(-1)added in the 20 °C AVR. Freeze-thaw and low-temperature thermal pretreatments promoted protein hydrolysis and increased methane yields by 32-50% at 20 °C, compared with untreated MB. Pretreatments also increased the mineralisation of nitrogen (41-57%) and phosphorus (76-84%) during digestion. MB digestion at ambient temperature was comparable with digestion at 20 °C, until temperature dropped below 16 °C. Copyright © 2014 Elsevier Ltd. All rights reserved.

  13. Development of a neutronic model for the fuel of a high temperature gas reactor type PBMR; Desarrollo de un modelo neutronico para el combustible de un reactor de gas de alta temperatura tipo PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Oropeza C, I.; Carmona H, R.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ivonucci@prodigy.net.mx

    2008-07-01

    In this work was developed the neutronic model of a fuel sphere of a nuclear reactor of gas of high temperature to modulate of bed of spheres (PBMR), using the Monte Carlo method with the MCNPx code. In order to be able to verify the fuel model constructed in this investigation, it is used a case of reference, based on an international exercise {sup b}enchmark{sup .} The benchmark report contains the results sent by different international participants for five phases with respect to the high temperature gas reactor (HTR), fed with uranium, plutonium and thorium. In particular, in first stage of benchmark an infinite adjustment of uranium compound fuel spheres is considered unique, with which our results were compared. This first stage considers two cases: cell calculations with spherical external frontier and cell calculations with cubic external frontier. The objective is to identify any increase in the uncertainty, related to the uranium fuel, that is associated with the plutonium and thorium fuels. In order to validate our results, the values of the neutron multiplication factor were taken in account, in cold and in the heat of the moment from the participants who sent their results obtained with Monte Carlo and deterministic calculations. The model of the fuel sphere developed in this work considers a regular distribution of 15000 Triso particles, in a cubic mesh centered within the sphere. For it was necessary to define the step firstly or {sup p}itch{sup o}f the cubic mesh. Generally, the results obtained by the participants of benchmark and those of this investigation present good agreement, nevertheless, appear some discrepancies, attributed to factors like different libraries of cross sections used, the nature of the solution: Monte Carlo or deterministic, and the difficulty of some participants to model the external frontier condition of reflection. (Author)

  14. High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 10, December 21, 1991--March 20, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, T.T.

    1992-07-01

    In this project we will study a novel process concept, i.e., the use of ceramic membrane reactors in upgrading of coal model compounds and coal derived liquids. In general terms, the USC research team is responsible for constructing and operating the membrane reactor apparatus and for testing various inorganic membranes for the upgrading of coal derived asphaltenes and coal model compounds. The USC effort will involve the principal investigator of this project and two graduate research assistants. The ALCOA team is responsible for the preparation of the inorganic membranes, for construction and testing of the ceramic membrane modules, and for measurement of their transport properties. The ALCOA research effort will involve Dr. Paul K. T. Liu, who is the project manager of the ALCOA research team, an engineer and a technician. UNOCAL`s contribution will be limited to overall technical assistance in catalyst preparation and the operation of the laboratory upgrading membrane reactor and for analytical back-up and expertise in oil analysis and materials characterization. UNOCAL is a no-cost contractor but will be involved in all aspects of the project, as deemed appropriate.

  15. High Temperature Humidity Sensor for Detection of Leak Through Slits and Cracks in Pressurized Nuclear Power Reactor Pipes

    Directory of Open Access Journals (Sweden)

    Debdulal Saha

    2007-03-01

    Full Text Available The leak before break (LBB concept is well known to nuclear power reactor. The problem is common to water power reactor. This is based on the premise that a detectable leak will develop before catastrophic break occurs. The main purpose of the present study is to develop tape cast MgCr2O4+35mole% TiO2 and gel cast g-Al2O3 humidity sensor for use in LBB applications at 3000C. The material capacitance varies with transient injection of water vapour adsorption. In actual plant, the sensors are placed on a steam pipe surrounded by heat insulation. The pipe unites the nuclear reactor and power generator. The analysis of humidity distribution in the annulus is calculated assuring leak rate 0.1gpm in a 30 m long tube. In this paper, analysis is done on the basis of the two types of sensor using AC frequency. Performance characteristics are observed for the LLB application.

  16. Natural Circulation Patterns in the VHTR Air-Ingress Accident and Related Issues

    Energy Technology Data Exchange (ETDEWEB)

    Chang Ho Oh; Eung Soo Kim; Hyung Seok Kang

    2010-10-01

    A natural circulation pattern in a Very High Gas-Cooled Reactor during a hypothetical air-ingress accident has been investigated using computational fluid dynamic (CFD) methods in order to compare with the previous 1-D flow path model for the air-ingress analyses. The GT-MHR 600 MWt reactor was selecte