1170 MW/sub t/ HTGR steamer cogeneration plant: design and cost study

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

A conceptual design and cost study is presented for intermediate size high temperature gas-cooled reactor (HTGR) for industrial energy applications performed by United Engineers and Constructors Inc., (UE and C) and The General Atomic Company (GAC). The study is part of a program at ORNL and has the objective to provide support in the evaluation of the technical and economic feasibility of a single unit 1170 MW/sub t/ HTGR steam cycle cogeneration plant (referred to as the Steamer plant) for the production of industrial process energy. Inherent in the achievement of this objective, it was essential to perform a numbermore » of basic tasks such as the development of plant concept, capital cost estimate, project schedule and annual operation and maintenance (O and M) cost.« less

Baseline Concept Description of a Small Modular High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 NGNPmore » 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 Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

Baseline Concept Description of a Small Modular High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans D.

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 NGNPmore » 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 Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richards, Matt; Hamilton, Chris

This report provides supplemental information to the assessment of target markets provided in Appendix A of the 2012 Next Generation Nuclear Plant (NGNP) Industry Alliance (NIA) business plan [NIA 2012] for deployment of High Temperature Gas-Cooled Reactors (HTGRs) in the 2025 – 2050 time frame. This report largely reiterates the [NIA 2012] assessment for potential deployment of 400 to 800 HTGR modules (100 to 200 HTGR plants with 4 reactor modules) in the 600-MWt class in North America by 2050 for electricity generation, co-generation of steam and electricity, oil sands operations, hydrogen production, and synthetic fuels production (e.g., coal tomore » liquids). As the result of increased natural gas supply from hydraulic fracturing, the current and historically low prices of natural gas remain a significant barrier to deployment of HTGRs and other nuclear reactor concepts in the U.S. However, based on U.S. Department of Energy (DOE) Energy Information Agency (EIA) data, U.S. natural gas prices are expected to increase by the 2030 – 2040 timeframe when a significant number of HTGR modules could be deployed. An evaluation of more recent EIA 2013 data confirms the assumptions in [NIA 2012] of future natural gas prices in the range of approximately $7/MMBtu to $10/MMBtu during the 2030 – 2040 timeframe. Natural gas prices in this range will make HTGR energy prices competitive with natural gas, even in the absence of carbon-emissions penalties. Exhibit ES-1 presents the North American projections in each market segment including a characterization of the market penetration logic. Adjustments made to the 2012 data (and reflected in Exhibit ES-1) include normalization to the slightly larger 625MWt reactor module, segregation between steam cycle and more advanced (higher outlet temperature) modules, and characterization of U.S. synthetic fuel process applications as a separate market segment.« less

IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, Gerhard; Bostelmann, F.

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained).more » SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V&V) of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.« less

AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pham, Binh T.; Einerson, Jeffrey J.; Scates, Dawn M.

A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of heliummore » coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes.« less

Energy alternative for industry: the high-temperature gas-cooled reactor steamer

DOE Office of Scientific and Technical Information (OSTI.GOV)

McMain, A.T. Jr.; Blok, F.J.

1978-04-01

Large industrial complexes are faced with new requirements that will lead to a transition from such fluid fuels as natural gas and oil to such solid fuels as coal and uranium for supply of industrial energy. Power plants using these latter fuels will be of moderate size (800 to 1200 MW(thermal)) and will generally have the capability of co-generating electric power and process steam. A study has been made regarding use of the 840-MW(thermal) Fort St. Vrain high-temperature gas-cooled reactor (HTGR) design for industrial applications. The initial conceptual design (referred to as the HTGR Steamer) is substantially simplified relative tomore » Fort St. Vrain in that outlet helium and steam temperatures are lower and the reheat section is deleted from the steam generators. The Steamer has four independent steam generating loops producing a total of 277 kg/s (2.2 x 10/sup 6/ lb/h) of prime steam at 4.5 MPa/672 K (650 psia/750/sup 0/F). The unit co-generates 46 MW(electric) and provides process steam at 8.31 MPa/762 K(1200 psia/912/sup 0/F). The basic configuration and much of the equipment are retained from the Fort St. Vrain design. The system has inherent safety features important for industrial applications. These and other features indicate that the HTGR Steamer is an industrial energy option deserving additional evaluation. Subsequent work will focus on parallel design optimization and application studies.« less

ICP-MS measurement of silver diffusion coefficient in graphite IG-110 between 1048K and 1284K

NASA Astrophysics Data System (ADS)

Carter, L. M.; Seelig, J. D.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

2018-01-01

Silver-110m has been shown to permeate intact silicon carbide and pyrolytic carbon coating layers of the TRISO fuel particles during normal High Temperature Gas-Cooled Reactor (HTGR) operational conditions. The diffusion coefficients for silver in graphite IG-110 measured using a release method designed to simulate HTGR conditions of high temperature and flowing helium in the temperature range 1048-1253 K are reported. The measurements were made using spheres milled from IG-110 graphite that were infused with silver using a pressure vessel technique. The Ag diffusion was measured using a time release technique with an ICP-MS instrument for detection. The results of this work are:

HTGR plant availability and reliability evaluations. Volume I. Summary of evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cadwallader, G.J.; Hannaman, G.W.; Jacobsen, F.K.

1976-12-01

The report (1) describes a reliability assessment methodology for systematically locating and correcting areas which may contribute to unavailability of new and uniquely designed components and systems, (2) illustrates the methodology by applying it to such components in a high-temperature gas-cooled reactor (Public Service Company of Colorado's Fort St. Vrain 330-MW(e) HTGR), and (3) compares the results of the assessment with actual experience. The methodology can be applied to any component or system; however, it is particularly valuable for assessments of components or systems which provide essential functions, or the failure or mishandling of which could result in relatively largemore » economic losses.« less

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

DOEpatents

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

2010-02-23

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). 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, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

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

DOEpatents

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.

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

DOEpatents

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.

Fabrication methods and anisotropic properties of graphite matrix compacts for use in HTGR

NASA Astrophysics Data System (ADS)

Yeo, Sunghwan; Yun, Jihae; Kim, Sungok; Cho, Moon Sung; Lee, Young-Woo

2018-02-01

This study investigated the anisotropic microstructural, mechanical, and thermal properties of fabricated graphite matrix prismatic compacts for High Temperature Gas Cooled Reactor (HTGR) fuel. When the observed alignment of graphite grains and the coke derived from phenolic resin is in the transverse direction, the result is severely anisotropic thermal properties. Compacts with such orientation in the transverse direction exhibited increases of thermal expansion and conductivity up to 5.8 times and 4.82 times, respectively, more than those in the axial direction. The formation of pores due to the pyrolysis of phenolic resin was observed predominantly on upper region of the fabricated compacts. This anisotropic pore formation created anisotropic Vickers hardness on the planes with different directions.

Master Logic Diagram: An Approach to Identify Initiating Events of HTGRs

NASA Astrophysics Data System (ADS)

Purba, J. H.

2018-02-01

Initiating events of a nuclear power plant being evaluated need to be firstly identified prior to applying probabilistic safety assessment on that plant. Various types of master logic diagrams (MLDs) have been proposedforsearching initiating events of the next generation of nuclear power plants, which have limited data and operating experiences. Those MLDs are different in the number of steps or levels and different in the basis for developing them. This study proposed another type of MLD approach to find high temperature gas cooled reactor (HTGR) initiating events. It consists of five functional steps starting from the top event representing the final objective of the safety functions to the basic event representing the goal of the MLD development, which is an initiating event. The application of the proposed approach to search for two HTGR initiating events, i.e. power turbine generator trip and loss of offsite power, is provided. The results confirmed that the proposed MLD is feasiblefor finding HTGR initiating events.

DOE Office of Scientific and Technical Information (OSTI.GOV)

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

System Evaluation and Economic Analysis of a HTGR Powered High-Temperature Electrolysis Hydrogen Production Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael G. McKellar; Edwin A. Harvego; Anastasia A. Gandrik

2010-10-01

A design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The power conversion unit will be a Rankine steam cycle with a power conversion efficiency of 40%. The reference hydrogen production plantmore » operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 40.4% at a hydrogen production rate of 1.75 kg/s and an oxygen production rate of 13.8 kg/s. An economic analysis of this plant was performed with realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a cost of $3.67/kg of hydrogen assuming an internal rate of return, IRR, of 12% and a debt to equity ratio of 80%/20%. A second analysis shows that if the power cycle efficiency increases to 44.4%, the hydrogen production efficiency increases to 42.8% and the hydrogen and oxygen production rates are 1.85 kg/s and 14.6 kg/s respectively. At the higher power cycle efficiency and an IRR of 12% the cost of hydrogen production is $3.50/kg.« less

Summary of the Advanced Reactor Design Criteria (ARDC) Phase 2 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holbrook, Mark Raymond

This report provides an end-of-year summary reflecting the progress and status of proposed regulatory design criteria for advanced non-LWR designs in accordance with the Level 3 milestone in M3AT-15IN2001017 in work package AT-15IN200101. These criteria have been designated as ARDC, and they provide guidance to future applicants for addressing the GDC that are currently applied specifically to LWR designs. The report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of example adaptations of ARDC for Sodium Fast Reactor (SFR) and modular High Temperature Gas-cooled Reactor (HTGR) designs.

JPRS Report, Science & Technology, China: Energy

DTIC Science & Technology

1988-06-29

capacity. There are currently two types of HTGR reactor designs: the particle-bed core , which uses spherical fuel elements, and the rod type core , in...and trial operating experience with the HTGR reactor. Its main design features are as follows. 1. A particle-bed core , continuous fueling and...Favorable for Development of Small-Scale HTGR (Xu Jiming; HE DONGLI GONGCHENG, Feb 88) 47 ERRATUM: In JPRS-CEN-88-003 of 25 April 1988 in article

Kinetics of Chronic Oxidation of NBG-17 Nuclear Graphite by Water Vapor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Contescu, Cristian I; Burchell, Timothy D; Mee, Robert

2015-05-01

This report presents the results of kinetic measurements during accelerated oxidation tests of NBG-17 nuclear graphite by low concentration of water vapor and hydrogen in ultra-high purity helium. The objective is to determine the parameters in the Langmuir-Hinshelwood (L-H) equation describing the oxidation kinetics of nuclear graphite in the helium coolant of high temperature gas-cooled reactors (HTGR). Although the helium coolant chemistry is strictly controlled during normal operating conditions, trace amounts of moisture (predictably < 0.2 ppm) cannot be avoided. Prolonged exposure of graphite components to water vapor at high temperature will cause very slow (chronic) oxidation over the lifetimemore » of graphite components. This behavior must be understood and predicted for the design and safe operation of gas-cooled nuclear reactors. The results reported here show that, in general, oxidation by water of graphite NBG-17 obeys the L-H mechanism, previously documented for other graphite grades. However, the characteristic kinetic parameters that best describe oxidation rates measured for graphite NBG-17 are different than those reported previously for grades H-451 (General Atomics, 1978) and PCEA (ORNL, 2013). In some specific conditions, certain deviations from the generally accepted L-H model were observed for graphite NBG-17. This graphite is manufactured in Germany by SGL Carbon Group and is a possible candidate for the fuel elements and reflector blocks of HTGR.« less

A preliminary systems-engineering study of an advanced nuclear-electrolytic hydrogen-production facility

NASA Technical Reports Server (NTRS)

Escher, W. J. D.; Donakowski, T. D.; Tison, R. R.

1975-01-01

An advanced nuclear-electrolytic hydrogen-production facility concept was synthesized at a conceptual level with the objective of minimizing estimated hydrogen-production costs. The concept is a closely-integrated, fully-dedicated (only hydrogen energy is produced) system whose components and subsystems are predicted on ''1985 technology.'' The principal components are: (1) a high-temperature gas-cooled reactor (HTGR) operating a helium-Brayton/ammonia-Rankine binary cycle with a helium reactor-core exit temperature of 980 C, (2) acyclic d-c generators, (3) high-pressure, high-current-density electrolyzers based on solid-polymer electrolyte technology. Based on an assumed 3,000 MWt HTGR the facility is capable of producing 8.7 million std cu m/day of hydrogen at pipeline conditions, 6,900 kPa. Coproduct oxygen is also available at pipeline conditions at one-half this volume. It has further been shown that the incorporation of advanced technology provides an overall efficiency of about 43 percent, as compared with 25 percent for a contemporary nuclear-electric plant powering close-coupled contemporary industrial electrolyzers.

Fission product palladium-silicon carbide interaction in htgr fuel particles

NASA Astrophysics Data System (ADS)

Minato, Kazuo; Ogawa, Toru; Kashimura, Satoru; Fukuda, Kousaku; Shimizu, Michio; Tayama, Yoshinobu; Takahashi, Ishio

1990-07-01

Interaction of fission product palladium (Pd) with the silicon carbide (SiC) layer was observed in irradiated Triso-coated uranium dioxide particles for high temperature gas-cooled reactors (HTGR) with an optical microscope and electron probe microanalyzers. The SiC layers were attacked locally or the reaction product formed nodules at the attack site. Although the main element concerned with the reaction was palladium, rhodium and ruthenium were also detected at the corroded areas in some particles. Palladium was detected on both the hot and cold sides of the particles, but the corroded areas and the palladium accumulations were distributed particularly on the cold side of the particles. The observed Pd-SiC reaction depths were analyzed on the assumption that the release of palladium from the fuel kernel controls the whole Pd-SiC reaction.

Analyzing the impact of reactive transport on the repository performance of TRISO fuel

NASA Astrophysics Data System (ADS)

Schmidt, Gregory

One of the largest determiners of the amount of electricity generated by current nuclear reactors is the efficiency of the thermodynamic cycle used for power generation. Current light water reactors (LWR) have an efficiency of 35% or less for the conversion of heat energy generated by the reactor to electrical energy. If this efficiency could be improved, more power could be generated from equivalent volumes of nuclear fuel. One method of improving this efficiency is to use a coolant flow that operates at a much higher temperature for electricity production. A reactor design that is currently proposed to take advantage of this efficiency is a graphite-moderated, helium-cooled reactor known as a High Temperature Gas Reactor (HTGR). There are significant differences between current LWR's and the proposed HTGR's but most especially in the composition of the nuclear fuel. For LWR's, the fuel elements consist of pellets of uranium dioxide or plutonium dioxide that are placed in long tubes made of zirconium metal alloys. For HTGR's, the fuel, known as TRISO (TRIstructural-ISOtropic) fuel, consists of an inner sphere of fissile material, a layer of dense pyrolytic carbon (PyC), a ceramic layer of silicon carbide (SiC) and a final dense outer layer of PyC. These TRISO particles are then compacted with graphite into fuel rods that are then placed in channels in graphite blocks. The blocks are then arranged in an annular fashion to form a reactor core. However, this new fuel form has unanswered questions on the environmental post-burn-up behavior. The key question for current once-through fuel operations is how these large irradiated graphite blocks with spent fuel inside will behave in a repository environment. Data in the literature to answer this question is lacking, but nevertheless this is an important question that must be answered before wide-spread adoption of HTGR's could be considered. This research has focused on answering the question of how the large quantity of graphite surrounding the spent HTGR fuel will impact the release of aqueous uranium from the TRISO fuel. In order to answer this question, the sorption and partitioning behavior of uranium to graphite under a variety of conditions was investigated. Key systematic variables that were analyzed include solution pH, dissolved carbonate concentration, uranium metal concentration and ionic strength. The kinetics and desorption characteristics of uranium/graphite partitioning were studied as well. The graphite used in these experiments was also characterized by a variety of techniques and conclusions are drawn about the relevant surface chemistry of graphite. This data was then used to generate a model for the reactive transport of uranium in a graphite matrix. This model was implemented with the software code CXTFIT and validated through the use of column studies mirroring the predicted system.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mcwilliams, A. J.

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 techniquesmore » 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.« less

The Shock and Vibration Digest. Volume 17, Number 2

DTIC Science & Technology

1985-02-01

phenomena relative to A computer program has been developed to -.- buildings, bridges, dams, and other struc- simulate the motions of bodies subjected to...1982). (57) Ikushima, T., Honma, T., and Ishiz- uka, H., "Seismic Research on Block-Type (47) Kadle, D.S. and Chwang, A.T., "Hy- HTGR Core ," Nucl...T., "A Seismic Study of High Temperature Gas-Cooled Reactor Core - (48) Yang, C.Y., Chiarito, V., and Dressel, with Block-Type Fuel ; 2nd Rept: An Ana

Improvement of Modeling HTGR Neutron Physics by Uncertainty Analysis with the Use of Cross-Section Covariance Information

NASA Astrophysics Data System (ADS)

Boyarinov, V. F.; Grol, A. V.; Fomichenko, P. A.; Ternovykh, M. Yu

2017-01-01

This work is aimed at improvement of HTGR neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for high temperature gas cooled reactors (HTGR) was generated. This library can be used for estimation of uncertainties, associated with nuclear data, in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k∞) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. Eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: 10B(capt), 238U(n,γ), ν5, 235U(n,γ), 238U(el), natC(el), 235U(fiss)-235U(n,γ), 235U(fiss).

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Themore » result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.« less

Briefing Book. Volume 1: The Evolution of the Nuclear Non-Proliferation Regime (Fourth Edition).

DTIC Science & Technology

1998-01-01

usually termed) nuclear reactors. The first of these is that they contain a core or mass of fissile material (the fuel ) which may weigh tens of tons... HTGR is cooled with helium gas and moderated with graphite. Highly enriched uranium is used as fuel (93 per cent U-235), though this may be mixed with...to convert U-238 in a blanket around the core into Pu-239 at a rate faster than its own consumption of fissile material. They thus produce more fuel

Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Hu, Rui; Lisowski, Darius

2016-04-17

The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at themore » NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.« less

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul Andrew

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A seriesmore » of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).« less

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul; Harp, Jason; Winston, Phil

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series ofmore » fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.« less

Results for Phase I of the IAEA Coordinated Research Program on HTGR Uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, Gerhard; Bostelmann, Friederike; Yoon, Su Jong

2015-01-01

The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied. High Temperature Gas-cooled Reactors (HTGR) has its own peculiarities, coated particle design, large graphite quantities, different materials and high temperatures that also require other simulationmore » requirements. The IAEA has therefore launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the HTR-PM (INET, China). This report summarizes the contributions of the HTGR Methods Simulation group at Idaho National Laboratory (INL) up to this point of the CRP. The activities at INL have been focused so far on creating the problem specifications for the prismatic design, as well as providing reference solutions for the exercises defined for Phase I. An overview is provided of the HTGR UAM objectives and scope, and the detailed specifications for Exercises I-1, I-2, I-3 and I-4 are also included here for completeness. The main focus of the report is the compilation and discussion of reference results for Phase I (i.e. for input parameters at their nominal or best-estimate values), which is defined as the first step of the uncertainty quantification process. These reference results can be used by other CRP participants for comparison with other codes or their own reference results. The status on the Monte Carlo modeling of the experimental VHTRC facility is also discussed. Reference results were obtained for the neutronics stand-alone cases (Ex. I-1 and Ex. I-2) using the (relatively new) Monte Carlo code Serpent, and comparisons were performed with the more established Monte Carlo codes MCNP and KENO-VI. For the thermal-fluids stand-alone cases (Ex. I-3 and I-4) the commercial CFD code CFX was utilized to obtain reference results that can be compared with lower fidelity tools.« less

Promethus Hot Leg Piping Concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

AM Girbik; PA Dilorenzo

2006-01-24

The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactormore » (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept.« less

Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

NASA Technical Reports Server (NTRS)

Peterman, D. D.; Fontaine, R. W.; Quade, R. N.; Halvers, L. J.; Jahromi, A. M.

1975-01-01

The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program.

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

An Overview of Reactor Concepts, a Survey of Reactor Designs.

DTIC Science & Technology

1985-02-01

may be very different. HTGRs may use highly enriched uranium, thereby yielding better fuel economy and a reduc- tion of the actual core size for a...specific power level. The HTGR core may have fuel and control rods placed in graphite arrays similar to PWR core con- figuration, or they may have fuel ...rods are pulled out. A Peach Bottom core design is another HTGR design. This design is featured by the fuel pin’s ability to purge itself of fission

A SURVEY OF CONVENTIONAL STEAM BOILER EXPERIENCE APPLICABLE TO THE HTGR STEAM GENERATORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paget, J.A.

1959-10-01

BS>The steam generator of a high temperature gas-cooled reactor consists of tubular heating surface inside a shell which forms part of the primary He circuit of the reactor. When a tube fails in such a steam generator, moisture in the form of steam is released into the He steam and is carried through the reactor where it will cause corrosion and mass transfer of C in the core. A paramount consideration in the design of a steam generator for a high temperature gas-cooled reactor is the prevention of tube failures. Preference, therefore, should be given to a forced circulation design.more » The Loeffler Boiler would be the best from this standpoint alone since only steam enters the tubes, and its circulation rate can be maintained at an adequate value to insure cool tubes regardless of load fluctuations. The next type in the order of preference would be the forced recirculation boiler, since at least the boiier tubes always have an adequate cooling flow regardless of output. The third type in order of preference would be a Sulzer Type boiler since it has a separator to remove dissolved material from the water which is comparible in efficiency to a standard boiler drum and although the flow through evaporator and superheater fluctuates with load, the Sulzer Boiler can be operated as a forced recirculation boiler at low loads. The least desirable type would be a Benson or supercritical boiler which is completely dependent on input water purity for its survival. It is not claimed that Benson or supercritical boilers should not or will not be used in the future for gas-cooled reactors, but only that their use would be the least conservative choice from a tube failure standpoint at the present time. (auth)« less

Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stroh, K.R.

1980-01-01

The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are beingmore » used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences.« less

Role of nuclear grade graphite in controlling oxidation in modular HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Windes, Willaim; Strydom, G.; Kane, J.

2014-11-01

The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of coremore » environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.« less

Progress report on the k{sub infinity} measurements of HTGR type lattices with the oscillation technique at zero reactivity; Rapport d'advancement sur les mesures de K {sub infinite} des reseaux du type HTGR avec la technique de l'oscillateur a'reactivite' nulle (in French)

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1971-01-15

The principles of measuring k {sub infinity} for a HTGR lattice using the oscillation technique with zero reactivity were already presented at the ''9th reactor physics meeting of countries participating in the Dragon project''. A brief summary of the essential characteristics of the experiment is followed by a status report on present work.

MODELING THE AMBIENT CONDITION EFFECTS OF AN AIR-COOLED NATURAL CIRCULATION SYSTEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui; Lisowski, Darius D.; Bucknor, Matthew

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performancemore » of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system. Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STARCCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.« less

Next Generation Nuclear Plant Defense-in-Depth Approach

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edward G. Wallace; Karl N. Fleming; Edward M. Burns

2009-12-01

The purpose of this paper is to (1) document the definition of defense-in-depth and the pproach that will be used to assure that its principles are satisfied for the NGNP project and (2) identify the specific questions proposed for preapplication discussions with the NRC. Defense-in-depth is a safety philosophy in which multiple lines of defense and conservative design and evaluation methods are applied to assure the safety of the public. The philosophy is also intended to deliver a design that is tolerant to uncertainties in knowledge of plant behavior, component reliability or operator performance that might compromise safety. This papermore » includes a review of the regulatory foundation for defense-in-depth, a definition of defense-in-depth that is appropriate for advanced reactor designs based on High Temperature Gas-cooled Reactor (HTGR) technology, and an explanation of how this safety philosophy is achieved in the NGNP.« less

Hydrogen production from coal using a nuclear heat source

NASA Technical Reports Server (NTRS)

Quade, R. N.

1976-01-01

A strong candidate for hydrogen production in the intermediate time frame of 1985 to 1995 is a coal-based process using a high-temperature gas-cooled reactor (HTGR) as a heat source. Expected process efficiencies in the range of 60 to 70% are considerably higher than all other hydrogen production processes except steam reforming of a natural gas. The process involves the preparation of a coal liquid, hydrogasification of that liquid, and steam reforming of the resulting gaseous or light liquid product. A study showing process efficiency and cost of hydrogen vs nuclear reactor core outlet temperature has been completed, and shows diminishing returns at process temperatures above about 1500 F. A possible scenario combining the relatively abundant and low-cost Western coal deposits with the Gulf Coast hydrogen users is presented which provides high-energy density transportation utilizing coal liquids and uranium.

Uranium extraction from TRISO-coated fuel particles using supercritical CO2 containing tri-n-butyl phosphate.

PubMed

Zhu, Liyang; Duan, Wuhua; Xu, Jingming; Zhu, Yongjun

2012-11-30

High-temperature gas-cooled reactors (HTGRs) are advanced nuclear systems that will receive heavy use in the future. It is important to develop spent nuclear fuel reprocessing technologies for HTGR. A new method for recovering uranium from tristructural-isotropic (TRISO-) coated fuel particles with supercritical CO(2) containing tri-n-butyl phosphate (TBP) as a complexing agent was investigated. TRISO-coated fuel particles from HTGR fuel elements were first crushed to expose UO(2) pellet fuel kernels. The crushed TRISO-coated fuel particles were then treated under O(2) stream at 750°C, resulting in a mixture of U(3)O(8) powder and SiC shells. The conversion of U(3)O(8) into solid uranyl nitrate by its reaction with liquid N(2)O(4) in the presence of a small amount of water was carried out. Complete conversion was achieved after 60 min of reaction at 80°C, whereas the SiC shells were not converted by N(2)O(4). Uranyl nitrate in the converted mixture was extracted with supercritical CO(2) containing TBP. The cumulative extraction efficiency was above 98% after 20 min of online extraction at 50°C and 25 MPa, whereas the SiC shells were not extracted by TBP. The results suggest an attractive strategy for reprocessing spent nuclear fuel from HTGR to minimize the generation of secondary radioactive waste. Copyright © 2012 Elsevier B.V. All rights reserved.

Methods for making a porous nuclear fuel element

DOEpatents

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

2014-12-30

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.

PHISICS/RELAP5-3D RESULTS FOR EXERCISES II-1 AND II-2 OF THE OECD/NEA MHTGR-350 BENCHMARK

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, Gerhard

2016-03-01

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) High-Temperature Gas-Cooled Reactor (HTGR) Methods group currently leads the Modular High-Temperature Gas-Cooled Reactor (MHTGR) 350 benchmark. The benchmark consists of a set of lattice-depletion, steady-state, and transient problems that can be used by HTGR simulation groups to assess the performance of their code suites. The paper summarizes the results obtained for the first two transient exercises defined for Phase II of the benchmark. The Parallel and Highly Innovative Simulation for INL Code System (PHISICS), coupled with the INL system code RELAP5-3D, was used to generate the results for the Depressurized Conductionmore » Cooldown (DCC) (exercise II-1a) and Pressurized Conduction Cooldown (PCC) (exercise II-2) transients. These exercises require the time-dependent simulation of coupled neutronics and thermal-hydraulics phenomena, and utilize the steady-state solution previously obtained for exercise I-3 of Phase I. This paper also includes a comparison of the benchmark results obtained with a traditional system code “ring” model against a more detailed “block” model that include kinetics feedback on an individual block level and thermal feedbacks on a triangular sub-mesh. The higher spatial fidelity that can be obtained by the block model is illustrated with comparisons of the maximum fuel temperatures, especially in the case of natural convection conditions that dominate the DCC and PCC events. Differences up to 125 K (or 10%) were observed between the ring and block model predictions of the DCC transient, mostly due to the block model’s capability of tracking individual block decay powers and more detailed helium flow distributions. In general, the block model only required DCC and PCC calculation times twice as long as the ring models, and it therefore seems that the additional development and calculation time required for the block model could be worth the gain that can be obtained in the spatial resolution« less

Uncertainty and Sensitivity Analyses of a Pebble Bed HTGR Loss of Cooling Event

DOE PAGES

Strydom, Gerhard

2013-01-01

The Very High Temperature Reactor Methods Development group at the Idaho National Laboratory identified the need for a defensible and systematic uncertainty and sensitivity approach in 2009. This paper summarizes the results of an uncertainty and sensitivity quantification investigation performed with the SUSA code, utilizing the International Atomic Energy Agency CRP 5 Pebble Bed Modular Reactor benchmark and the INL code suite PEBBED-THERMIX. Eight model input parameters were selected for inclusion in this study, and after the input parameters variations and probability density functions were specified, a total of 800 steady state and depressurized loss of forced cooling (DLOFC) transientmore » PEBBED-THERMIX calculations were performed. The six data sets were statistically analyzed to determine the 5% and 95% DLOFC peak fuel temperature tolerance intervals with 95% confidence levels. It was found that the uncertainties in the decay heat and graphite thermal conductivities were the most significant contributors to the propagated DLOFC peak fuel temperature uncertainty. No significant differences were observed between the results of Simple Random Sampling (SRS) or Latin Hypercube Sampling (LHS) data sets, and use of uniform or normal input parameter distributions also did not lead to any significant differences between these data sets.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James; Bayless, Paul; Strydom, Gerhard

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.more » 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.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bostelmann, Friederike; Strydom, Gerhard; Reitsma, Frederik

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

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

DOE PAGES

Bostelmann, Friederike; Strydom, Gerhard; Reitsma, Frederik; ...

2016-01-11

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

The Use of Thorium within the Nuclear Power Industry - 13472

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Keith

2013-07-01

Thorium is 3 to 4 times more abundant than uranium and is widely distributed in nature as an easily exploitable resource in many countries. Unlike natural uranium, which contains ∼0.7% fissile {sup 235}U isotope, natural thorium does not contain any fissile material and is made up of the fertile {sup 232}Th isotope only. Therefore thorium and thorium-based fuel as metal, oxide or carbide, has been utilized in combination with fissile {sup 235}U or {sup 239}Pu in nuclear research and power reactors for conversion to fissile {sup 233}U, thereby enlarging fissile material resources. During the pioneering years of nuclear energy, frommore » the mid 1950's to mid 1970's, there was considerable interest worldwide to develop thorium fuels and fuel cycles in order to supplement uranium reserves. Thorium fuels and fuel cycles are particularly relevant to countries having large thorium deposits but very limited uranium reserves for their long term nuclear power programme. The feasibility of thorium utilization in high temperature gas cooled reactors (HTGR), light water reactors (LWR), pressurized heavy water reactors (PHWRs), liquid metal cooled fast breeder reactors (LMFBR) and molten salt breeder reactors (MSBR) were demonstrated. The initial enthusiasm for thorium fuels and fuel cycles was not sustained among the developing countries later, due to new discovery of uranium deposits and their improved availability. However, in recent times, the need for proliferation-resistance, longer fuel cycles, higher burnup, and improved waste form characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material has led to renewed interest in thorium-based fuels and fuel cycles. (authors)« less

PIE on Safety-Tested AGR-1 Compact 5-1-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactormore » (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.« less

Effect of carbon ion irradiation on Ag diffusion in SiC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leng, Bin; Ko, Hyunseok; Gerczak, Tyler J.

Transport of Ag fission product through the silicon-carbide (SiC) diffusion barrier layer in TRISO fuel particles is of considerable interest given the application of this fuel type in high temperature gas-cooled reactor (HTGR) and other future reactor concepts. The reactor experiments indicate that radiation may play an important role in release of Ag; however so far the isolated effect of radiation on Ag diffusion has not been investigated in controlled laboratory experiments. In this study, we investigate the diffusion couples of Ag and polycrystalline 3C–SiC, as well as Ag and single crystalline 4H–SiC samples before and after irradiation with Cmore » 2+ ions. The diffusion couple samples were exposed to temperatures of 1500 °C, 1535 °C, and 1569 °C, and the ensuing diffusion profiles were analyzed by secondary ion mass spectrometry (SIMS). We found that diffusion coefficients calculated from these measurements indicate that Ag diffusion was greatly enhanced by carbon irradiation due to a combined effect of radiation damage on diffusion and the presence of grain boundaries in polycrystalline SiC samples.« less

Effect of carbon ion irradiation on Ag diffusion in SiC

DOE PAGES

Leng, Bin; Ko, Hyunseok; Gerczak, Tyler J.; ...

2015-11-14

Transport of Ag fission product through the silicon-carbide (SiC) diffusion barrier layer in TRISO fuel particles is of considerable interest given the application of this fuel type in high temperature gas-cooled reactor (HTGR) and other future reactor concepts. The reactor experiments indicate that radiation may play an important role in release of Ag; however so far the isolated effect of radiation on Ag diffusion has not been investigated in controlled laboratory experiments. In this study, we investigate the diffusion couples of Ag and polycrystalline 3C–SiC, as well as Ag and single crystalline 4H–SiC samples before and after irradiation with Cmore » 2+ ions. The diffusion couple samples were exposed to temperatures of 1500 °C, 1535 °C, and 1569 °C, and the ensuing diffusion profiles were analyzed by secondary ion mass spectrometry (SIMS). We found that diffusion coefficients calculated from these measurements indicate that Ag diffusion was greatly enhanced by carbon irradiation due to a combined effect of radiation damage on diffusion and the presence of grain boundaries in polycrystalline SiC samples.« less

Design and evaluation of a nondestructive fissile assay device for HTGR fuel samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

McNeany, S. R.; Knoll, R. W.; Jenkins, J. D.

1979-02-01

Nondestructive assay of fissile material plays an important role in nuclear fuel processing facilities. Information for product quality control, plant criticality safety, and nuclear materials accountability can be obtained from assay devices. All of this is necessary for a safe, efficient, and orderly operation of a production plant. Presented here is a design description and an operational evaluation of a device developed to nondestructively assay small samples of High-Temperature Gas-Cooled Reactor (HTGR) fuel. The measurement technique employed consists in thermal-neutron irradiation of a sample followed by pneumatic transfer to a high-efficiency neutron detector where delayed neutrons are counted. In general,more » samples undergo several irradiation and count cycles during a measurement. The total number of delayed-neutron counts accumulated is translated into grams of fissile mass through comparison with the counts accumulated in an identical irradiation and count sequence of calibration standards. Successful operation of the device through many experiments over a one-year period indicates high operational reliability. Tests of assay precision show this to be better than 0.25% for measurements of 10 min. Assay biases may be encountered if calibration standards are not representative of unknown samples, but reasonable care in construction and control of standards should lead to no more than 0.2% bias in the measurements. Nondestructive fissile assay of HTGR fuel samples by thermal-neutron irradiation and delayed-neutron detection has been demonstrated to be a rapid and accurate analysis technique. However, careful attention and control must be given to calibration standards to see that they remain representative of unknown samples.« less

Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

Setiadipura, T.; Suwoto; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Gerczak, Tyler J.; Morris, Robert Noel

2016-04-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in themore » Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.« less

A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petti, David Andrew

2017-04-01

Modular high temperature gas-cooled reactor (HTGR) designs were developed to provide natural safety, which prevents core damage under all licensing basis events. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. The required level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to source terms for other reactor types and allows a graded approach to emergency planning and the potential elimination of the need for evacuation and sheltering beyond a small exclusion area. Achieving this level, however,more » is predicated on exceptionally high coated-particle fuel fabrication quality and excellent performance under normal operation and accident conditions. The design goal of modular HTGRs is to meet the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) for offsite dose at the Exclusion Area Boundary (EAB). To achieve this, the reactor design concepts require a level of fuel integrity that is far better than that achieved for all prior U.S.-manufactured tristructural isotropic (TRISO) coated particle fuel.« less

Operation and postirradiation examination of ORR capsule OF-2: accelerated testing of HTGR fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiegs, T.N.; Thoms, K.R.

1979-03-01

Irradiation capsule OF-2 was a test of High-Temperature Gas-Cooled Reactor fuel types under accelerated irradiation conditions in the Oak Ridge Research Reactor. The results showed good irradiation performance of Triso-coated weak-acid-resin fissile particles and Biso-coated fertile particles. These particles had been coated by a fritted gas distributor in the 0.13-m-diam furnace. Fast-neutron damage (E > 0.18 MeV) and matrix-particle interaction caused the outer pyrocarbon coating on the Triso-coated particles to fail. Such failure depended on the optical anisotropy, density, and open porosity of the outer pyrocarbon coating, as well as on the coke yield of the matrix. Irradiation of specimensmore » with values outside prescribed limits for these properties increased the failure rate of their outer pyrocarbon coating. Good irradiation performance was observed for weak-acid-resin particles with conversions in the range from 15 to 75% UC/sub 2/.« less

NGNP Project 2011 Status and Path Forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

L.E. Demick

2011-12-01

High Temperature Gas Reactor (HTGR) technology can play an important role in the United States’ energy future by extending the use of nuclear energy for non-electricity energy production missions as well as continuing to provide a considerable base load electric power generation capability. Extending nuclear energy into the industrial and transportation sectors through the co-production of process heat and electricity provides safe and reliable energy for these sectors in an environmentally responsible manner. The safety case for the modular HTGR provides a substantial improvement in nuclear plant safety for the protection of the public and the environment, and supports collocationmore » of the HTGR with major industrial facilities. The NGNP Project at the Idaho National Laboratory has been working toward an objective of commercializing the HTGR technology under DOE direction since 2006. The Project is undergoing a quantum shift in direction and scope as a result of recent DOE decisions. This paper summarizes where the Project has been, where it is at the time of this writing and what is needed in future activities to commercialize HTGR technology.« less

High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.

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, simplemore » 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 allows the owner of the facility to select the preferred designation, and that either designation can be acceptable.« less

Comparison between instrumented precracked Charpy and compact specimen tests of carbon steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nanstad, R.K.

1980-01-01

The General Atomic Company High Temperature Gas-Cooled Reactor (HTGR) is housed within a prestressed concrete reactor vessel (PCRV). Various carbon steel structural members serve as closures at penetrations in the vessel. A program of testing and evaluation is underway to determine the need for reference fracture toughness (K/sub IR/) and indexing procedures for these materials as described in Appendix G to Section III, ASME Code for light water reactor steels. The materials of interest are carbon steel forgings (SA508, Class 1) and plates (SA537, Classes 1 and 2) as well as weldments of these steels. The fracture toughness behavior ismore » characterized with instrumented precracked Charpy V-votch specimens (PCVN) - slow-bend and dynamic - and compact specimens (10-mm and 25-mm thicknesses) using both linear elastic (ASTM E399) and elastic-plastic (equivalent Energy and J-Integral) analytical procedures. For the dynamic PCVN tests, force-time traces are analyzed according to the procedures of the Pressure Vessel Research Council (PVRC)/Metal Properties Council (MPC). Testing and analytical procedures are discussed and PCVN results are compared to those obtained with compact specimens.« less

Development of a coolant channel helium and nitrogen gas ratio sensor for a high temperature gas reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cadell, S. R.; Woods, B. G.

2012-07-01

To measure the changing gas composition of the coolant during a postulated High Temperature Gas Reactor (HTGR) accident, an instrument is needed. This instrument must be compact enough to measure the ratio of the coolant versus the break gas in an individual coolant channel. This instrument must minimally impact the fluid flow and provide for non-direct signal routing to allow minimal disturbance to adjacent channels. The instrument must have a flexible geometry to allow for the measurement of larger volumes such as in the upper or lower plenum of a HTGR. The instrument must be capable of accurately functioning throughmore » the full operating temperature and pressure of a HTGR. This instrument is not commercially available, but a literature survey has shown that building off of the present work on Capacitance Sensors and Cross-Capacitors will provide a basis for the development of the desired instrument. One difficulty in developing and instrument to operate at HTGR temperatures is acquiring an electrical conductor that will not melt at 1600 deg. C. This requirement limits the material selection to high temperature ceramics, graphite, and exotic metals. An additional concern for the instrument is properly accounting for the thermal expansion of both the sensing components and the gas being measured. This work covers the basic instrument overview with a thorough discussion of the associated uncertainty in making these measurements. (authors)« less

High-temperature Gas Reactor (HTGR)

NASA Astrophysics Data System (ADS)

Abedi, Sajad

2011-05-01

General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, Dane F.

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],more » 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 O 2 cannot be ignored, especially for the FHR, in which environment the product, SiO 2, 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.« less

NRC ARDC Guidance Support Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holbrook, Mark R.

This report provides a summary that reflects the progress and status of proposed regulatory design criteria for advanced non-light water reactor (LWR) designs in accordance with the Level 3 milestone M3AT-17IN2001013 in work package AT-17IN200101. These criteria have been designated as advanced reactor design criteria (ARDC) and they provide guidance to future applicants for addressing the general design criteria (GDC) that are currently applied specifically to LWR designs. This report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of ARDC regulatory guidance for sodium fast reactor (SFR) andmore » modular high-temperature gas-cooled reactor (HTGR) designs. Status Report Organization: Section 2 discusses the origin of the GDC and their application to LWRs. Section 3 addresses the objective of this initiative and how it benefits the advanced non-LWR reactor vendors. Section 4 discusses the scope and structure of the initiative. Section 5 provides background on the U.S. Department of Energy (DOE) ARDC team’s original development of the proposed ARDC that were submitted to the NRC for consideration. Section 6 provides a summary of recent ARDC Phase 2 activities. Appendices A through E document the DOE ARDC team’s public comments on various sections of the NRC’s draft regulatory guide DG–1330, “Guidance for Developing Principal Design Criteria for Non-Light Water Reactors.”« less

COL Application Content Guide for HTGRs: Revision to RG 1.206, Part 1 - Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wayne Moe

2012-08-01

A combined license (COL) application is required by the Nuclear Regulatory Commission (NRC) for all proposed nuclear plants. The information requirements for a COL application are set forth in 10 CFR 52.79, “Contents of Applications; Technical Information in Final Safety Analysis Report.” An applicant for a modular high temperature gas-cooled reactor (HTGR) must develop and submit for NRC review and approval a COL application which conforms to these requirements. The technical information necessary to allow NRC staff to evaluate a COL application and resolve all safety issues related to a proposed nuclear plant is detailed and comprehensive. To this, Regulatorymore » Guide (RG) 1.206, “Combined License Applications for Nuclear Power Plants” (LWR Edition), was developed to assist light water reactor (LWR) applicants in incorporating and effectively formatting required information for COL application review (Ref. 1). However, the guidance prescribed in RG 1.206 presumes a LWR design proposal consistent with the systems and functions associated with large LWR power plants currently operating under NRC license.« less

Turbomachinery design considerations for the nuclear HTGR-GT power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDonald, C.F.; Smith, M.J.

1979-11-01

For several years, design studies have been under way in the USA on a nuclear closed-cycle gas turbine plant (HTGR-GT). Design aspects of the helium turbomachine portion of these studies are presented. Gas dynamic and mechanical design considerations are presented for helium turbomachines in the 400-MW(e) (non-intercooled) and 600-MW(e) (intercooled) power range. Design of the turbomachine is a key element in the overall power plant program effort, which is currently directed toward the selection of a reference HTGR-GT commercial plant configuration for the US utility market. A conservative design approach has been emphasized to provide maximum safety and durability. Themore » studies presented for the integrated plant concept outline the necessary close working relationship between the reactor primary system and turbomachine designers.« less

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franken, Daniel; Gould, Daniel; Jain, Prashant K.

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

DOE PAGES

Franken, Daniel; Gould, Daniel; Jain, Prashant K.; ...

2017-09-21

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

DOE PAGES

Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; ...

2016-08-15

Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representativemore » of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.« less

AGC-2 Irradiation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohrbaugh, David Thomas; Windes, William; Swank, W. David

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a completemore » properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphites is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.« less

Impact of New Nuclear Data Libraries on Small Sized Long Life CANDLE HTGR Design Parameters

NASA Astrophysics Data System (ADS)

Liem, Peng Hong; Hartanto, Donny; Tran, Hoai Nam

2017-01-01

The impact of new evaluated nuclear data libraries (JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1) on the core characteristics of small-sized long-life CANDLE High Temperature Gas-Cooled Reactors (HTGRs) with uranium and thorium fuel cycles was investigated. The most important parameters of the CANDLE core characteristics investigated here covered (1) infinite multiplication factor of the fresh fuel containing burnable poison, (2) the effective multiplication factor of the equilibrium core, (3) the moving velocity of the burning region, (4) the attained discharge burnup, and (5) the maximum power density. The reference case was taken from the current JENDL-3.3 results. For the uranium fuel cycle, the impact of the new libraries was small, while significant impact was found for thorium fuel cycle. The findings indicated the needs of more accurate nuclear data libraries for nuclides involved in thorium fuel cycle in the future.

Metallic impurities-silicon carbide interaction in HTGR fuel particles

NASA Astrophysics Data System (ADS)

Minato, Kazuo; Ogawa, Toru; Kashimura, Satoru; Fukuda, Kousaku; Shimizu, Michio; Tayama, Yoshinobu; Takahashi, Ishio

1990-12-01

Corrosion of the coating layers of silicon carbide (SiC) by metallic impurities was observed in irradiated Triso-coated uranium dioxide particles for high temperature gas-cooled reactors with an optical microscope and an electron probe micro-analyzer. The SiC layers were attacked from the outside of the particles. The main element observed in the corroded areas was iron, but sometimes iron and nickel were found. These elements must have been contained as impurities in the graphite matrix in which the coated particles were dispersed. Since these elements are more stable thermodynamically in the presence of SiC than in the presence of graphite at irradiation temperatures, they were transferred to the SiC layer to form more stable silicides. During fuel manufacturing processes, intensive care should be taken to prevent the fuel from being contaminated with those elements which react with SiC.

Efficiency and cost advantages of an advanced-technology nuclear electrolytic hydrogen-energy production facility

NASA Technical Reports Server (NTRS)

Donakowski, T. D.; Escher, W. J. D.; Gregory, D. P.

1977-01-01

The concept of an advanced-technology (viz., 1985 technology) nuclear-electrolytic water electrolysis facility was assessed for hydrogen production cost and efficiency expectations. The facility integrates (1) a high-temperature gas-cooled nuclear reactor (HTGR) operating a binary work cycle, (2) direct-current (d-c) electricity generation via acyclic generators, and (3) high-current-density, high-pressure electrolyzers using a solid polymer electrolyte (SPE). All subsystems are close-coupled and optimally interfaced for hydrogen production alone (i.e., without separate production of electrical power). Pipeline-pressure hydrogen and oxygen are produced at 6900 kPa (1000 psi). We found that this advanced facility would produce hydrogen at costs that were approximately half those associated with contemporary-technology nuclear electrolysis: $5.36 versus $10.86/million Btu, respectively. The nuclear-heat-to-hydrogen-energy conversion efficiency for the advanced system was estimated as 43%, versus 25% for the contemporary system.

Development Program of IS Process Pilot Test Plant for Hydrogen Production With High-Temperature Gas-Cooled Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin Iwatsuki; Atsuhiko Terada; Hiroyuki Noguchi

2006-07-01

At the present time, we are alarmed by depletion of fossil energy and effects on global environment such as acid rain and global warming, because our lives depend still heavily on fossil energy. So, it is universally recognized that hydrogen is one of the best energy media and its demand will be increased greatly in the near future. In Japan, the Basic Plan for Energy Supply and Demand based on the Basic Law on Energy Policy Making was decided upon by the Cabinet on 6 October, 2003. In the plan, efforts for hydrogen energy utilization were expressed as follows; hydrogenmore » is a clean energy carrier without carbon dioxide (CO{sub 2}) emission, and commercialization of hydrogen production system using nuclear, solar and biomass, not fossil fuels, is desired. However, it is necessary to develop suitable technology to produce hydrogen without CO{sub 2} emission from a view point of global environmental protection, since little hydrogen exists naturally. Hydrogen production from water using nuclear energy, especially the high-temperature gas-cooled reactor (HTGR), is one of the most attractive solutions for the environmental issue, because HTGR hydrogen production by water splitting methods such as a thermochemical iodine-sulfur (IS) process has a high possibility to produce hydrogen effectively and economically. The Japan Atomic Energy Agency (JAEA) has been conducting the HTTR (High-Temperature Engineering Test Reactor) project from the view to establishing technology base on HTGR and also on the IS process. In the IS process, raw material, water, is to be reacted with iodine (I{sub 2}) and sulfur dioxide (SO{sub 2}) to produce hydrogen iodide (HI) and sulfuric acid (H{sub 2}SO{sub 4}), the so-called Bunsen reaction, which are then decomposed endo-thermically to produce hydrogen (H{sub 2}) and oxygen (O{sub 2}), respectively. Iodine and sulfur dioxide produced in the decomposition reactions can be used again as the reactants in the Bunsen reaction. In JAEA, continuous hydrogen production was demonstrated with the hydrogen production rate of about 30 NL/hr for one week using a bench-scale test apparatus made of glass. Based on the test results and know-how obtained through the bench-scale tests, a pilot test plant that can produce hydrogen of about 30 Nm{sup 3}/hr is being designed. The test plant will be fabricated with industrial materials such as glass coated steel, SiC ceramics etc, and operated under high pressure condition up to 2 MPa. The test plant will consist of a IS process plant and a helium gas (He) circulation facility (He loop). The He loop can simulate HTTR operation conditions, which consists of a 400 kW-electric heater for He hating, a He circulator and a steam generator working as a He cooler. In parallel to the design study, key components of the IS process such as the sulfuric acid (H{sub 2}SO{sub 4}) and the sulfur trioxide (SO{sub 3}) decomposers working under-high temperature corrosive environments have been designed and test-fabricated to confirm their fabricability. Also, other R and D's are under way such as corrosion, processing of HIx solutions. This paper describes present status of these activities. (authors)« less

Summary of Planned Implementation for the HTGR Lessons Learned Applicable to the NGNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ian Mckirdy

2011-09-01

This document presents a reconciliation of the lessons learned during a 2010 comprehensive evaluation of pertinent lessons learned from past and present high temperature gas-cooled reactors that apply to the Next Generation Nuclear Plant Project along with current and planned activities. The data used are from the latest Idaho National Laboratory research and development plans, the conceptual design report from General Atomics, and the pebble bed reactor technology readiness study from AREVA. Only those lessons related to the structures, systems, and components of the Next Generation Nuclear Plant (NGNP), as documented in the recently updated lessons learned report are addressed.more » These reconciliations are ordered according to plant area, followed by the affected system, subsystem, or component; lesson learned; and finally an NGNP implementation statement. This report (1) provides cross references to the original lessons learned document, (2) describes the lesson learned, (3) provides the current NGNP implementation status with design data needs associated with the lesson learned, (4) identifies the research and development being performed related to the lesson learned, and (5) summarizes with a status of how the lesson learned has been addressed by the NGNP Project.« less

FINAL PROJECT REPORT - EVALUATION AND TESTING OF HTGR REACTOR BUILDING RESPONSE TO DEPRESSURIZATION ACCIDENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

ALLIANCE LIMITED, NGNP INDUSTRY

This report provides a description of the project, summarizes each phase of the project, and ends with project conclusions. In addition, the report contains a descriptive index of the technical reports generated during the course of the project.

Nuclear fuels status

NASA Technical Reports Server (NTRS)

Kania, Michael

1991-01-01

A discussion on coated particle fuel performance from a modular High Temperature Gas Reactor (HTGR) is presented along with experimental results. The following topics are covered: (1) the coated particle fuel concept; (2) the functional requirements; (3) performance limiting mechanisms; (4) fuel performance; and (5) methods/techniques for characterizing performance.

ICP-MS measurement of iodine diffusion in IG-110 graphite for HTGR/VHTR

NASA Astrophysics Data System (ADS)

Carter, L. M.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

2016-05-01

Graphite functions as a structural material and as a barrier to fission product release in HTGR/VHTR designs, and elucidation of transport parameters for fission products in reactor-grade graphite is thus required for reactor source terms calculations. We measured iodine diffusion in spheres of IG-110 graphite using a release method based on Fickain diffusion kinetics. Two sources of iodine were loaded into the graphite spheres; molecular iodine (I2) and cesium iodide (CsI). Measurements of the diffusion coefficient were made over a temperature range of 873-1293 K. We have obtained the following Arrhenius expressions for iodine diffusion:DI , CsI infused =(6 ×10-12 2/s) exp(30,000 J/mol RT) And,DI , I2 infused =(4 ×10-10 m2/s) exp(-11,000 J/mol RT ) The results indicate that iodine diffusion in IG-110 graphite is not well-described by Fickan diffusion kinetics. To our knowledge, these are the first measurements of iodine diffusion in IG-110 graphite.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rouxelin, Pascal Nicolas; Strydom, Gerhard

Best-estimate plus uncertainty analysis of reactors is replacing the traditional conservative (stacked uncertainty) method for safety and licensing analysis. To facilitate uncertainty analysis applications, a comprehensive approach and methodology must be developed and applied. High temperature gas cooled reactors (HTGRs) have several features that require techniques not used in light-water reactor analysis (e.g., coated-particle design and large graphite quantities at high temperatures). The International Atomic Energy Agency has therefore launched the Coordinated Research Project on HTGR Uncertainty Analysis in Modeling to study uncertainty propagation in the HTGR analysis chain. The benchmark problem defined for the prismatic design is represented bymore » the General Atomics Modular HTGR 350. The main focus of this report is the compilation and discussion of the results obtained for various permutations of Exercise I 2c and the use of the cross section data in Exercise II 1a of the prismatic benchmark, which is defined as the last and first steps of the lattice and core simulation phases, respectively. The report summarizes the Idaho National Laboratory (INL) best estimate results obtained for Exercise I 2a (fresh single-fuel block), Exercise I 2b (depleted single-fuel block), and Exercise I 2c (super cell) in addition to the first results of an investigation into the cross section generation effects for the super-cell problem. The two dimensional deterministic code known as the New ESC based Weighting Transport (NEWT) included in the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1.2 package was used for the cross section evaluation, and the results obtained were compared to the three dimensional stochastic SCALE module KENO VI. The NEWT cross section libraries were generated for several permutations of the current benchmark super-cell geometry and were then provided as input to the Phase II core calculation of the stand alone neutronics Exercise II 1a. The steady state core calculations were simulated with the INL coupled-code system known as the Parallel and Highly Innovative Simulation for INL Code System (PHISICS) and the system thermal-hydraulics code known as the Reactor Excursion and Leak Analysis Program (RELAP) 5 3D using the nuclear data libraries previously generated with NEWT. It was observed that significant differences in terms of multiplication factor and neutron flux exist between the various permutations of the Phase I super-cell lattice calculations. The use of these cross section libraries only leads to minor changes in the Phase II core simulation results for fresh fuel but shows significantly larger discrepancies for spent fuel cores. Furthermore, large incongruities were found between the SCALE NEWT and KENO VI results for the super cells, and while some trends could be identified, a final conclusion on this issue could not yet be reached. This report will be revised in mid 2016 with more detailed analyses of the super-cell problems and their effects on the core models, using the latest version of SCALE (6.2). The super-cell models seem to show substantial improvements in terms of neutron flux as compared to single-block models, particularly at thermal energies.« less

Criticality calculations of the Very High Temperature reactor Critical Assembly benchmark with Serpent and SCALE/KENO-VI

DOE PAGES

Bostelmann, Friederike; Hammer, Hans R.; Ortensi, Javier; ...

2015-12-30

Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, criticality calculations of the Very High Temperature Critical Assembly experiment were performed as the validation reference to the prismatic MHTGR-350 lattice calculations. Criticality measurements performed at several temperature points at this Japanese graphite-moderated facility were recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and represent one of the few data sets available for the validation of HTGR lattice physics. Here, this work compares VHTRC criticality simulations utilizing the Monte Carlo codes Serpent and SCALE/KENO-VI. Reasonable agreement was found between Serpent andmore » KENO-VI, but only the use of the latest ENDF cross section library release, namely the ENDF/B-VII.1 library, led to an improved match with the measured data. Furthermore, the fourth beta release of SCALE 6.2/KENO-VI showed significant improvements from the current SCALE 6.1.2 version, compared to the experimental values and Serpent.« less

Role of nuclear energy to a future society of shortage of energy resources and global warming

NASA Astrophysics Data System (ADS)

Saito, Shinzo

2010-03-01

Human society entered into the society of large energy consumption since the industrial revolution and consumes more than 10 billion tons of oil equivalent energy a year in the world in the present time, in which over 80% is provided by fossil fuels such as coal, oil and natural gas. Total energy consumption is foreseen to increase year by year from now on due to significant economical and population growth in the developing countries such as China and India. However, fossil fuel resources are limited with conventional crude oil estimated to last about 40 years, and it is said that the peak oil production time has come now. On the other hand, global warming due to green house gases (GHG) emissions, especially carbon dioxide, has become a serious issue. Nuclear energy plays an important role as means to resolve energy security and global warming issues. Four hundred twenty-nine nuclear power plants are operating world widely producing 16% of the total electric power with total plant capacity of 386 GWe without emission of CO 2 as of 2006. It is estimated that another 250 GWe nuclear power is needed to keep the same level contribution of electricity generation in 2030. On the other hand, the Japan Atomic Energy Research Institute (JAERI) developed the very high temperature gas-cooled reactor (HTGR) named high temperature gas-cooled engineering test reactor (HTTR) and carbon free hydrogen production process (IS process). Nuclear energy utilization will surely widen in, not only electricity generation, but also various industries such as steel making, chemical industries, together with hydrogen production for transportation by introduction of HTGRs. The details of development of the HTTR and IS process are also described.

Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jason M. Harp; Paul A. Demkowicz

2014-10-01

In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10 -4 to 10 -5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materialsmore » is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.« less

Safety philosophy of gas turbine high temperature reactor (GTHTR300)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

2002-07-01

Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Majormore » features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)« less

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

NASA Technical Reports Server (NTRS)

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.

Water NSTF Design, Instrumentation, and Test Planning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, Darius D.; Gerardi, Craig D.; Hu, Rui

The following report serves as a formal introduction to the water-based Natural convection Shutdown heat removal Test Facility (NSTF) program at Argonne. Since 2005, this US Department of Energy (DOE) sponsored program has conducted large scale experimental testing to generate high-quality and traceable validation data for guiding design decisions of the Reactor Cavity Cooling System (RCCS) concept for advanced reactor designs. The most recent facility iteration, and focus of this report, is the operation of a 1/2 scale model of a water-RCCS concept. Several features of the NSTF prototype align with the conceptual design that has been publicly released formore » the AREVA 625 MWt SC-HTGR. The design of the NSTF also retains all aspects common to a fundamental boiling water thermosiphon, and thus is well poised to provide necessary experimental data to advance basic understanding of natural circulation phenomena and contribute to computer code validation. Overall, the NSTF program operates to support the DOE vision of aiding US vendors in design choices of future reactor concepts, advancing the maturity of codes for licensing, and ultimately developing safe and reliable reactor technologies. In this report, the top-level program objectives, testing requirements, and unique considerations for the water cooled test assembly are discussed, and presented in sufficient depth to support defining the program’s overall scope and purpose. A discussion of the proposed 6-year testing program is then introduced, which outlines the specific strategy and testing plan for facility operations. The proposed testing plan has been developed to meet the toplevel objective of conducting high-quality test operations that span across a broad range of single- and two-phase operating conditions. Details of characterization, baseline test cases, accident scenario, and parametric variations are provided, including discussions of later-stage test cases that examine the influence of geometric variations and off-normal configurations. The facility design follows, including as-built dimensions and specifications of the various mechanical and liquid systems, design choices for the test section, water storage tank, and network piping. Specifications of the instrumentation suite are then presented, along with specific information on performance windows, measurement uncertainties, and installation locations. Finally, descriptions of the control systems and heat removal networks are provided, which have been engineered to support precise quantification of energy balances and facilitate well-controlled test operations.« less

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

DOEpatents

Hunsbedt, Anstein; Busboom, Herbert J.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Passive cooling safety system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Indirect passive cooling system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1990-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element

DTIC Science & Technology

1990-06-01

long fuel elements, arranged to form a core , were analyzed for an up-power transient from 0 MWt to approximately 18 MWt. The simple model significantly...VARIATIONS IN FUEL ELEMENT GEOMETRY ............. 60 4.4 VARIATIONS IN THE MANNER OF TRANSIENT CONTROL ..... 62 4.5 CORE REPRESENTATION BY MULTIPLE FUEL ...the HTGR , however, the PBR packs small fuel particles between inner and outer retention elements, designated as frits. The PBR is appropriate for a

Status of Chronic Oxidation Studies of Graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Contescu, Cristian I.; Mee, Robert W.

Graphite will undergo extremely slow, but continuous oxidation by traces of moisture that will be present, albeit at very low levels, in the helium coolant of HTGR. This chronic oxidation may cause degradation of mechanical strength and thermal properties of graphite components if a porous oxidation layer penetrates deep enough in the bulk of graphite components during the lifetime of the reactor. The current research on graphite chronic oxidation is motivated by the acute need to understand the behavior of each graphite grade during prolonged exposure to high temperature chemical attack by moisture. The goal is to provide the elementsmore » needed to develop predictive models for long-time oxidation behavior of graphite components in the cooling helium of HTGR. The tasks derived from this goal are: (1) Oxidation rate measurements in order to determine and validate a comprehensive kinetic model suitable for prediction of intrinsic oxidation rates as a function of temperature and oxidant gas composition; (2) Characterization of effective diffusivity of water vapor in the graphite pore system in order to account for the in-pore transport of moisture; and (3) Development and validation of a predictive model for the penetration depth of the oxidized layer, in order to assess the risk of oxidation caused damage of particular graphite grades after prolonged exposure to the environment of helium coolant in HTGR. The most important and most time consuming of these tasks is the measurement of oxidation rates in accelerated oxidation tests (but still under kinetic control) and the development of a reliable kinetic model. This report summarizes the status of chronic oxidation studies on graphite, and then focuses on model development activities, progress of kinetic measurements, validation of results, and improvement of the kinetic models. Analysis of current and past results obtained with three grades of showed that the classical Langmuir-Hinshelwood model cannot reproduce all data collected so far. Starting from here we propose a modification of the LH model to include temperature activation of graphite surface as a Boltzmann activation function. The enhanced Boltzmann-Langmuir-Hinshelwood model (BLH) was tested successfully on three grades of graphite. The model is a robust, comprehensive mathematical function that allows better fitting of experimental results spanning a wide range of temperature and partial pressures of water vapor and hydrogen. However, the model did not fit satisfactorily the data extracted from the old report on graphite H-451 oxidation by water.« less

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

Liquid metal cooled nuclear reactors with passive cooling system

DOEpatents

Hunsbedt, Anstein; Fanning, Alan W.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, C.W.

1985-02-19

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, Charles W.

1987-01-01

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

Cooling system for a nuclear reactor

DOEpatents

Amtmann, Hans H.

1982-01-01

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Liquid metal cooled nuclear reactor plant system

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 1: Objectives, summary results and introduction

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The objective was to determine which reactor, conversion, and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. Specifically, the requirement was 10 megawatts for 5 years of full power operation and 10 years systems life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study. The concepts are: a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heat pipe and pumped tube-fin heat rejection; a lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator; a lithium cooled reactor with potassium Rankine turbine-alternator and heat pipe radiator; and a lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the lithium cooled incore thermionic reactor with heat pipe radiator.

THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Colley, J.R.

1962-12-01

The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Thermodynamic analysis of the efficiency of high-temperature steam electrolysis system for hydrogen production

NASA Astrophysics Data System (ADS)

Mingyi, Liu; Bo, Yu; Jingming, Xu; Jing, Chen

High-temperature steam electrolysis (HTSE), a reversible process of solid oxide fuel cell (SOFC) in principle, is a promising method for highly efficient large-scale hydrogen production. In our study, the overall efficiency of the HTSE system was calculated through electrochemical and thermodynamic analysis. A thermodynamic model in regards to the efficiency of the HTSE system was established and the quantitative effects of three key parameters, electrical efficiency (η el), electrolysis efficiency (η es), and thermal efficiency (η th) on the overall efficiency (η overall) of the HTSE system were investigated. Results showed that the contribution of η el, η es, η th to the overall efficiency were about 70%, 22%, and 8%, respectively. As temperatures increased from 500 °C to 1000 °C, the effect of η el on η overall decreased gradually and the η es effect remained almost constant, while the η th effect increased gradually. The overall efficiency of the high-temperature gas-cooled reactor (HTGR) coupled with the HTSE system under different conditions was also calculated. With the increase of electrical, electrolysis, and thermal efficiency, the overall efficiencies were anticipated to increase from 33% to a maximum of 59% at 1000 °C, which is over two times higher than that of the conventional alkaline water electrolysis.

Passive cooling system for top entry liquid metal cooled nuclear reactors

DOEpatents

Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

1992-01-01

A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 4: Concepts selection, conceptual designs, recommendations

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

A study was conducted by NASA Lewis Research Center for the Triagency SP-100 program office. The objective was to determine which reactor, conversion and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. The requirement was 10 megawatts for 5 years of full power operation and 10 years system life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study: (1) a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heatpipe and pumped tube fin rejection, (2) a Lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator,(3) a Lithium cooled reactor with a Potassium Rankine turbine-alternator and heat pipe radiator, and (4) a Lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the Lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the Lithium cooled incore thermionic reactor with heat pipe radiator.

Reactor water cleanup system

DOEpatents

Gluntz, D.M.; Taft, W.E.

1994-12-20

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

NASA Astrophysics Data System (ADS)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

TUNABLE IRRADIATION TESTBED

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Asner, David M.

PNNL has developed and continues to develop innovative methods for characterizing irradiated materials from nuclear reactors and particle accelerators for various clients and collaborators around the world. The continued development of these methods, in addition to the ability to perform unique scientific investigations of the effects of radiation on materials could be greatly enhanced with easy access to irradiation facilities. A Tunable Irradiation Testbed with customized targets (a 30 MeV, 1mA cyclotron or similar coupled to a unique target system) is shown to provide a much more flexible and cost-effective source of irradiating particles than a test reactor or isotopicmore » source. The configuration investigated was a single shielded building with multiple beam lines from a small, flexible, high flux irradiation source. Potential applications investigated were the characterization of radiation damage to materials applicable to advanced reactors, fusion reactor, legacy waste, (via neutron spectra tailored to HTGR, molten salt, LWR, LMR, fusion environments); 252Cf replacement; characterization of radiation damage to materials of interest to High Energy Physics to enable the neutrino program; and research into production of short lived isotopes for potential medical and other applications.« less

Issues and Potential Program on Denatured Fuel Utilization.

DTIC Science & Technology

1978-12-01

HTGR fuel develop - ment program ; 4. coated particles of (U,Th)02 have been extensively tested as potential HTGR fuels . A detailed summary of the...current scrap and waste treatment requirements. dBase case for all HTGR (Prismatic Fuel Element) cases based on data in "Summary Program Plan...Alternate Program for HTGR Fuel Recycle," April 11, 1975, Draft. 19 a --- AC8NCi09 The principal factors that result in a nominally-higher cost for

Integrating Nuclear Energy to Oilfield Operations – Two Case Studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eric P. Robertson; Lee O. Nelson; Michael G. McKellar

2011-11-01

Fossil fuel resources that require large energy inputs for extraction, such as the Canadian oil sands and the Green River oil shale resource in the western USA, could benefit from the use of nuclear power instead of power generated by natural gas combustion. This paper discusses the technical and economic aspects of integrating nuclear energy with oil sands operations and the development of oil shale resources. A high temperature gas reactor (HTGR) that produces heat in the form of high pressure steam (no electricity production) was selected as the nuclear power source for both fossil fuel resources. Both cases weremore » based on 50,000 bbl/day output. The oil sands case was a steam-assisted, gravity-drainage (SAGD) operation located in the Canadian oil sands belt. The oil shale development was an in-situ oil shale retorting operation located in western Colorado, USA. The technical feasibility of the integrating nuclear power was assessed. The economic feasibility of each case was evaluated using a discounted cash flow, rate of return analysis. Integrating an HTGR to both the SAGD oil sands operation and the oil shale development was found to be technically feasible for both cases. In the oil sands case, integrating an HTGR eliminated natural gas combustion and associated CO2 emissions, although there were still some emissions associated with imported electrical power. In the in situ oil shale case, integrating an HTGR reduced CO2 emissions by 88% and increased natural gas production by 100%. Economic viabilities of both nuclear integrated cases were poorer than the non-nuclear-integrated cases when CO2 emissions were not taxed. However, taxing the CO2 emissions had a significant effect on the economics of the non-nuclear base cases, bringing them in line with the economics of the nuclear-integrated cases. As we move toward limiting CO2 emissions, integrating non-CO2-emitting energy sources to the development of energy-intense fossil fuel resources is becoming increasingly important. This paper attempts to reduce the barriers that have traditionally separated fossil fuel development and application of nuclear power and to promote serious discussion of ideas about hybrid energy systems.« less

The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

NASA Astrophysics Data System (ADS)

Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

NASA Technical Reports Server (NTRS)

Nast, T. C.

1983-01-01

The utilization of solid hydrogen in space for sensor and instrument cooling is a very efficient technique for long term cooling or for cooling at high heat rates. The solid hydrogen can provide temperatures as low as 7 to 8 K to instruments. Vapor cooling is utilized to reduce parasitic heat inputs to the 7 to 8 K stage and is effective in providing intermediate cooling for instrument components operating at higher temperatures. The use of solid hydrogen in place of helium may lead to weight reductions as large as a factor of ten and an attendent reduction in system volume. The results of an investigation of a catalytic reactor for use with a solid hydrogen cooling system is presented. Trade studies were performed on several configurations of reactor to meet the requirements of high reactor efficiency with low pressure drop. Results for the selected reactor design are presented for both liquid hydrogen systems operating at near atmospheric pressure and the solid hydrogen cooler operating as low as 1 torr.

Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

DOEpatents

Hunsbedt, A.; Boardman, C.E.

1995-04-11

A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1995-01-01

A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

Design of conduction cooling system for a high current HTS DC reactor

NASA Astrophysics Data System (ADS)

Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

2017-07-01

A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki

2002-07-01

Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heatmore » removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)« less

System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

2004-03-15

Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 smore » after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.« less

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

User's manual for the Composite HTGR Analysis Program (CHAP-1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilbert, J.S.; Secker, P.A. Jr.; Vigil, J.C.

1977-03-01

CHAP-1 is the first release version of an HTGR overall plant simulation program with both steady-state and transient solution capabilities. It consists of a model-independent systems analysis program and a collection of linked modules, each representing one or more components of the HTGR plant. Detailed instructions on the operation of the code and detailed descriptions of the HTGR model are provided. Information is also provided to allow the user to easily incorporate additional component modules, to modify or replace existing modules, or to incorporate a completely new simulation model into the CHAP systems analysis framework.

METHOD OF FIXING NITROGEN FOR PRODUCING OXIDES OF NITROGEN

DOEpatents

Harteck, P.; Dondes, S.

1959-08-01

A method is described for fixing nitrogen from air by compressing the air, irradiating the compressed air in a nuclear reactor, cooling to remove NO/ sub 2/, compressing the cooled gas, further cooling to remove N/sub 2/O and recirculating the cooled compressed air to the reactor.

Assessment of Sensor Technologies for Advanced Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

NEUTRONIC REACTOR POWER PLANT

DOEpatents

Metcalf, H.E.

1962-12-25

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

Method and apparatus for enhancing reactor air-cooling system performance

DOEpatents

Hunsbedt, Anstein

1996-01-01

An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. 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 usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-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. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. 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 usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-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. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

SPOUTED BED DESIGN CONSIDERATIONS FOR COATED NUCLEAR FUEL PARTICLES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, Douglas W.

High Temperature Gas Cooled Reactors (HTGRs) are fueled with tristructural isotropic (TRISO) coated nuclear fuel particles embedded in a carbon-graphite fuel body. TRISO coatings consist of four layers of pyrolytic carbon and silicon carbide that are deposited on uranium ceramic fuel kernels (350µm – 500µm diameters) in a concatenated series of batch depositions. Each layer has dedicated functions such that the finished fuel particle has its own integral containment to minimize and control the release of fission products into the fuel body and reactor core. The TRISO coatings are the primary containment structure in the HTGR reactor and must havemore » very high uniformity and integrity. To ensure high quality TRISO coatings, the four layers are deposited by chemical vapor deposition (CVD) using high purity precursors and are applied in a concatenated succession of batch operations before the finished product is unloaded from the coating furnace. These depositions take place at temperatures ranging from 1230°C to 1550°C and use three different gas compositions, while the fuel particle diameters double, their density drops from 11.1 g/cm3 to 3.0 g/cm3, and the bed volume increases more than 8-fold. All this is accomplished without the aid of sight ports or internal instrumentation that could cause chemical contamination within the layers or mechanical damage to thin layers in the early stages of each layer deposition. The converging section of the furnace retort was specifically designed to prevent bed stagnation that would lead to unacceptably high defect fractions and facilitate bed circulation to avoid large variability in coating layer dimensions and properties. The gas injection nozzle was designed to protect precursor gases from becoming overheated prior to injection, to induce bed spouting and preclude bed stagnation in the bottom of the retort. Furthermore, the retort and injection nozzle designs minimize buildup of pyrocarbon and silicon carbide on the retort wall and manage nozzle orifice accretions. The equipment and operating methods have yielded very good reproducibility in the TRISO coated particles batches.« less

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, R.B.; Fero, A.H.; Sejvar, J.

1997-12-16

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

1997-01-01

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

Nuclear reactor vessel fuel thermal insulating barrier

DOEpatents

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

NEUTRONIC REACTORS

DOEpatents

Vernon, H.C.

1959-01-13

A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murav’ev, V. P., E-mail: murval1@mail.ru

The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

158. ARAIII Reactor building (ARA608) Secondary cooling loop and piping ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

158. ARA-III Reactor building (ARA-608) Secondary cooling loop and piping plan. This drawing was selected as a typical example of piping arrangements within reactor building. Aerojet/general 880-area/GCRE-608-P-16. Date: February 1958. INeel index code no. 063-0608-50-013-102641. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Method and apparatus for a catalytic firebox reactor

DOEpatents

Smith, Lance L.; Etemad, Shahrokh; Ulkarim, Hasan; Castaldi, Marco J.; Pfefferle, William C.

2001-01-01

A catalytic firebox reactor employing an exothermic catalytic reaction channel and multiple cooling conduits for creating a partially reacted fuel/oxidant mixture. An oxidation catalyst is deposited on the walls forming the boundary between the multiple cooling conduits and the exothermic catalytic reaction channel, on the side of the walls facing the exothermic catalytic reaction channel. This configuration allows the oxidation catalyst to be backside cooled by any fluid passing through the cooling conduits. The heat of reaction is added to both the fluid in the exothermic catalytic reaction channel and the fluid passing through the cooling conduits. After discharge of the fluids from the exothermic catalytic reaction channel, the fluids mix to create a single combined flow. A further innovation in the reactor incorporates geometric changes in the exothermic catalytic reaction channel to provide streamwise variation of the velocity of the fluids in the reactor.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

Method and apparatus for enhancing reactor air-cooling system performance

DOEpatents

Hunsbedt, A.

1996-03-12

An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahmed, K. K.; Scarlat, R. O.; Hu, R.

Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

COOLING TOWER PUMP HOUSE, TRA606. THREE OF SIX SECTIONS OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

COOLING TOWER PUMP HOUSE, TRA-606. THREE OF SIX SECTIONS OF COOLING TOWER ARE VISIBLE ABOVE RAILING. PUMP HOUSE IN FOREGROUND IS ON SOUTH SIDE OF COOLING TOWER. NOTE THREE PIPES TAKING WATER FROM PUMP HOUSE TO HOT DECK OF COOLING TOWER. EMERGENCY WATER SUPPLY TOWER IS ALSO IN VIEW. INL NEGATIVE NO. 6197. Unknown Photographer, 6/27/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Pre-test CFD Calculations for a Bypass Flow Standard Problem

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rich Johnson

The bypass flow in a prismatic high temperature gas-cooled reactor (HTGR) is the flow that occurs between adjacent graphite blocks. Gaps exist between blocks due to variances in their manufacture and installation and because of the expansion and shrinkage of the blocks from heating and irradiation. Although the temperature of fuel compacts and graphite is sensitive to the presence of bypass flow, there is great uncertainty in the level and effects of the bypass flow. The Next Generation Nuclear Plant (NGNP) program at the Idaho National Laboratory has undertaken to produce experimental data of isothermal bypass flow between three adjacentmore » graphite blocks. These data are intended to provide validation for computational fluid dynamic (CFD) analyses of the bypass flow. Such validation data sets are called Standard Problems in the nuclear safety analysis field. Details of the experimental apparatus as well as several pre-test calculations of the bypass flow are provided. Pre-test calculations are useful in examining the nature of the flow and to see if there are any problems associated with the flow and its measurement. The apparatus is designed to be able to provide three different gap widths in the vertical direction (the direction of the normal coolant flow) and two gap widths in the horizontal direction. It is expected that the vertical bypass flow will range from laminar to transitional to turbulent flow for the different gap widths that will be available.« less

Heat exchanger with auxiliary cooling system

DOEpatents

Coleman, John H.

1980-01-01

A heat exchanger with an auxiliary cooling system capable of cooling a nuclear reactor should the normal cooling mechanism become inoperable. A cooling coil is disposed around vertical heat transfer tubes that carry secondary coolant therethrough and is located in a downward flow of primary coolant that passes in heat transfer relationship with both the cooling coil and the vertical heat transfer tubes. A third coolant is pumped through the cooling coil which absorbs heat from the primary coolant which increases the downward flow of the primary coolant thereby increasing the natural circulation of the primary coolant through the nuclear reactor.

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beyer, Brian David; Beddingfield, David H; Durst, Philip

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguardsmore » criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.« less

Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

NASA Astrophysics Data System (ADS)

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0237] Cost-Benefit Analysis for Radwaste Systems for Light... (RG) 1.110, ``Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors... components for light water nuclear power reactors. ADDRESSES: Please refer to Docket ID NRC-2013-0237 when...

Application of a Self-Actuating Shutdown System (SASS) to a Gas-Cooled Fast Reactor (GCFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Germer, J.H.; Peterson, L.F.; Kluck, A.L.

1980-09-01

The application of a SASS (Self-Actuated Shutdown System) to a GCFR (Gas-Cooled Fast Reactor) is compared with similar systems designed for an LMFBR (Liquid Metal Fast Breeder Reactor). A comparison of three basic SASS concepts is given: hydrostatic holdup, fluidic control, and magnetic holdup.

PROCESS FOR COOLING A NUCLEAR REACTOR

DOEpatents

Borst, L.B.

1962-12-11

This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

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

NASA Astrophysics Data System (ADS)

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.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

LIQUID METAL REACTOR COOLING SYSTEMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aberdam, M.; Gros, G.

1965-02-01

This report is part of a series of bibliographies. The specific purpose of this report is to describe the various elements of the cooling systems in the principal liquid-metal-cooled reactors now operating, being contsructed, or in the design stage. The information given is drawn from reports or publicatios received during or before September 1964.

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

Passive heat transfer means for nuclear reactors

DOEpatents

Burelbach, James P.

1984-01-01

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

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.

Control rod system useable for fuel handling in a gas-cooled nuclear reactor

DOEpatents

Spurrier, Francis R.

1976-11-30

A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

NASA Astrophysics Data System (ADS)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Houts, Michael G.

2011-01-01

Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans David

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 ofmore » 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.« less

Passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1989-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

ADVANCED COURSE ON FUEL ELEMENTS FOR WATER COOLED POWER REACTORS, ORGANIZED BY THE NETHERLANDS'-NORWEGIAN REACTOR SCHOOL AT INSTITUTT FOR ATOMENERGI, KJELLER, NORWAY, 22nd AUGUST-3rd SEPTEMBER,1960. VOLUME III

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aas, S.; Barendregt, T.J.; Chesne, A.

1960-07-01

A series of lectures on fuel elements for water-cooled power reactors are presented. Topics covered include fabrication, properties, cladding, radiation damage, design, cycling, storage and transpont, and reprocessing. Separate records have been prepared for each section.

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

NASA Astrophysics Data System (ADS)

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

2012-05-01

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Natural circulating passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1990-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

Passive heat-transfer means for nuclear reactors. [LMFBR

DOEpatents

Burelbach, J.P.

1982-06-10

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

77 FR 64148 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-18

... Regulatory Guides (RG) RG 1.79, ````Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors,'' Revision 2 and RG 1.79.1, ``Initial Test Program of Emergency Core Cooling Systems for...

DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

DOEpatents

Snell, A.H.; Allison, S.K.

1958-02-11

This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

High-Temperature Gas-Cooled Test Reactor Point Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

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.

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2013 CFR

2013-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2012 CFR

2012-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2014 CFR

2014-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

Passive containment cooling system

DOEpatents

Conway, Lawrence E.; Stewart, William A.

1991-01-01

A containment cooling system utilizes a naturally induced air flow and a gravity flow of water over the containment shell which encloses a reactor core to cool reactor core decay heat in two stages. When core decay heat is greatest, the water and air flow combine to provide adequate evaporative cooling as heat from within the containment is transferred to the water flowing over the same. The water is heated by heat transfer and then evaporated and removed by the air flow. After an initial period of about three to four days when core decay heat is greatest, air flow alone is sufficient to cool the containment.

Dynamic modeling of temperature change in outdoor operated tubular photobioreactors.

PubMed

Androga, Dominic Deo; Uyar, Basar; Koku, Harun; Eroglu, Inci

2017-07-01

In this study, a one-dimensional transient model was developed to analyze the temperature variation of tubular photobioreactors operated outdoors and the validity of the model was tested by comparing the predictions of the model with the experimental data. The model included the effects of convection and radiative heat exchange on the reactor temperature throughout the day. The temperatures in the reactors increased with increasing solar radiation and air temperatures, and the predicted reactor temperatures corresponded well to the measured experimental values. The heat transferred to the reactor was mainly through radiation: the radiative heat absorbed by the reactor medium, ground radiation, air radiation, and solar (direct and diffuse) radiation, while heat loss was mainly through the heat transfer to the cooling water and forced convection. The amount of heat transferred by reflected radiation and metabolic activities of the bacteria and pump work was negligible. Counter-current cooling was more effective in controlling reactor temperature than co-current cooling. The model developed identifies major heat transfer mechanisms in outdoor operated tubular photobioreactors, and accurately predicts temperature changes in these systems. This is useful in determining cooling duty under transient conditions and scaling up photobioreactors. The photobioreactor design and the thermal modeling were carried out and experimental results obtained for the case study of photofermentative hydrogen production by Rhodobacter capsulatus, but the approach is applicable to photobiological systems that are to be operated under outdoor conditions with significant cooling demands.

Preliminary design of high temperature ultrasonic transducers for liquid sodium environments

NASA Astrophysics Data System (ADS)

Prowant, M. S.; Dib, G.; Qiao, H.; Good, M. S.; Larche, M. R.; Sexton, S. S.; Ramuhalli, P.

2018-04-01

Advanced reactor concepts include fast reactors (including sodium-cooled fast reactors), gas-cooled reactors, and molten-salt reactors. Common to these concepts is a higher operating temperature (when compared to light-water-cooled reactors), and the proposed use of new alloys with which there is limited operational experience. Concerns about new degradation mechanisms, such as high-temperature creep and creep fatigue, that are not encountered in the light-water fleet and longer operating cycles between refueling intervals indicate the need for condition monitoring technology. Specific needs in this context include periodic in-service inspection technology for the detection and sizing of cracking, as well as technologies for continuous monitoring of components using in situ probes. This paper will discuss research on the development and evaluation of high temperature (>550°C; >1022°F) ultrasonic probes that can be used for continuous monitoring of components. The focus of this work is on probes that are compatible with a liquid sodium-cooled reactor environment, where the core outlet temperatures can reach 550°C (1022°F). Modeling to assess sensitivity of various sensor configurations and experimental evaluation have pointed to a preferred design and concept of operations for these probes. This paper will describe these studies and ongoing work to fabricate and fully evaluate survivability and sensor performance over extended periods at operational temperatures.

A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

NASA Astrophysics Data System (ADS)

Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

2005-02-01

A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

DOEpatents

DeVolpi, A.

1984-07-20

The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

Advanced Diesel Oil Fuel Processor Development

DTIC Science & Technology

1986-06-01

water exit 29 sample quencher: gas sample line inlet 30 sample quencher: gas sample line exit 31 sample quencher: cooling water inlet 32 desulfuriser ...exit line 33, 34 desulfurimer 35 heat exchanger: process gas exit (to desulfuriser ) 38 shift reactor inlet (top) 37 shift reactor: cooling air exit

Development work for a borax internal core-catcher for a gas-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho; Lee, Sangjin; Jin, Yoon-Su; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scheele, Randall D.; Casella, Andrew M.

2010-09-28

This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

NEUTRONIC REACTOR

DOEpatents

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbajo, Juan J; Qualls, A L

2016-01-01

INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a smallmore » version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.« less

Emergency core cooling system

DOEpatents

Schenewerk, William E.; Glasgow, Lyle E.

1983-01-01

A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

Core cooling under accident conditions at the high-flux beam reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tichler, P.; Cheng, L.; Fauske, H.

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fullymore » developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.« less

A 100-kWt NaK-Cooled Space Reactor Concept for an Early-Flight Mission

NASA Astrophysics Data System (ADS)

Poston, David I.

2003-01-01

A stainless-steel (SS) sodium-potassium (NaK) cooled reactor could potentially be the first step in utilizing fission technology in space. The sum of all system-level experience for liquid-metal-cooled space reactors has been with NaK, including the SNAP-10a, the only reactor ever launched by the US. This paper describes a 100-kWt NaK reactor, the NaK-100, which is designed to be developed with minimal technical risk. In additional to NaK technology heritage, the NaK-100 uses a proven fuel-form (SS/UO2) and is designed for simplified system integration and testing. The pins are placed within a solid SS prism, and the NaK flows in an annulus between the pins and the prism. The nuclear and thermal-hydraulic performance of the NaK-100 is presented, as well as the major differences between the NaK-100 and SNAP-10a.

PBF Cooling Tower. View from highbay roof of Reactor Building ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Cooling Tower. View from high-bay roof of Reactor Building (PER-620). Camera faces northwest. East louvered face has been installed. Inlet pipes protrude from fan deck. Two redwood vents under construction at top. Note piping, control, and power lines at sub-grade level in trench leading to Reactor Building. Photographer: Kirsh. Date: June 6, 1969. INEEL negative no. 69-3466 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Multi-Megawatt Power System Trade Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Longhurst, Glen Reed; Schnitzler, Bruce Gordon; Parks, Benjamin Travis

2001-11-01

As part of a larger task, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform a trade study comparing liquid-metal cooled reactors having Rankine power conversion systems with gas-cooled reactors having Brayton power conversion systems. This report summarizes the approach, the methodology, and the results of that trade study. Findings suggest that either approach has the possibility to approach the target specific mass of 3-5 kg/kWe for the power system, though it appears either will require improvements to achieve that. Higher reactor temperatures have the most potential for reducing the specific mass of gas-cooled reactors but domore » not necessarily have a similar effect for liquid-cooled Rankine systems. Fuels development will be the key to higher reactor operating temperatures. Higher temperature turbines will be important for Brayton systems. Both replacing lithium coolant in the primary circuit with gallium and replacing potassium with sodium in the power loop for liquid systems increase system specific mass. Changing the feed pump turbine to an electric motor in Rankine systems has little effect. Key technologies in reducing specific mass are high reactor and radiator operating temperatures, low radiator areal density, and low turbine/generator system masses. Turbine/generator mass tends to dominate overall power system mass for Rankine systems. Radiator mass was dominant for Brayton systems.« less

AIR COOLED NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Szilard, L.

1958-05-27

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Fuel leak detection apparatus for gas cooled nuclear reactors

DOEpatents

Burnette, Richard D.

1977-01-01

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, Terry L.

1993-01-01

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, T.L.

1993-10-19

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

System Study: Reactor Core Isolation Cooling 1998-2014

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

Transient Load Following and Control Analysis of Advanced S-CO2 Power Conversion with Dry Air Cooling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, Anton; Sienicki, James J.

2016-01-01

Supercritical carbon dioxide (S-CO2) Brayton cycles are under development as advanced energy converters for advanced nuclear reactors, especially the Sodium-Cooled Fast Reactor (SFR). The use of dry air cooling for direct heat rejection to the atmosphere ultimate heat sink is increasingly becoming a requirement in many regions due to restrictions on water use. The transient load following and control behavior of an SFR with an S-CO2 cycle power converter utilizing dry air cooling have been investigated. With extension and adjustment of the previously existing control strategy for direct water cooling, S-CO2 cycle power converters can also be used for loadmore » following operation in regions where dry air cooling is a requirement« less

Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

DOEpatents

Peterson, Per F.

2013-05-14

A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM BOTTOM TO TOP: ETR COOLING TOWER, ELECTRICAL BUILDING AND LOW-BAY SECTION OF ETR BUILDING, HEAT EXCHANGER BUILDING (WITH U SHAPED YARD), COMPRESSOR BUILDING. MTR REACTOR SERVICES BUILDING IS ATTACHED TO SOUTH WALL OF MTR. WING A IS ATTACHED TO BALCONY FLOOR OF MTR. NEAR UPPER RIGHT CORNER OF VIEW IS MTR PROCESS WATER BUILDING. WING B IS AT FAR WEST END OF COMPLEX. NEAR MAIN GATE IS GAMMA FACILITY, WITH "COLD" BUILDINGS BEYOND: RAW WATER STORAGE TANKS, STEAM PLANT, MTR COOLING TOWER PUMP HOUSE AND COOLING TOWER. INL NEGATIVE NO. 56-4101. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Nuclear reactor insulation and preheat system

DOEpatents

Wampole, Nevin C.

1978-01-01

An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

A Comparison of Fission Power System Options for Lunar and Mars Surface Applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

This paper presents a comparison of reactor and power conversion design options for 50 kWe class lunar and Mars surface power applications with scaling from 25 to 200 kWe. Design concepts and integration approaches are provided for three reactor-converter combinations: gas-cooled Brayton, liquid-metal Stirling, and liquid-metal thermoelectric. The study examines the mass and performance of low temperature, stainless steel based reactors and higher temperature refractory reactors. The preferred system implementation approach uses crew-assisted assembly and in-situ radiation shielding via installation of the reactor in an excavated hole. As an alternative, self-deployable system concepts that use earth-delivered, on-board radiation shielding are evaluated. The analyses indicate that among the 50 kWe stainless steel reactor options, the liquid-metal Stirling system provides the lowest mass at about 5300 kg followed by the gas-cooled Brayton at 5700 kg and the liquid-metal thermoelectric at 8400 kg. The use of a higher temperature, refractory reactor favors the gas-cooled Brayton option with a system mass of about 4200 kg as compared to the Stirling and thermoelectric options at 4700 and 5600 kg, respectively. The self-deployed concepts with on-board shielding result in a factor of two system mass increase as compared to the in-situ shielded concepts.

DE-NE0008277_PROTEUS final technical report 2018

DOE Office of Scientific and Technical Information (OSTI.GOV)

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Reactor core isolation cooling system

DOEpatents

Cooke, Franklin E.

1992-01-01

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

COOLED NEUTRONIC REACTOR

DOEpatents

Binner, C.R.; Wilkie, C.B.

1958-03-18

This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Permana, Sidik; Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132; Sekimoto, Hiroshi

2010-12-23

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period hasmore » been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.« less

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradin, Michael; Anderson, M.; Muci, M.

This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintainmore » similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.« less

PBF Reactor Building (PER620). Plot plan shows layout, including auxiliary ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Plot plan shows layout, including auxiliary buildings: Emergency Generator (621), Hose House (622), Cooling Tower Auxiliary (624), Maintenance and Storage Warehouse (625), Gas Cylinder Storage (627), Hose House (628), Cooling Tower (720), Substation (719), and other features. Road connections between PBF Reactor, its control building, and SPERT-I site. Note cable trenches along road to control building. Date: July 1965. Ebasco Services, PER-U-101. INEEL index no. 761-0100-00-205-123005 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Low exchange element for nuclear reactor

DOEpatents

Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

1985-01-01

A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

Monitoring of tritium

DOEpatents

Corbett, James A.; Meacham, Sterling A.

1981-01-01

The fluid from a breeder nuclear reactor, which may be the sodium cooling fluid or the helium reactor-cover-gas, or the helium coolant of a gas-cooled reactor passes over the portion of the enclosure of a gaseous discharge device which is permeable to hydrogen and its isotopes. The tritium diffused into the discharge device is radioactive producing beta rays which ionize the gas (argon) in the discharge device. The tritium is monitored by measuring the ionization current produced when the sodium phase and the gas phase of the hydrogen isotopes within the enclosure are in equilibrium.

U.S./CIS eye joint nuclear rocket venture

NASA Technical Reports Server (NTRS)

Clark, John S.; Mcilwain, Melvin C.; Smetanikov, Vladimir; D'Yakov, Evgenij K.; Pavshuk, Vladimir A.

1993-01-01

An account is given of the significance for U.S. spacecraft development of a nuclear thermal rocket (NTR) reactor concept that has been developed in the (formerly Soviet) Commonwealth of Independent States (CIS). The CIS NTR reactor employs a hydrogen-cooled zirconium hydride moderator and ternary carbide fuels; the comparatively cool operating temperatures associated with this design promise overall robustness.

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2010 CFR

2010-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents SECTION I. Introduction. Section 50.34a provides that an application for a construction...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2011 CFR

2011-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents SECTION I. Introduction. Section 50.34a provides that an application for a construction...

The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hagen, T.H.J.J. van der; Stekelenburg, A.J.C.

1995-09-01

The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

Apparatus for controlling nuclear core debris

DOEpatents

Jones, Robert D.

1978-01-01

Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yacout, A. M.; Billone, M. C.

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of datamore » were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.« less

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

76 FR 79229 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on January 19-20, 2012, 11545 Rockville... Cooling Systems for Light- Water Nuclear Power Reactors'' (Open)--The Committee will hear presentations by...

Radial blanket assembly orificing arrangement

DOEpatents

Patterson, J.F.

1975-07-01

A nuclear reactor core for a liquid metal cooled fast breeder reactor is described in which means are provided for increasing the coolant flow through the reactor fuel assemblies as the reactor ages by varying the coolant flow rate with the changing coolant requirements during the core operating lifetime. (auth)

Progressing batch hydrolysis process

DOEpatents

Wright, J.D.

1985-01-10

A progressive batch hydrolysis process is disclosed for producing sugar from a lignocellulosic feedstock. It comprises passing a stream of dilute acid serially through a plurality of percolation hydrolysis reactors charged with feed stock, at a flow rate, temperature and pressure sufficient to substantially convert all the cellulose component of the feed stock to glucose. The cooled dilute acid stream containing glucose, after exiting the last percolation hydrolysis reactor, serially fed through a plurality of pre-hydrolysis percolation reactors, charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the hemicellulose component of said feedstock to glucose. The dilute acid stream containing glucose is cooled after it exits the last prehydrolysis reactor.

Progressing batch hydrolysis process

DOEpatents

Wright, John D.

1986-01-01

A progressive batch hydrolysis process for producing sugar from a lignocellulosic feedstock, comprising passing a stream of dilute acid serially through a plurality of percolation hydrolysis reactors charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the cellulose component of the feedstock to glucose; cooling said dilute acid stream containing glucose, after exiting the last percolation hydrolysis reactor, then feeding said dilute acid stream serially through a plurality of prehydrolysis percolation reactors, charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the hemicellulose component of said feedstock to glucose; and cooling the dilute acid stream containing glucose after it exits the last prehydrolysis reactor.

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamada, K.; Aksan, S. N.

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

Very high temperature behavior of HTGR core materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soo, P.; Uneberg, G.; Sabatini, R.

1978-01-01

A description is given of experiments to investigate the behavior of HTGR core materials during hypothetical heatup accidents in which the core temperature is assumed to reach values between 2400/sup 0/C and the graphite sublimation range (>3600/sup 0/C). The work includes BISO coated fuel particle failure, simulated fission product migration in core graphite, and graphite sublimation behavior.

77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on November 1-3, 2012, 11545 Rockville...-Term Core Cooling Approach for the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project...

Development of toroid-type HTS DC reactor series for HVDC system

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

NASA Astrophysics Data System (ADS)

Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

2004-02-01

One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

NASA Astrophysics Data System (ADS)

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 transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Reflux cooling experiments on the NCSU scaled PWR facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doster, J.M.; Giavedoni, E.

1993-01-01

Under loss of forced circulation, coupled with the loss or reduction in primary side coolant inventory, horizontal stratified flows can develop in the hot and cold legs of pressurized water reactors (PWRs). Vapor produced in the reactor vessel is transported through the hot leg to the steam generator tubes where it condenses and flows back to the reactor vessel. Within the steam generator tubes, the flow regimes may range from countercurrent annular flow to single-phase convection. As a result, a number of heat transfer mechanisms are possible, depending on the loop configuration, total heat transfer rate, and the steam flowmore » rate within the tubes. These include (but are not limited to) two-phase natural circulation, where the condensate flows concurrent to the vapor stream and is transported to the cold leg so that the entire reactor coolant loop is active, and reflux cooling, where the condensate flows back down the interior of the coolant tubes countercurrent to the vapor stream and is returned to the reactor vessel through the hot leg. While operating in the reflux cooling mode, the cold leg can effectively be inactive. Heat transfer can be further influenced by noncondensables in the vapor stream, which accumulate within the upper regions of the steam generator tube bundle. In addition to reducing the steam generator's effective heat transfer area, under these conditions operation under natural circulation may not be possible, and reflux cooling may be the only viable heat transfer mechanism. The scaled PWR (SPWR) facility in the nuclear engineering department at North Carolina State Univ. (NCSU) is being used to study the effectiveness of two-phase natural circulation and reflux cooling under conditions associated with loss of forced circulation, midloop coolant levels, and noncondensables in the primary coolant system.« less

Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Webster, K. L.

2007-01-01

Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.

2013-11-01

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Radiant vessel auxiliary cooling system

DOEpatents

Germer, John H.

1987-01-01

In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takeda, T.; Shimazu, Y.; Hibi, K.

2012-07-01

Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of thismore » project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)« less

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moriarty, M.P.

1993-01-15

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system

NASA Astrophysics Data System (ADS)

Moriarty, Michael P.

1993-01-01

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Characteristics of potential repository wastes. Volume 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-07-01

The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continuemore » and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities.« less

Nuclear Electric Magnetohydrodynamic Propulsion for Submarine

DTIC Science & Technology

1989-05-01

develop - ment strategies for the future. The base program includes the development of the LMFBR, and HTGR to...events. Oxide fuel is -134- being retained as a backup, pending the outcome of the metal fuel development program . The design allows for a quick fuel ... HTGR plants can be developed with much higher source temperature and core power density. High efficiency and low power den- sity characteristics

Catalog of experimental projects for a fissioning plasma reactor

NASA Technical Reports Server (NTRS)

Lanzo, C. D.

1973-01-01

Experimental and theoretical investigations were carried out to determine the feasibility of using a small scale fissioning uranium plasma as the power source in a driver reactor. The driver system is a light water cooled and moderated reactor of the MTR type. The eight experiments and proposed configurations for the reactor are outlined.

Application of Molten Salt Reactor Technology to Nuclear Electric Propulsion Mission

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Nuclear electric propulsion (NEP) and planetary surface power missions require reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional gas cooled, liquid metal, and heat pipe space reactors.

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. Amore » list of variable names and a listing for BRENDA are included as appendices.« less

Engine management during NTRE start up

NASA Technical Reports Server (NTRS)

Bulman, Mel; Saltzman, Dave

1993-01-01

The topics are presented in viewgraph form and include the following: total engine system management critical to successful nuclear thermal rocket engine (NTRE) start up; NERVA type engine start windows; reactor power control; heterogeneous reactor cooling; propellant feed system dynamics; integrated NTRE start sequence; moderator cooling loop and efficient NTRE starting; analytical simulation and low risk engine development; accurate simulation through dynamic coupling of physical processes; and integrated NTRE and mission performance.

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

DOEpatents

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to make scoping calculations for mission analysis.

ETR, TRA642. EASTWEST SECTION, LOOKING NORTH. PATH OF COOLING WATER ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. EAST-WEST SECTION, LOOKING NORTH. PATH OF COOLING WATER PIPE TUNNEL. WORKING AND STORAGE CANAL. SUB-PILE ROOM. CONTROL ROD ACCESS ROOM. FLOOR NAMES. (THIS WAS A CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-5, 11/1955. INL INDEX NO. 532-0642-00-486-100913. REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Application of the aqueous self-cooled blanket concept to fusion reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deutsch, L.; Steiner, D.; Embrechts, M.J.

1986-01-01

The development of a reliable, safe, and economically attractive tritium breeding blanket is an essential requirement in the path to commercial fusion power. The primary objective of the recently completed Blanket Comparison and Selection Study (BCSS) was to evaluate previously proposed concepts, and thereby identify a limited number of preferred options that would provide the focus for an R and D program. The water-cooled concepts in the BCSS scored relatively low. We consider it prudent that a promising water-cooled blanket concept be included in this program since nearly all power producing reactors currently rely on water technology. It is inmore » this context that we propose the novel water-cooled blanket concept described herein.« less

Continuous production of tritium in an isotope-production reactor with a separate circulation system

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

Space Nuclear Reactor Engineering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David Irvin

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE PAGES

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; ...

2017-02-26

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

SCW Pressure-Channel Nuclear Reactor Some Design Features

NASA Astrophysics Data System (ADS)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

139. ARAIII Index of drwaings of gascooled reactor experiment buildings. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

139. ARA-III Index of drwaings of gas-cooled reactor experiment buildings. Aerojet-general 880-area/GCRE-100. Date: February 1958. Ineel index code no. 063-9999-80-013-102505. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

ETR COOLING TOWER. PUMP HOUSE (TRA645) IN SHADOW OF TOWER ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR COOLING TOWER. PUMP HOUSE (TRA-645) IN SHADOW OF TOWER ON LEFT. AT LEFT OF VIEW, HIGH-BAY BUILDING IS ETR. ONE STORY ATTACHMENT IS ETR ELECTRICAL BUILDING. STACK AT RIGHT IS ETR STACK; MTR STACK IS TOWARD LEFT. CAMERA FACING NORTHEAST. INL NEGATIVE NO. 56-3799. Jack L. Anderson, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MEANS FOR SHIELDING AND COOLING REACTORS

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

NASA Astrophysics Data System (ADS)

Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

2010-12-01

Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

Numerical modelling of heat and mass transfer in adsorption solar reactor of ammonia on active carbon

NASA Astrophysics Data System (ADS)

Aroudam, El. H.

In this paper, we present a modelling of the performance of a reactor of a solar cooling machine based carbon-ammonia activated bed. Hence, for a solar radiation, measured in the Energetic Laboratory of the Faculty of Sciences in Tetouan (northern Morocco), the proposed model computes the temperature distribution, the pressure and the ammonia concentration within the activated carbon bed. The Dubinin-Radushkevich formula is used to compute the ammonia concentration distribution and the daily cycled mass necessary to produce a cooling effect for an ideal machine. The reactor is heated at a maximum temperature during the day and cool at the night. A numerical simulation is carried out employing the recorded solar radiation data measured locally and the daily ambient temperature for the typical clear days. Initially the reactor is at ambient temperature, evaporating pressure; Pev=Pst(Tev=0 ∘C) and maintained at uniform concentration. It is heated successively until the threshold temperature corresponding to the condensing pressure; Pcond=Pst(Tam) (saturation pressure at ambient temperature; in the condenser) and until a maximum temperature at a constant pressure; Pcond. The cooling of the reactor is characterised by a fall of temperature to the minimal values at night corresponding to the end of a daily cycle. We use the mass balance equations as well as energy equation to describe heat and mass transfer inside the medium of three phases. A numerical solution of the obtained non linear equations system based on the implicit finite difference method allows to know all parameters characteristic of the thermodynamic cycle and consider principally the daily evolution of temperature, ammonia concentration for divers positions inside the reactor. The tube diameter of the reactor shows the dependence of the optimum value on meteorological parameters for 1 m2 of collector surface.

Methanation assembly using multiple reactors

DOEpatents

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24

A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

ETR, TRA642. ETR COMPLEX NEARLY COMPLETE. CAMERA FACES NORTHWEST, PROBABLY ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. ETR COMPLEX NEARLY COMPLETE. CAMERA FACES NORTHWEST, PROBABLY FROM TOP DECK OF COOLING TOWER. SHADOW IS CAST BY COOLING TOWER UNITS OFF LEFT OF VIEW. HIGH-BAY REACTOR BUILDING IS SURROUNDED BY ITS ATTACHED SERVICES: ELECTRICAL (TRA-648), HEAT EXCHANGER (TRA-644 WITH U-SHAPED YARD), AND COMPRESSOR (TRA-643). THE CONTROL BUILDING (TRA-647) ON THE NORTH SIDE IS HIDDEN FROM VIEW. AT UPPER RIGHT IS MTR BUILDING, TRA-603. INL NEGATIVE NO. 56-3798. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Top shield temperatures, C and K Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agar, J.D.

1964-12-28

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locationsmore » in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.« less

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

NASA Astrophysics Data System (ADS)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

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

DOEpatents

Kelly, Sean M

2016-09-27

A system and method for air 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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, Sean M.

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.

Dynamic System Simulation of the KRUSTY Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klein, Steven Karl; Kimpland, Robert Herbert

2016-05-09

The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is expected to be approximately four (4) to five (5) kilowatts with a core temperature above 1,000 K. No data is available on any historical reactor employing HEU metal that operated over the temperature range required for the KRUSTY experiment. Further, no reactor has operated with heat pipes as the primary cooling mechanism. Historic power reactors have employed either naturalmore » or forced convection so data on their operation is not directly applicable to the KRUSTY experiment. The primary purpose of the system model once developed and refined by data from these component experiments, will be used to plan the KRUSTY experiment. This planning will include expected behavior of the reactor from start-up, through various transient conditions where cooling begins to become present and effective, and finally establishment of steady-state. In addition, the model can provide indicators of anticipated off-normal events and appropriate operator response to those conditions. This information can be used to develop specific experiment operating procedures and aids to guide the operators in conduct of the experiment.« less

Single Channel Testing for Characterization of the Direct Gas Cooled Reactor and the SAFE-100 Heat Exchanger

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bragg-Sitton, S.M.; Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812; Kapernick, R.

2004-02-04

Experiments have been designed to characterize the coolant gas flow in two space reactor concepts that are currently under investigation by NASA Marshall Space Flight Center and Los Alamos National Laboratory: the direct-drive gas-cooled reactor (DDG) and the SAFE-100 heatpipe-cooled reactor (HPR). For the DDG concept, initial tests have been completed to measure pressure drop versus flow rate for a prototypic core flow channel, with gas exiting to atmospheric pressure conditions. The experimental results of the completed DDG tests presented in this paper validate the predicted results to within a reasonable margin of error. These tests have resulted in amore » re-design of the flow annulus to reduce the pressure drop. Subsequent tests will be conducted with the re-designed flow channel and with the outlet pressure held at 150 psi (1 MPa). Design of a similar test for a nominal flow channel in the HPR heat exchanger (HPR-HX) has been completed and hardware is currently being assembled for testing this channel at 150 psi. When completed, these test programs will provide the data necessary to validate calculated flow performance for these reactor concepts (pressure drop and film temperature rise)« less

Modeling Hybrid Nuclear Systems With Chilled-Water Storage

DOE PAGES

Misenheimer, Corey T.; Terry, Stephen D.

2016-06-27

Air-conditioning loads during the warmer months of the year are large contributors to an increase in the daily peak electrical demand. Traditionally, utility companies boost output to meet daily cooling load spikes, often using expensive and polluting fossil fuel plants to match the demand. Likewise, heating, ventilation, and air conditioning (HVAC) system components must be sized to meet these peak cooling loads. However, the use of a properly sized stratified chilled-water storage system in conjunction with conventional HVAC system components can shift daily energy peaks from cooling loads to off-peak hours. This process is examined in light of the recentmore » development of small modular nuclear reactors (SMRs). In this paper, primary components of an air-conditioning system with a stratified chilled-water storage tank were modeled in FORTRAN 95. A basic chiller operation criterion was employed. Simulation results confirmed earlier work that the air-conditioning system with thermal energy storage (TES) capabilities not only reduced daily peaks in energy demand due to facility cooling loads but also shifted the energy demand from on-peak to off-peak hours, thereby creating a more flattened total electricity demand profile. Thus, coupling chilled-water storage-supplemented HVAC systems to SMRs is appealing because of the decrease in necessary reactor power cycling, and subsequently reduced associated thermal stresses in reactor system materials, to meet daily fluctuations in cooling demand. Finally and also, such a system can be used as a thermal sink during reactor transients or a buffer due to renewable intermittency in a nuclear hybrid energy system (NHES).« less

Modeling Hybrid Nuclear Systems With Chilled-Water Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Misenheimer, Corey T.; Terry, Stephen D.

Air-conditioning loads during the warmer months of the year are large contributors to an increase in the daily peak electrical demand. Traditionally, utility companies boost output to meet daily cooling load spikes, often using expensive and polluting fossil fuel plants to match the demand. Likewise, heating, ventilation, and air conditioning (HVAC) system components must be sized to meet these peak cooling loads. However, the use of a properly sized stratified chilled-water storage system in conjunction with conventional HVAC system components can shift daily energy peaks from cooling loads to off-peak hours. This process is examined in light of the recentmore » development of small modular nuclear reactors (SMRs). In this paper, primary components of an air-conditioning system with a stratified chilled-water storage tank were modeled in FORTRAN 95. A basic chiller operation criterion was employed. Simulation results confirmed earlier work that the air-conditioning system with thermal energy storage (TES) capabilities not only reduced daily peaks in energy demand due to facility cooling loads but also shifted the energy demand from on-peak to off-peak hours, thereby creating a more flattened total electricity demand profile. Thus, coupling chilled-water storage-supplemented HVAC systems to SMRs is appealing because of the decrease in necessary reactor power cycling, and subsequently reduced associated thermal stresses in reactor system materials, to meet daily fluctuations in cooling demand. Finally and also, such a system can be used as a thermal sink during reactor transients or a buffer due to renewable intermittency in a nuclear hybrid energy system (NHES).« less

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

NASA Astrophysics Data System (ADS)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

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

NASA Astrophysics Data System (ADS)

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Per; Greenspan, Ehud

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 designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. 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-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

10 CFR 50.55a - Codes and standards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... specified in § 50.55, except that each combined license for a boiling or pressurized water-cooled nuclear... boiling or pressurized water-cooled nuclear power facility is subject to the conditions in paragraphs (f... performed. (2) Systems and components of boiling and pressurized water-cooled nuclear power reactors must...

77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-15

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core... acceptable to implement with regard to initial testing features of emergency core cooling systems (ECCSs) for...

BOILING REACTORS

DOEpatents

Untermyer, S.

1962-04-10

A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

NASA Astrophysics Data System (ADS)

Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

2012-06-01

Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

Reactor vessel lower head integrity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rubin, A.M.

1997-02-01

On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) inmore » this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.« less

Process for making silicon from halosilanes and halosilicons

NASA Technical Reports Server (NTRS)

Levin, Harry (Inventor)

1988-01-01

A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

Process for making silicon

NASA Technical Reports Server (NTRS)

Levin, Harry (Inventor)

1987-01-01

A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCulloch, R.W.; Post, D.W.; Lovell, R.T.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relatemore » this profile to that generated by the coils in completed fuel pin simulators.« less

Cooling water distribution system

DOEpatents

Orr, Richard

1994-01-01

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, Paul R.

1994-01-01

A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

Decay Heat Removal from a GFR Core by Natural Convection

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, Wesley C.; Hejzlar, Pavel; Driscoll, Michael J.

2004-07-01

One of the primary challenges for Gas-cooled Fast Reactors (GFR) is decay heat removal after a loss of coolant accident (LOCA). Due to the fact that thermal gas cooled reactors currently under design rely on passive mechanisms to dissipate decay heat, there is a strong motivation to accomplish GFR core cooling through natural phenomena. This work investigates the potential of post-LOCA decay heat removal from a GFR core to a heat sink using an external convection loop. A model was developed in the form of the LOCA-COLA (Loss of Coolant Accident - Convection Loop Analysis) computer code as a meansmore » for 1D steady state convective heat transfer loop analysis. The results show that decay heat removal by means of gas cooled natural circulation is feasible under elevated post-LOCA containment pressure conditions. (authors)« less

ETR COMPLEX. CAMERA FACING SOUTH. FROM BOTTOM OF VIEW TO ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR COMPLEX. CAMERA FACING SOUTH. FROM BOTTOM OF VIEW TO TOP: MTR, MTR SERVICE BUILDING, ETR CRITICAL FACILITY, ETR CONTROL BUILDING (ATTACHED TO ETR), ETR BUILDING (HIGH-BAY), COMPRESSOR BUILDING (ATTACHED AT LEFT OF ETR), HEAT EXCHANGER BUILDING (JUST BEYOND COMPRESSOR BUILDING), COOLING TOWER PUMP HOUSE, COOLING TOWER. OTHER BUILDINGS ARE CONTRACTORS' CONSTRUCTION BUILDINGS. INL NEGATIVE NO. 56-4105. Unknown Photographer, ca. 1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Optimization of a heat-pipe-cooled space radiator for use with a reactor-powered Stirling engine

NASA Technical Reports Server (NTRS)

Moriarty, Michael P.; French, Edward P.

1987-01-01

The design optimization of a reactor-Stirling heat-pipe-cooled radiator is presented. The radiator is a self-deploying concept that uses individual finned heat pipe 'petals' to reject waste heat from a Stirling engine. Radiator optimization methodology is presented, and the results of a parametric analysis of the radiator design variables for a 100-kW(e) system are given. The additional steps of optiminzing the radiator resulted in a net system mass savings of 3 percent.

Fast quench reactor and method

DOEpatents

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

1998-05-12

A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

Key Assets for a Sustainable Low Carbon Energy Future

NASA Astrophysics Data System (ADS)

Carre, Frank

2011-10-01

Since the beginning of the 21st century, concerns of energy security and climate change gave rise to energy policies focused on energy conservation and diversified low-carbon energy sources. Provided lessons of Fukushima accident are evidently accounted for, nuclear energy will probably be confirmed in most of today's nuclear countries as a low carbon energy source needed to limit imports of oil and gas and to meet fast growing energy needs. Future challenges of nuclear energy are then in three directions: i) enhancing safety performance so as to preclude any long term impact of severe accident outside the site of the plant, even in case of hypothetical external events, ii) full use of Uranium and minimization long lived radioactive waste burden for sustainability, and iii) extension to non-electricity energy products for maximizing the share of low carbon energy source in transportation fuels, industrial process heat and district heating. Advanced LWRs (Gen-III) are today's best available technologies and can somewhat advance nuclear energy in these three directions. However, breakthroughs in sustainability call for fast neutron reactors and closed fuel cycles, and non-electric applications prompt a revival of interest in high temperature reactors for exceeding cogeneration performances achievable with LWRs. Both types of Gen-IV nuclear systems by nature call for technology breakthroughs to surpass LWRs capabilities. Current resumption in France of research on sodium cooled fast neutron reactors (SFRs) definitely aims at significant progress in safety and economic competitiveness compared to earlier reactors of this type in order to progress towards a new generation of commercially viable sodium cooled fast reactor. Along with advancing a new generation of sodium cooled fast reactor, research and development on alternative fast reactor types such as gas or lead-alloy cooled systems (GFR & LFR) is strategic to overcome technical difficulties and/or political opposition specific to sodium. In conclusion, research and technology breakthroughs in nuclear power are needed for shaping a sustainable low carbon future. International cooperation is key for sharing costs of research and development of the required novel technologies and cost of first experimental reactors needed to demonstrate enabling technologies. At the same time technology breakthroughs are developed, pre-normative research is required to support codification work and harmonized regulations that will ultimately apply to safety and security features of resulting innovative reactor types and fuel cycles.

Heat Pipe Technology: A bibliography with abstracts

NASA Technical Reports Server (NTRS)

1974-01-01

This bibliography lists 149 references with abstracts and 47 patents dealing with applications of heat pipe technology. Topics covered include: heat exchangers for heat recovery; electrical and electronic equipment cooling; temperature control of spacecraft; cryosurgery; cryogenic, cooling; nuclear reactor heat transfer; solar collectors; laser mirror cooling; laser vapor cavitites; cooling of permafrost; snow melting; thermal diodes variable conductance; artery gas venting; and venting; and gravity assisted pipes.

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

77 FR 42771 - License Renewal for the Dow Chemical TRIGA Research Reactor

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-20

... Chemical Company in Midland, MI and is a part of the Analytical Sciences Laboratory. The reactor is housed...-Radiological Impacts The Dow TRIGA Research Reactor core is cooled by a light water primary system consisting... provided by the volume of primary coolant allows several hours of full-power operation without any...

Experiments on rehabilitation of radioactive metallic waste (RMW) of reactor stainless steels of Siberian chemical plant

NASA Astrophysics Data System (ADS)

Kolpakov, G. N.; Zakusilov, V. V.; Demyanenko, N. V.; Mishin, A. S.

2016-06-01

Stainless steel pipes, used to cool a reactor plant, have a high cost, and after taking a reactor out of service they must be buried together with other radioactive waste. Therefore, the relevant problem is the rinse of pipes from contamination, followed by returning to operation.

Coupled field-structural analysis of HGTR fuel brick using ABAQUS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, S.; Jain, R.; Majumdar, S.

2012-07-01

High-temperature, gas-cooled reactors (HTGRs) are usually helium-gas cooled, with a graphite core that can operate at reactor outlet temperatures much higher than can conventional light water reactors. In HTGRs, graphite components moderate and reflect neutrons. During reactor operation, high temperature and high irradiation cause damage to the graphite crystal and grains and create other defects. This cumulative structural damage during the reactor lifetime leads to changes in graphite properties, which can alter the ability to support the designed loads. The aim of the present research is to develop a finite-element code using commercially available ABAQUS software for the structural integritymore » analysis of graphite core components under extreme temperature and irradiation conditions. In addition, the Reactor Geometry Generator tool-kit, developed at Argonne National Laboratory, is used to generate finite-element mesh for complex geometries such as fuel bricks with multiple pin holes and coolant flow channels. This paper presents the proposed concept and discusses results of stress analysis simulations of a fuel block with H-451 grade material properties. (authors)« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sridharan, Kumar; Allen, Todd; Anderson, Mark

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,more » 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.« less

Research Program of a Super Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Passive decay heat removal system for water-cooled nuclear reactors

DOEpatents

Forsberg, Charles W.

1991-01-01

A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

WATER BOILER REACTOR

DOEpatents

King, L.D.P.

1960-11-22

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

PBF Cooling Tower contextual view. Camera facing southwest. West wing ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Cooling Tower contextual view. Camera facing southwest. West wing and north facade (rear) of Reactor Building (PER-620) is at left; Cooling Tower to right. Photographer: Kirsh. Date: November 2, 1970. INEEL negative no. 70-4913 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Bragg-Sitton, Shannon; VanDyke, Melissa

2003-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

NASA Technical Reports Server (NTRS)

Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

2002-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns

2014-01-01

The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during themore » development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.« less

New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

DOE PAGES

Takamatsu, Kuniyoshi; Hu, Rui

2014-11-27

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient airmore » as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, such as the maximum operating temperature of the RPV, 713.15 (K) = 440 (°C), and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). Finally, when the RCCS can remove 600 (kW) of the rated nominal state even during LOCA, the safety review for building the HTTR could confirm that the temperature distribution of the HTTR is within the temperature limits of the structures to secure structures and fuels after the shutdown because the large heat capacity of the graphite core can absorb heat from the fuel in a short period. Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal.« less

Design of megawatt power level heat pipe reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Online Oxide Contamination Measurement and Purification Demonstration

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Godfroy, T. J.; Webster, K. L.; Garber, A. E.; Polzin, K. A.; Childers, D. J.

2011-01-01

Liquid metal sodium-potassium (NaK) has advantageous thermodynamic properties indicating its use as a fission reactor coolant for a surface (lunar, martian) power system. A major area of concern for fission reactor cooling systems is system corrosion due to oxygen contaminants at the high operating temperatures experienced. A small-scale, approximately 4-L capacity, simulated fission reactor cooling system employing NaK as a coolant was fabricated and tested with the goal of demonstrating a noninvasive oxygen detection and purification system. In order to generate prototypical conditions in the simulated cooling system, several system components were designed, fabricated, and tested. These major components were a fully-sealed, magnetically-coupled mechanical NaK pump, a graphite element heated reservoir, a plugging indicator system, and a cold trap. All system components were successfully demonstrated at a maximum system flow rate of approximately 150 cc/s at temperatures up to 550 C. Coolant purification was accomplished using a cold trap before and after plugging operations which showed a relative reduction in oxygen content.

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

DOE PAGES

Cheng, Lap-Yan; Wei, Thomas Y. C.

2009-01-01

The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

Effect of Silver or Copper Nanoparticles-Dispersed Silane Coatings on Biofilm Formation in Cooling Water Systems

PubMed Central

Ogawa, Akiko; Kanematsu, Hideyuki; Sano, Katsuhiko; Sakai, Yoshiyuki; Ishida, Kunimitsu; Beech, Iwona B.; Suzuki, Osamu; Tanaka, Toshihiro

2016-01-01

Biofouling often occurs in cooling water systems, resulting in the reduction of heat exchange efficiency and corrosion of the cooling pipes, which raises the running costs. Therefore, controlling biofouling is very important. To regulate biofouling, we focus on the formation of biofilm, which is the early step of biofouling. In this study, we investigated whether silver or copper nanoparticles-dispersed silane coatings inhibited biofilm formation in cooling systems. We developed a closed laboratory biofilm reactor as a model of a cooling pipe and used seawater as a model for cooling water. Silver or copper nanoparticles-dispersed silane coating (Ag coating and Cu coating) coupons were soaked in seawater, and the seawater was circulated in the laboratory biofilm reactor for several days to create biofilms. Three-dimensional images of the surface showed that sea-island-like structures were formed on silane coatings and low concentration Cu coating, whereas nothing was formed on high concentration Cu coatings and low concentration Ag coating. The sea-island-like structures were analyzed by Raman spectroscopy to estimate the components of the biofilm. We found that both the Cu coating and Ag coating were effective methods to inhibit biofilm formation in cooling pipes. PMID:28773758

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

NASA Astrophysics Data System (ADS)

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 and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bohachek, Randolph Charles

2015-09-01

The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactorsmore » is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.« less

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

SAM Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

NASA Technical Reports Server (NTRS)

Clark, John S.; Walton, James T.; Mcguire, Melissa L.

1992-01-01

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

DOEpatents

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki, E-mail: szaki@fi.itba.c.id; Sekimoto, H., E-mail: hsekimot@gmail.com

2014-09-30

Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature canmore » be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.« less

Nuclear engine flow reactivity shim control

DOEpatents

Walsh, J.M.

1973-12-11

A nuclear engine control system is provided which automatically compensates for reactor reactivity uncertainties at the start of life and reactivity losses due to core corrosion during the reactor life in gas-cooled reactors. The coolant gas flow is varied automatically by means of specially provided control apparatus so that the reactor control drums maintain a predetermined steady state position throughout the reactor life. This permits the reactor to be designed for a constant drum position and results in a desirable, relatively flat temperature profile across the core. (Official Gazette)

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

NEUTRONIC REACTOR CHARGING AND DISCHARGING

DOEpatents

Zinn, W.H.

1959-07-14

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Nuclear energy center site survey reactor plant considerations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harty, H.

The Energy Reorganization Act of 1974 required the Nuclear Regulatory Commission (NRC) to make a nuclear energy center site survey (NECSS). Background information for the NECSS report was developed in a series of tasks which include: socioeconomic inpacts; environmental impact (reactor facilities); emergency response capability (reactor facilities); aging of nuclear energy centers; and dry cooled nuclear energy centers.

Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Tyler A.; Edwards, Michael J.

A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

HEAVY WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Szilard, L.

1958-04-29

A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, P.R.

1994-12-27

A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

Passive containment cooling water distribution device

DOEpatents

Conway, Lawrence E.; Fanto, Susan V.

1994-01-01

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

NASA Astrophysics Data System (ADS)

Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

2017-01-01

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

Neutronic reactor thermal shield

DOEpatents

Wende, Charles W. J.

1976-06-15

1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

Vented target elements for use in an isotope-production reactor. [LMFBR

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium gas in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins equipped with vents, and tritium gas is recovered from the coolant.

Assemblies with both target and fuel pins in an isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

Removal of hydrogen sulfide as ammonium sulfate from hydropyrolysis product vapors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marker, Terry L.; Felix, Larry G.; Linck, Martin B.

A system and method for processing biomass into hydrocarbon fuels that includes processing a biomass in a hydropyrolysis reactor resulting in hydrocarbon fuels and a process vapor stream and cooling the process vapor stream to a condensation temperature resulting in an aqueous stream. The aqueous stream is sent to a catalytic reactor where it is oxidized to obtain a product stream containing ammonia and ammonium sulfate. A resulting cooled product vapor stream includes non-condensable process vapors comprising H.sub.2, CH.sub.4, CO, CO.sub.2, ammonia and hydrogen sulfide.

METHOD AND APPARATUS FOR PRODUCING POWER

DOEpatents

Wollan, E.O.

1961-06-27

A neutronic reactor comprising two discrete zones; namely, an inner zone containing fissionable material and an outer zone containing fertile material is described. The inner zone is operated at a low temperature and is cooled by pressurized water. The outer zone is operated at a substantially higher temperature and is cooled by steam flashed from the inner zone. The reactor is particularly advantageous in that it produces high temperature steam; yet the materials of construction in the core (inner zone) are not restricted to materials capable of withstanding high temperature operation.

Piloted rich-catalytic lean-burn hybrid combustor

DOEpatents

Newburry, Donald Maurice

2002-01-01

A catalytic combustor assembly which includes, an air source, a fuel delivery means, a catalytic reactor assembly, a mixing chamber, and a means for igniting a fuel/air mixture. The catalytic reactor assembly is in fluid communication with the air source and fuel delivery means and has a fuel/air plenum which is coated with a catalytic material. The fuel/air plenum has cooling air conduits passing therethrough which have an upstream end. The upstream end of the cooling conduits is in fluid communication with the air source but not the fuel delivery means.

Removal of hydrogen sulfide as ammonium sulfate from hydropyrolysis product vapors

DOEpatents

Marker, Terry L; Felix, Larry G; Linck, Martin B; Roberts, Michael J

2014-10-14

A system and method for processing biomass into hydrocarbon fuels that includes processing a biomass in a hydropyrolysis reactor resulting in hydrocarbon fuels and a process vapor stream and cooling the process vapor stream to a condensation temperature resulting in an aqueous stream. The aqueous stream is sent to a catalytic reactor where it is oxidized to obtain a product stream containing ammonia and ammonium sulfate. A resulting cooled product vapor stream includes non-condensable process vapors comprising H.sub.2, CH.sub.4, CO, CO.sub.2, ammonia and hydrogen sulfide.

Development of Inspection and Repair Technology for Heat Exchanger Tubes in Fast Breeder Reactors

DTIC Science & Technology

2009-06-01

Technology for Heat Exchanger Tubes in Fast Breeder Reactors Akihiko NISHIMURA *1 , Takahisa SHOBU, Kiyoshi OKA, Toshihiko YAMAGUCHI, Yukihiro SHIMADA...fast breeder reactors (FBRs). It comprises a laser processing head combined with an eddy current testing unit. Ultrashort laser pulse ablation is used...be applied in the main- tenance of large structures such as nuclear reactors and chemical factories [1]. Internal access to a blanket cooling pipe

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

An intrinsically safe facility for forefront research and training on nuclear technologies — General description of the system

NASA Astrophysics Data System (ADS)

Mansani, L.; Bruzzone, M.; Frambati, S.; Reale, M.

2014-04-01

In the framework of research on generation-IV reactors, it is very important to have infrastructures specifically dedicated to the study of fundamental parameters in dynamics and kinetics of future fast-neutron reactors. Among various options pursued by international groups, Italy focused on lead-cooled reactors, which guarantee minimal neutron slowdown and capture and efficient cooling. In this paper it is described the design of a the low-power prototype generator, LEADS, that could be used within research facilities such as the National Laboratory of Legnaro of the INFN. The LEADS has a high safety standard in order to be used as a training facility, but it has also a good flexibility so as to allow a wide range of measurements and experiments. A high safety standard is achieved by limiting the reactor power to less than few hundred kW and the neutron multiplication factor k eff to less than 0.95 (a limiting value for spent fuel pool), by using a pure-uranium fuel (no plutonium) and by using solid lead as a diffuser. The proposed core is therefore intrinsically subcritical and has to be driven by an external neutron source generated by a proton beam impinging in a target. Preliminary simulations, performed with the MCNPX code indicated, for a 0.75mA continuous proton beam current at 70MeV proton energy, a reactor power of about 190kW when using a beryllium converter. The enriched-uranium fuel elements are immersed in a solid-lead matrix and contained within a steel vessel. The system is cooled by helium gas, which is transparent to neutrons and does not undergo activation. The gas is pumped by a compressor through specific holes at the entrance of the active volume with a temperature which varies according to the operating conditions and a pressure of about 1.1MPa. The hot gas coming out of the vessel is cooled by an external helium-water heat exchanger. The beryllium converter is cooled by its dedicated helium gas cooling system. After shutdown, the decay is completely dissipated by conduction through the lead reflector and steel vessel, and then evacuated by irradiation from the vessel surface to the external ambient air.

A roadmap for nuclear energy technology

NASA Astrophysics Data System (ADS)

Sofu, Tanju

2018-01-01

The prospects for the future use of nuclear energy worldwide can best be understood within the context of global population growth, urbanization, rising energy need and associated pollution concerns. As the world continues to urbanize, sustainable development challenges are expected to be concentrated in cities of the lower-middle-income countries where the pace of urbanization is fastest. As these countries continue their trajectory of economic development, their energy need will also outpace their population growth adding to the increased demand for electricity. OECD IEA's energy system deployment pathway foresees doubling of the current global nuclear capacity by 2050 to reduce the impact of rapid urbanization. The pending "retirement cliff" of the existing U.S. nuclear fleet, representing over 60 percent of the nation's emission-free electricity, also poses a large economic and environmental challenge. To meet the challenge, the U.S. DOE has developed the vision and strategy for development and deployment of advanced reactors. As part of that vision, the U.S. government pursues programs that aim to expand the use of nuclear power by supporting sustainability of the existing nuclear fleet, deploying new water-cooled large and small modular reactors to enable nuclear energy to help meet the energy security and climate change goals, conducting R&D for advanced reactor technologies with alternative coolants, and developing sustainable nuclear fuel cycle strategies. Since the current path relying heavily on water-cooled reactors and "once-through" fuel cycle is not sustainable, next generation nuclear energy systems under consideration aim for significant advances over existing and evolutionary water-cooled reactors. Among the spectrum of advanced reactor options, closed-fuel-cycle systems using reactors with fast-neutron spectrum to meet the sustainability goals offer the most attractive alternatives. However, unless the new public-private partnership models emerge to tackle the licensing and demonstration challenges for these advanced reactor concepts, realization of their enormous potential is not likely, at least in the U.S.

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

NASA Astrophysics Data System (ADS)

Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

2017-01-01

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Fuel pins with both target and fuel pellets in an isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

NEUTRONIC REACTORS

DOEpatents

Anderson, H.L.

1958-10-01

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Design, fabrication, and testing of an external fuel (UO2), full-length thermionic converter

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

The development of a full-length external-fuel thermionic converter for in-pile testing is described. The development program includes out-of-pile performance testing of the fully fueled-converter, using RF-induction heating, before its installation in the in-pile test capsule. The external-fuel converter is cylindrical in shape, and consists of an inner, centrally cooled collector, and an outer emitter surrounded by nuclear fuel. The term full-length denotes that the converter is long enough to extend over the full height of the reactor core. Thus, the converter is not a scaled-down test device, but a full-scale fuel element of the thermionic reactor. The external-fuel converter concept permits a number of different design options, particularly with respect to the fuel composition and shape, and the collector cooling arrangement. The converter described was developed for the Jet Propulsion Laboratory, and is based on their concept for a thermionic reactor with uninsulated collector cooling as previously described. The converter is double-ended, with through-flow cooling, and with ceramic seals and emitter and collector power take-offs at both ends. The design uses a revolver-shaped tungsten emitter body, with the central emitter hole surrounded by six peripheral fuel holes loaded with cylindrical UO2 pellets.

Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen (Technical Monitor)

2002-01-01

Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multimegawatt nuclear reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multimegawatt gas-cooled and liquid metal concepts.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

1998-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-09-24

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Summary of space nuclear reactor power systems, 1983--1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Z.; Southwest Science and Technology Univ., No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031; Chen, Y.

2012-07-01

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjustedmore » to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)« less

Ichthyoplankton entrainment study at the SRS Savannah River water intakes for Westinghouse Savannah River Company. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paller, M.

1992-03-26

Cooling water for L and K Reactors and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pump houses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water and passed through the reactor`s heat exchangers where temperatures may reach 70{degrees}C during full power operation. Ichthyoplankton mortality under such conditions is assumed to be 100 percent. The number of ichthyoplankton entrained into the cooling system depends on a variety of variables, including time of year, density and distribution of ichthyoplankton in themore » river, discharge levels in the river, and the volume of water withdrawn by the pumps. Entrainment at the 1 G pump house, which is immediately downstream from the confluence of Upper Three Runs Creek and the Savannah River, is also influenced by discharge rates and ichthyoplankton densities in Upper Three Runs Creek. Because of the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River, the Department of Energy requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory sample ichthyoplankton at the SRS Savannah River intakes. Dams & Moore, Inc., under a contract with Westinghouse Savannah River Company performed the sampling and data analysis for the ESS.« less

HTGR technology economic/ business analysis and trade studies impacts. Impacts of HTGR commericialization on the U.S. economy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Silady, Fred

The approach to this task was to initially review the 2012 Business Plan and supporting analyses for the above impacts. With that understanding as a base, the Business Plan impacts are updated in terms of the GDP and job creation as a result of additional studies and inputs such as the revised market assessment from Task 1.1. For the impacts on U.S. competitiveness, the NGNP Industry Alliance team members have been utilized to provide inputs on supplier infrastructure development and on vendor capability.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« less

Dynamics of heat-pipe reactors

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1971-01-01

A split-core heat pipe reactor, fueled with either U(233)C or U(235)C in a tungsten cermet and cooled by 7-Li-W heat pipes, was examined for the effects of the heat pipes on reactor while trying to safely absorb large reactivity inputs through inherent shutdown mechanisms. Limits on ramp reactivity inputs due to fuel melting temperature and heat pipe wall heat flux were mapped for the reactor in both startup and at-power operating modes.

HEDL FACILITIES CATALOG 400 AREA

DOE Office of Scientific and Technical Information (OSTI.GOV)

MAYANCSIK BA

1987-03-01

The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

Thermal baffle for fast-breeder reacton

DOEpatents

Rylatt, John A.

1977-01-01

A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel.

Municipal Waste Incinerator Public Works Center, Yokosuka Japan Evaluation and Recommendations

DTIC Science & Technology

1993-04-01

Incinerator and Pollution Control Equipment 24 XIV. Gas Cooling Chamber Water Injection Sites and Control Valve 25 XV. Quencher Reactor 27 XVI...discussed below.I 11I.B.1. Exhaust Gas Cooling Chamber Within the exhaust gas cooling chamber, water is atomized into the gas stream cools the gases...as it evaporates. The feed rate of water is controlled to provide gases entering the quencher at 3000C (Figure XIV). The gases exit the exhaust gas

Technical note on results of production test IP-628-AI: Crosstie flow test and flushing calibration, and the earlier production tests IP-573-AI and IP-499-AI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jones, S.S.

1964-06-15

A series of three experimental tests was conducted on the operation and adequacy of the K-Reactors` third, or last-ditch, cooling system. The first test showed considerable line corrosion the second test was performed directly after line cleaning, and the third test showed a significant amount of additional line corrosion after only nine months of service. The present cooling adequacy of this last-ditch system at the KE and KW reactors is summarized these show the power levels for which we have adequate last-ditch cooling as a function of the crosstie coolant temperature. These figures include the effects of increasing the numbermore » of pumps that remain in operation at the other K-Reactor, and various other operating or emergency conditions. These curves are for diesel pump speeds of 750 rpm which are planned for June of this year. In these figures the crosstie temperature is assumed to be a conservative 5{degrees}C above the process water inlet temperature.« less

THERMAL PROPERTIES AND HEATING AND COOLING DURABILITY OF REACTOR SHIELDING CONCRETE (in Japanese)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hosoi, J.; Chujo, K.; Saji, K.

1959-01-01

A study was made of the thermal properties of various concretes made of domestic raw materials for radiation shields of a power reactor and of a high- flux research reactor. The results of measurements of thermal expansion coefficient, specific heat, thermal diffusivity, thermal conductivity, cyclical heating, and cooling durability are described. Relationships between thermal properties and durability are discussed and several photographs of the concretes are given. It is shown that the heating and cooling durability of such a concrete which has a large thermal expansion coefficient or a considerable difference between the thermal expansion of coarse aggregate and themore » one of cement mortar part or aggregates of lower strength is very poor. The decreasing rates of bending strength and dynamical modulus of elasticity and the residual elongation of the concrete tested show interesting relations with the modified thermal stress resistance factor containing a ratio of bending strength and thermal expansion coefficient. The thermal stress resistance factor seems to depend on the conditions of heat transfer on the surface and on heat release in the concrete. (auth)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

PubMed

Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

2017-12-01

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Geomechanical Analysis of Underground Coal Gasification Reactor Cool Down for Subsequent CO2 Storage

NASA Astrophysics Data System (ADS)

Sarhosis, Vasilis; Yang, Dongmin; Kempka, Thomas; Sheng, Yong

2013-04-01

Underground coal gasification (UCG) is an efficient method for the conversion of conventionally unmineable coal resources into energy and feedstock. If the UCG process is combined with the subsequent storage of process CO2 in the former UCG reactors, a near-zero carbon emission energy source can be realised. This study aims to present the development of a computational model to simulate the cooling process of UCG reactors in abandonment to decrease the initial high temperature of more than 400 °C to a level where extensive CO2 volume expansion due to temperature changes can be significantly reduced during the time of CO2 injection. Furthermore, we predict the cool down temperature conditions with and without water flushing. A state of the art coupled thermal-mechanical model was developed using the finite element software ABAQUS to predict the cavity growth and the resulting surface subsidence. In addition, the multi-physics computational software COMSOL was employed to simulate the cavity cool down process which is of uttermost relevance for CO2 storage in the former UCG reactors. For that purpose, we simulated fluid flow, thermal conduction as well as thermal convection processes between fluid (water and CO2) and solid represented by coal and surrounding rocks. Material properties for rocks and coal were obtained from extant literature sources and geomechanical testings which were carried out on samples derived from a prospective demonstration site in Bulgaria. The analysis of results showed that the numerical models developed allowed for the determination of the UCG reactor growth, roof spalling, surface subsidence and heat propagation during the UCG process and the subsequent CO2 storage. It is anticipated that the results of this study can support optimisation of the preparation procedure for CO2 storage in former UCG reactors. The proposed scheme was discussed so far, but not validated by a coupled numerical analysis and if proved to be applicable it could provide a significant optimisation of the UCG process by means of CO2 storage efficiency. The proposed coupled UCG-CCS scheme allows for meeting EU targets for greenhouse gas emissions and increases the coal yield otherwise impossible to exploit.

Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Pointer, William David; Sieger, Matt

2016-04-01

The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

Compatibility tests of materials for a lithium-cooled space power reactor concept

NASA Technical Reports Server (NTRS)

Sinclair, J. H.

1973-01-01

Materials for a lithium-cooled space power reactor concept must be chemically compatible for up to 50,000 hr at high temperature. Capsule tests at 1040 C (1900 F) were made of material combinations of prime interest: T-111 in direct contact with uranium mononitride (UN), Un in vacuum separated from T-111 by tungsten wire, UN with various oxygen impurity levels enclosed in tungsten wire lithium-filled T-111 capsules, and TZM and lithium together in T-111 capsules. All combinations were compatible for over 2800 hr except for T-111 in direct contact with UN.

Design of a Mechanical NaK Pump for Fission Space Power Systems

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Bradley, David; Godfroy, Thomas

2010-01-01

Alkali liquid metal cooled fission reactor concepts are under development for mid-range spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid. Traditionally, linear induction pumps have been used to provide the required flow and head conditions for liquid metal systems but can be limited in performance. This paper details the design, build, and check-out test of a mechanical NaK pump. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

LFR "Lead-Cooled Fast Reactor"

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cinotti, L; Fazio, C; Knebel, J

2006-05-11

The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small,more » transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology development already being carried out in different institutes participating in this STREP. This is particularly true in Europe where a large R&D program associated with the development of Accelerator Driven Systems (ADS) is being actively pursued. The general objective of the ELSY project is to design an innovative lead-cooled fast reactor complemented by an analytical effort to assess the existing knowledge base in the field of lead-alloy coolants (i.e., lead-bismuth eutectic (LBE) and also lead/lithium) in order to extrapolate this knowledge base to pure lead. This analysis effort will be complemented with some limited R&D activities to acquire missing or confirmatory information about fundamental topics for ELSY that are not sufficiently covered in the ongoing European ADS program or elsewhere.« less

Nuclear reactor spacer grid and ductless core component

DOEpatents

Christiansen, David W.; Karnesky, Richard A.

1989-01-01

The invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

Expert systems for fault diagnosis in nuclear reactor control

NASA Astrophysics Data System (ADS)

Jalel, N. A.; Nicholson, H.

1990-11-01

An expert system for accident analysis and fault diagnosis for the Loss Of Fluid Test (LOFT) reactor, a small scale pressurized water reactor, was developed for a personal computer. The knowledge of the system is presented using a production rule approach with a backward chaining inference engine. The data base of the system includes simulated dependent state variables of the LOFT reactor model. Another system is designed to assist the operator in choosing the appropriate cooling mode and to diagnose the fault in the selected cooling system. The response tree, which is used to provide the link between a list of very specific accident sequences and a set of generic emergency procedures which help the operator in monitoring system status, and to differentiate between different accident sequences and select the correct procedures, is used to build the system knowledge base. Both systems are written in TURBO PROLOG language and can be run on an IBM PC compatible with 640k RAM, 40 Mbyte hard disk and color graphics.

Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium-Plutonium Carbide based fuel and 300 MWt Power

NASA Astrophysics Data System (ADS)

Clief Pattipawaej, Sandro; Su'ud, Zaki

2017-01-01

A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

PBF Cooling Tower under construction. Cold water basin is five ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Cooling Tower under construction. Cold water basin is five feet deep. Foundation and basin walls are reinforced concrete. Camera facing west. Pipe openings through wall in front are outlets for return flow of cool water to reactor building. Photographer: John Capek. Date: September 4, 1968. INEEL negative no. 68-3473 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Applications of the Aqueous Self-Cooled Blanket concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steiner, D.; Embrechts, M.J.; Varsamis, G.

1986-11-01

In this paper a novel water-cooled blanket concept is examined. This concept, designated the Aqueous Self-Cooled Blanket (ASCB), employs water with small amounts of dissolved fertile compounds as both the coolant and the breeding medium. The ASCB concept is reviewed and its application in three different contexts is examined: (1) power reactors; (2) near-term devices such as NET; and (3) fusion-fission hybrids.

Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

NASA Astrophysics Data System (ADS)

Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

2018-01-01

ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical experiment yields that we could verify the effects of heterogeneity of propagation medium on waves in Liquid sodium.

Sensor Failure Detection of FASSIP System using Principal Component Analysis

NASA Astrophysics Data System (ADS)

Sudarno; Juarsa, Mulya; Santosa, Kussigit; Deswandri; Sunaryo, Geni Rina

2018-02-01

In the nuclear reactor accident of Fukushima Daiichi in Japan, the damages of core and pressure vessel were caused by the failure of its active cooling system (diesel generator was inundated by tsunami). Thus researches on passive cooling system for Nuclear Power Plant are performed to improve the safety aspects of nuclear reactors. The FASSIP system (Passive System Simulation Facility) is an installation used to study the characteristics of passive cooling systems at nuclear power plants. The accuracy of sensor measurement of FASSIP system is essential, because as the basis for determining the characteristics of a passive cooling system. In this research, a sensor failure detection method for FASSIP system is developed, so the indication of sensor failures can be detected early. The method used is Principal Component Analysis (PCA) to reduce the dimension of the sensor, with the Squarred Prediction Error (SPE) and statistic Hotteling criteria for detecting sensor failure indication. The results shows that PCA method is capable to detect the occurrence of a failure at any sensor.

RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hongbin Zhang; Haihua Zhao; Cliff Davis

2008-06-01

An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heatmore » from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment.« less

Nuclear reactor overflow line

DOEpatents

Severson, Wayne J.

1976-01-01

The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

Effect on the tritium breeding ratio for a distributed ICRF antenna in a DEMO reactor

NASA Astrophysics Data System (ADS)

Garcia, A.; Noterdaeme, J.-M.; Fischer, U.; Dies, J.

2015-12-01

The paper reports results of MCNP-5 calculations to assess the effect on the Tritium Breeding Ratio (TBR) when integrating a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna in the blanket of DEMO fusion power reactor. The calculations consider different parameters such as the ICRF covering ratio and the type of breeding blanket including the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) concepts. For an antenna with a full toroidal circumference of 360°, located poloidally at 40° with a poloidal extension of 1 m, the reduction of the TBR is -0.349% for the HCPB blanket and -0.532% for the HCLL blanket. The distributed ICRF antenna is thus shown to have only a marginal effect on the TBR of the DEMO reactor.

Power flattening on modified CANDLE small long life gas-cooled fast reactor

NASA Astrophysics Data System (ADS)

Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

2014-09-01

Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

Fibre optic sensor based measurements of flow-induced vibration in a liquid metal cooled nuclear reactor set-up

NASA Astrophysics Data System (ADS)

De Pauw, B.; Vanlanduit, S.; Van Tichelen, K.; Geernaert, T.; Thienpont, H.; Berghmans, F.

2017-04-01

Fuel assembly vibrations in nuclear reactor cores should not be excessive as these can compromise the lifetime of the assembly and lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. We therefore demonstrate accurate measurements of the vibrations of a fuel assembly in a lead-bismuth eutectic cooled installation with fibre Bragg grating (FBG) based sensors. The use of FBGs in combination with a dedicated sensor integration approach allows accounting for the severe geometrical constraints and providing for the required minimal intrusiveness of the instrumentation, identifying the vibration modes with required accuracy and observing the differences between the vibration amplitudes of the individual fuel pins as well as evidencing a low frequency fuel pin vibration mode resulting from the supports.

Ichthyoplankton entrainment study at the SRS Savannah River water intakes for Westinghouse Savannah River Company

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paller, M.

1992-03-26

Cooling water for L and K Reactors and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pump houses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water and passed through the reactor's heat exchangers where temperatures may reach 70[degrees]C during full power operation. Ichthyoplankton mortality under such conditions is assumed to be 100 percent. The number of ichthyoplankton entrained into the cooling system depends on a variety of variables, including time of year, density and distribution of ichthyoplankton in themore » river, discharge levels in the river, and the volume of water withdrawn by the pumps. Entrainment at the 1 G pump house, which is immediately downstream from the confluence of Upper Three Runs Creek and the Savannah River, is also influenced by discharge rates and ichthyoplankton densities in Upper Three Runs Creek. Because of the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River, the Department of Energy requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory sample ichthyoplankton at the SRS Savannah River intakes. Dams Moore, Inc., under a contract with Westinghouse Savannah River Company performed the sampling and data analysis for the ESS.« less

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perret, G.; Pattupara, R. M.; Girardin, G.

2012-07-01

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fastmore » Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)« less

ETR, TRA642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF FLOOR AND FACES NORTH. OUTER WALL OF STORAGE CANAL IS AT RIGHT. SHIELDING IS THICKER AT LOWER LEVEL, WHERE SPENT FUEL ELEMENTS WILL COOL AFTER REMOVAL FROM REACTOR. INL NEGATIVE NO. 56-1401. Jack L. Anderson, Photographer, 5/1/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

DOEpatents

McCorkle, W.H.

1960-02-23

BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

Convective cooling in a pool-type research reactor

NASA Astrophysics Data System (ADS)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

Westinghouse Small Modular Reactor balance of plant and supporting systems design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Memmott, M. J.; Stansbury, C.; Taylor, C.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operationmore » of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)« less

Thermos reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

Cooling by conversion of para to ortho-hydrogen

NASA Technical Reports Server (NTRS)

Sherman, A. (Inventor)

1983-01-01

The cooling capacity of a solid hydrogen cooling system is significantly increased by exposing vapor created during evaporation of a solid hydrogen mass to a catalyst and thereby accelerating the endothermic para-to-ortho transition of the vapor to equilibrium hydrogen. Catalyst such as nickel, copper, iron or metal hydride gels of films in a low pressure drop catalytic reactor are suitable for accelerating the endothermic para-to-ortho conversion.

Agile Port and High Speed Ship Technologies, Vol 1: FY05 Projects 3-6 and 8-10

DTIC Science & Technology

2008-07-02

Computational Fluid Dynamics DTMB - David Taylor Model Basin JVR - Jet Velocity Ratio NSWCCD - Naval Surface Warfare Center, Carderock Division SDD - Systems...immature current state of the technology employed for the reactor system (multiple closed Brayton Cycle, Helium Cooled Gas reactors); (iii) several

78 FR 53482 - Entergy Operations, Inc., River Bend Station, Unit 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-29

... facility consists of a boiling-water reactor located in West Feliciana Parish, Louisiana. 2.0 Request... Containment Leakage Testing for Water- Cooled Power Reactors,'' requires that components which penetrate containment be periodically leak tested at the ``P a, '' defined as the ``calculated peak containment internal...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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.

MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

DOEpatents

Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

1962-06-26

A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

77 FR 5063 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-01

... Subcommittee on Thermal-Hydraulics Phenomena; Notice of Meeting The ACRS Subcommittee on Thermal-Hydraulics... Regulatory Guide (1.79), ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water... Water Reactors.'' The Subcommittee will hear presentations by and hold discussions with the NRC staff...

Reactor pressure vessel with forged nozzles

DOEpatents

Desai, Dilip R.

1993-01-01

Inlet nozzles for a gravity-driven cooling system (GDCS) are forged with a cylindrical reactor pressure vessel (RPV) section to which a support skirt for the RPV is attached. The forging provides enhanced RPV integrity around the nozzle and substantial reduction of in-service inspection costs by eliminating GDCS nozzle-to-RPV welds.

Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

Quenching behavior of molten pool with different strategies – A review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shrikant,, E-mail: 2014rmt9018@mnit.ac.in; Pandel, U.; Duchaniya, R. K.

After the major severe accident in nuclear reactor, there has been lot of concerns regarding long term core melt stabilization following a severe accident in nuclear reactors. Numerous strategies have been though for quenching and stabilization of core melt like top flooding, bottom flooding, indirect cooling, etc. However, the effectiveness of these schemes is yet to be determined properly, for which, lot of experiments are needed. Several experiments have been performed for coolability of melt pool under bottom flooding as well as for indirect cooling. Besides these tests are very scattered because they involve different simulants material initial temperatures andmore » masses of melt, which makes it very complex to judge the effectiveness of a particular technique and advantage over the other. In this review paper, a study has been carried on different cooling techniques of simulant materials with same mass. Three techniques have been compared here and the results are discussed. Under top flooding technique it took several hours to cool the melt under without decay heat condition. In bottom flooding technique was found to be the best technique among in indirect cooling technique, top flooded technique, and bottom flooded technique.« less

ETR COMPLEX. CAMERA FACING EAST. FROM LEFT TO RIGHT: ETRCRITICAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR COMPLEX. CAMERA FACING EAST. FROM LEFT TO RIGHT: ETR-CRITICAL FACILITY BUILDING, ETR CONTROL BUILDING (ATTACHED TO HIGH-BAY ETR), ETR, ONE-STORY SECTION OF ETR BUILDING, ELECTRICAL BUILDING, COOLING TOWER PUMP HOUSE, COOLING TOWER. COMPRESSOR AND HEAT EXCHANGER BUILDING ARE PARTLY IN VIEW ABOVE ETR. DARK-COLORED DUCTS PROCEED FROM GROUND CONNECTION TO ETR WASTE GAS STACK. OTHER STACK IS MTR STACK WITH FAN HOUSE IN FRONT OF IT. RECTANGULAR STRUCTURE NEAR TOP OF VIEW IS SETTLING BASIN. INL NEGATIVE NO. 56-4102. Unknown Photographer, ca. 1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Device for cooling and humidifying reformate

DOEpatents

Zhao, Jian Lian; Northrop, William F.

2008-04-08

Devices for cooling and humidifying a reformate stream from a reforming reactor as well as related methods, modules and systems includes a heat exchanger and a sprayer. The heat exchanger has an inlet, an outlet, and a conduit between the inlet and the outlet. The heat exchanger is adapted to allow a flow of a first fluid (e.g. water) inside the conduit and to establish a heat exchange relationship between the first fluid and a second fluid (e.g. reformate from a reforming reactor) flowing outside the conduit. The sprayer is coupled to the outlet of the heat exchanger for spraying the first fluid exiting the heat exchanger into the second fluid.

Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

NASA Astrophysics Data System (ADS)

Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

2014-04-01

CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

Method for fabricating wrought components for high-temperature gas-cooled reactors and product

DOEpatents

Thompson, Larry D.; Johnson, Jr., William R.

1985-01-01

A method and alloys for fabricating wrought components of a high-temperature gas-cooled reactor are disclosed. These wrought, nickel-based alloys, which exhibit strength and excellent resistance to carburization at elevated temperatures, include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength.

MANUFACTURE OF UF$sub 4$

DOEpatents

Calcott, W.S.

1959-10-13

The manufacture of uranium tetrafluoride from urarium dioxide is described. Uranium dioxide is heated to about 500 deg C in a reactor. Anhydrous hydrogen fluoride is passed through the reactor in contact with uranium dioxide for several hours, the flow of hydrogen fluoride is discontinued, and hydrogen passed through the reactor for less than an hour. The flow of hydrogen fluoride is resumed for several hours, and then nitrogen is passed for a few minutes to expel unreacted hydrogen fluoride as water vapor. The reactor is cooled to room temperature and the uranium tetrafluoride removed.

Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoit, J. C.; Bourdot, P.; Eschbach, R.

2012-07-01

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

Prospective scenarios of nuclear energy evolution over the 21. century

DOE Office of Scientific and Technical Information (OSTI.GOV)

Massara, S.; Tetart, P.; Garzenne, C.

2006-07-01

In this paper, different world scenarios of nuclear energy development over the 21. century are analyzed, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE - STRATEGIE. Three nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on natural uranium consumption and industrial requirements related to the nuclear reactors and the associated fuel cycle facilities. Both thermal-spectrum systems (Pressurized Water Reactor and High Temperature Gas-cooled Reactor) and Fast Reactors are investigated. (authors)

Packed fluidized bed blanket for fusion reactor

DOEpatents

Chi, John W. H.

1984-01-01

A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

Reactor for fluidized bed silane decomposition

NASA Technical Reports Server (NTRS)

Iya, Sridhar K. (Inventor)

1989-01-01

An improved heated fluidized bed reactor and method for the production of high purity polycrystalline silicon by silane pyrolysis wherein silicon seed particles are heated in an upper heating zone of the reactor and admixed with particles in a lower zone, in which zone a silane-containing gas stream, having passed through a lower cooled gas distribution zone not conducive to silane pyrolysis, contacts the heated seed particles whereon the silane is heterogeneously reduced to silicon.

Transient Effects in Turbulence Modelling.

DTIC Science & Technology

1979-12-01

plenum region of a liquid-metal- cooled fast breeder reactor (LMFBR). The efficient heat transfer characteristics of liquid metal coolant, combined...Transients in Generalized Liquid-Metal Fast Breeder Reactor Outlet Plenums," Nuclear Technology, Vol. 44, July 1979, p. 210. 135 15. Lorenz, J. J., "MIX... Sodium Coolant in the Outlet Plenum of a Fast Nuclear Reactor ," Int. J. Heat Mass Transfer, Vol. 21, 1978, pp. 1565-1579. 19. Chen, Y. B., Golay, M. W

Solar Power Satellites - A Review of the Space Transportation Options.

DTIC Science & Technology

1980-03-01

already exists with such systems, gained mainly through liquid-metal breeder reactor programmes. 0 For example, inlet temperatures of 970 C can be handled...alternatives exist. In addition, there would be extreme reluctance on the part of most governments to allow large C- reactors , producing gigawatts of power, to...antenna. The reactors employed are high-temperature gas- cooled breeders , which convert U238 into fissile plutonium. Each of the modules includes a

Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

Removal of dissolved and colloidal silica

DOEpatents

Midkiff, William S.

2002-01-01

Small amorphous silica particles are used to provide a relatively large surface area upon which silica will preferentially adsorb, thereby preventing or substantially reducing scaling caused by deposition of silica on evaporative cooling tower components, especially heat exchange surfaces. The silica spheres are contacted by the cooling tower water in a sidestream reactor, then separated using gravity separation, microfiltration, vacuum filtration, or other suitable separation technology. Cooling tower modifications for implementing the invention process have been designed.

Cooling molten salt reactors using "gas-lift"

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

2014-08-01

This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

Method of detecting leakage of reactor core components of liquid metal cooled fast reactors

DOEpatents

Holt, Fred E.; Cash, Robert J.; Schenter, Robert E.

1977-01-01

A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE PAGES

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

2018-02-02

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanson, G.H.

1978-06-05

The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. LTR 113-47 has shown that the LOFT ECCS can be safely bypassed or disabled when the total core power does not exceed 25 kW. A modified policy involves disabling the automatic actuation of the LOFT ECCS, but still retaining the manualmore » activation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 70 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS.« less

Effect of microstructure on the corrosion of CVD-SiC exposed to supercritical water

NASA Astrophysics Data System (ADS)

Tan, L.; Allen, T. R.; Barringer, E.

2009-10-01

Silicon carbide (SiC) is an important engineering material being studied for potential use in multiple nuclear energy systems including high-temperature gas-cooled reactors and water-cooled reactors. The corrosion behavior of SiC exposed to supercritical water (SCW) is critical for examining its applications in nuclear reactors. Although the hydrothermal corrosion of SiC has been the subject of many investigations, the study on the microstructural effects on the corrosion is limited. This paper presents the effect of residual strain, grain size, grain boundary types, and surface orientations on the corrosion of chemical vapor deposited (CVD) β-SiC exposed to SCW at 500 °C and 25 MPa. Weight loss occurred on all the samples due to localized corrosion. Residual strains associated with small grains showed the most significant effect on the corrosion compared to the other factors.

Effect on the tritium breeding ratio for a distributed ICRF antenna in a DEMO reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garcia, A., E-mail: albert.garcia.hp@gmail.com; Karlsruhe Institute of Technology; Polytechnic University of Catalonia

The paper reports results of MCNP-5 calculations to assess the effect on the Tritium Breeding Ratio (TBR) when integrating a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna in the blanket of DEMO fusion power reactor. The calculations consider different parameters such as the ICRF covering ratio and the type of breeding blanket including the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) concepts. For an antenna with a full toroidal circumference of 360°, located poloidally at 40° with a poloidal extension of 1 m, the reduction of the TBR is −0.349% for the HCPB blanket andmore » −0.532% for the HCLL blanket. The distributed ICRF antenna is thus shown to have only a marginal effect on the TBR of the DEMO reactor.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Method of shielding a liquid-metal-cooled reactor

DOEpatents

Sayre, Robert K.

1978-01-01

The primary heat transport system of a nuclear reactor -- particularly for a liquid-metal-cooled fast-breeder reactor -- is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system.

State of Fukushima nuclear fuel debris tracked by Cs137 in cooling water.

PubMed

Grambow, B; Mostafavi, M

2014-11-01

It is still difficult to assess the risk originating from the radioactivity inventory remaining in the damaged Fukushima nuclear reactors. Here we show that cooling water analyses provide a means to assess source terms for potential future releases. Until now already about 34% of the inventories of (137)Cs of three reactors has been released into water. We found that the release rate of (137)Cs has been constant for 2 years at about 1.8% of the inventory per year indicating ongoing dissolution of the fuel debris. Compared to laboratory studies on spent nuclear fuel behavior in water, (137)Cs release rates are on the higher end, caused by the strong radiation field and oxidant production by water radiolysis and by impacts of accessible grain boundaries. It is concluded that radionuclide analyses in cooling water allow tracking of the conditions of the damaged fuel and the associated risks.

Earth storable bimodal engine, phase 1

NASA Technical Reports Server (NTRS)

1973-01-01

An in-depth study of an Earth Storable Bimodal (ESB) Engine using earth storable propellants N2O/N2H4 and operating in either a monopropellant or bipropellant mode was conducted. Detailed studies were completed for both a hot-gas, regeneratively cooled thrust chamber and a ducted hot-gas, film cooled thrust chamber. Hydrazine decomposition products were used for cooling in either configuration. The various arrangements and configurations of hydrazine reactors, secondary injectors, chambers and gimbal methods were considered. The two basic materials selected for the major components were columbium alloys and L-605. The secondary injector types considered were previously demonstrated by JPL and consisted of a liquid-on-gas triplet, a liquid-on-gas doublet, and a liquid-on-gas coaxial injector. Various design tradeoffs were made with different reactor types located at: the secondary injector station, the thrust chamber throat, and the nozzle/extension interface. Associated thermal, structural, and mass analyses were completed.

76 FR 48184 - Exelon Nuclear, Peach Bottom Atomic Power Station, Unit 1; Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-08

... nuclear reactor facility. PBAPS Unit 1 was a high-temperature, gas-cooled reactor that was operated from... the safeguards contingency plan.'' Part 73 of 10 CFR, ``Physical Protection of Plant and Materials... physical protection system which will have capabilities for the protection of special nuclear material at...

75 FR 76051 - Northern States Power Company-Minnesota, Prairie Island Nuclear Generating Plant, Units 1 and 2...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-07

..., 2010 (Agencywide Documents Access and Management System Accession Nos. ML093280883 and ML101480083... systems for light-water nuclear power reactors,'' and appendix K to 10 CFR part 50, ``ECCS Evaluation... core cooling system (ECCS) for reactors fueled with zircaloy or ZIRLO\\TM\\ cladding. In addition...

The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, R.R.

1986-01-01

This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

An intrinsically safe facility for forefront research and training on nuclear technologies

NASA Astrophysics Data System (ADS)

Mansani, L.; Monti, S.; Ricco, G.; Ricotti, M.

2014-04-01

In this short paper the motivations for the development of fast spectrum lead-cooled reactors are briefly summarized. In particular the importance of subcritical research reactors, like the one described in this Focus Point, for the investigation of various scientifical and technological aspects and the training of students, is discussed.

77 FR 7610 - Notice of Availability of Environmental Assessment and Finding of No Significant Impact for...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-13

... and reinforced concrete floors acting as diaphragms in distributing loads to vertically resisting... reinforced concrete foundation. The reactor is fueled with standard low-enriched TRIGA (Training, Research... cooled by a light water primary system consisting of the reactor pool and a heat removal system to remove...

76 FR 53979 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-30

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal Hydraulics Phenomena; Notice of Meeting The ACRS Subcommittee on Thermal Hydraulics... Revision 4 to Regulatory Guide 1.82, ``Water Sources for Long-Term Recirculation Cooling Following a Loss...

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Evaluation of an Integrated Gas-Cooled Reactor Simulator and Brayton Turbine-Generator

NASA Technical Reports Server (NTRS)

Hissam, David Andy; Stewart, Eric T.

2006-01-01

A closed-loop brayton cycle, powered by a fission reactor, offers an attractive option for generating both planetary and in-space electric power. Non-nuclear testing of this type of system provides the opportunity to safely work out integration and system control challenges for a modest investment. Recognizing this potential, a team at Marshall Space Flight Center has evaluated the viability of integrating and testing an existing gas-cooled reactor simulator and a modified commercially available, off-the-shelf, brayton turbine-generator. Since these two systems were developed independently of one another, this evaluation had to determine if they could operate together at acceptable power levels, temperatures, and pressures. Thermal, fluid, and structural analyses show that this combined system can operate at acceptable power levels and temperatures. In addition, pressure drops across the reactor simulator, although higher than desired, are also viewed as acceptable. Three potential working fluids for the system were evaluated: N2, He/Ar, and He/Xe. Other potential issues, such as electrical breakdown in the generator and the operation of the brayton foil bearings using various gas mixtures, were also investigated.

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

Closure of the R Reactor Disassembly Basin at the SRS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Austin, W.E.

The Facilities Disposition Division (FDD) at the Savannah River Site is engaged in planning the deactivation/closure of three of the site's five reactor disassembly basins. Activities are currently underway at R-Reactor Disassembly Basin and will continue with the P and C disassembly basins. The basins still contain the cooling and shielding water that was present when operations ceased. Low concentrations of radionuclides are present, with tritium, Cs-137, and Sr-90 being the major contributors. Although there is no evidence that any of the basins have leaked, the 50-year-old facilities will eventually contaminate the surrounding groundwaters. The FDD is pursuing a pro-activemore » solution to close the basins in-place and prevent a release to the groundwater. In-situ ion exchange is currently underway at the R-Reactor Disassembly Basin to reduce the Cs and Sr concentrations to levels that would allow release of the treated water to previously used on-site cooling ponds or to prevent ground water impact. The closure will be accomplished under CERCLA.« less

Reliability Analysis of RSG-GAS Primary Cooling System to Support Aging Management Program

NASA Astrophysics Data System (ADS)

Deswandri; Subekti, M.; Sunaryo, Geni Rina

2018-02-01

Multipurpose Research Reactor G.A. Siwabessy (RSG-GAS) which has been operating since 1987 is one of the main facilities on supporting research, development and application of nuclear energy programs in BATAN. Until now, the RSG-GAS research reactor has been successfully operated safely and securely. However, because it has been operating for nearly 30 years, the structures, systems and components (SSCs) from the reactor would have started experiencing an aging phase. The process of aging certainly causes a decrease in reliability and safe performances of the reactor, therefore the aging management program is needed to resolve the issues. One of the programs in the aging management is to evaluate the safety and reliability of the system and also screening the critical components to be managed.One method that can be used for such purposes is the Fault Tree Analysis (FTA). In this papers FTA method is used to screening the critical components in the RSG-GAS Primary Cooling System. The evaluation results showed that the primary isolation valves are the basic events which are dominant against the system failure.

First-wall structural analysis of the self-cooled water blanket concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Brien, D.A.; Steiner, D.; Embrechts, M.J.

1986-01-01

A novel blanket concept recently proposed utilizes water with small amounts of dissolved lithium compound as both coolant and breeder. The inherent simplicity of this idea should result in an attractive breeding blanket for fusion reactors. In addition, the available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate this concept. First-wall and blanket designs have been developed first for the tandem mirror reactor (TMR) due to the obvious advantages of this geometry. First-wall and blanket designs will also be developed for toroidal reactors. A simple plate designmore » with coolant tubes welded on the back (side away from plasma) was chosen as the first wall for the TMR application. Dimensions and materials were chosen to minimize temperature differences and thermal stresses. A finite element code (STRAW), originally developed for the analysis of core components subjected to high-pressure transients in the fast breeder program, was utilized to evaluate stresses in the first wall.« less

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

Access to small size distributions of nanoparticles by microwave-assisted synthesis. Formation of Ag nanoparticles in aqueous carboxymethylcellulose solutions in batch and continuous-flow reactors

NASA Astrophysics Data System (ADS)

Horikoshi, Satoshi; Abe, Hideki; Torigoe, Kanjiro; Abe, Masahiko; Serpone, Nick

2010-08-01

This article examines the effect(s) of the 2.45-GHz microwave (MW) radiation in the synthesis of silver nanoparticles in aqueous media by reduction of the diaminesilver(i) complex, [Ag(NH3)2]+, with carboxymethylcellulose (CMC) in both batch-type and continuous-flow reactor systems with a particular emphasis on the characteristics of the microwaves in this process and the size distributions. This microwave thermally-assisted synthesis is compared to a conventional heating (CH) method, both requiring a reaction temperature of 100 °C to produce the nanoparticles, in both cases leading to the formation of silver colloids with different size distributions. Reduction of the diaminesilver(i) precursor complex, [Ag(NH3)2]+, by CMC depended on the solution temperature. Cooling the reactor during the heating process driven with 390-Watt microwaves (MW-390W/Cool protocol) yielded silver nanoparticles with sizes spanning the range 1-2 nm. By contrast, the size distribution of Ag nanoparticles with 170-Watt microwaves (no cooling; MW-170W protocol) was in the range 1.4-3.6 nm (average size ~3 nm). The overall results suggest the potential for a scale-up process in the microwave-assisted synthesis of nanoparticles. Based on the present data, a flow-through microwave reactor system is herein proposed for the continuous production of silver nanoparticles. The novel flow reactor system (flow rate, 600 mL min-1) coupled to 1200-Watt microwave radiation generated silver nanoparticles with a size distribution 0.7-2.8 nm (average size ca. 1.5 nm).

Scaling analysis for the direct reactor auxiliary cooling system for FHRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lv, Q.; Kim, I. H.; Sun, X.

2015-04-01

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and tomore » activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from both the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a scientific design of a scaled-down high-temperature DRACS test facility.« less

Low-cost, compact, cooled photomultiplier assembly for use in magnetic fields up to 1400 Gauss

NASA Technical Reports Server (NTRS)

Patch, R. W.; Tashjian, R. A.; Jentner, T. A.

1975-01-01

Use of vortex tube for cooling and concentric shielding have produced smaller and more compact unit than was previously available. Future uses of device could include installation in gas chromatographs and mass spectrometers. Additional uses would include measurements and controls in magnetohydrodynamic power generators and fusion reactors.

Bulge-Formed Cooling Channels In A Wall

NASA Technical Reports Server (NTRS)

Mcaninch, Michael D.; Holbrook, Richard L.; Lacount, Dale F.; Kawashige, Chester M.; Crapuchettes, John M.; Scala, James

1996-01-01

Vessels bounded by walls shaped as surfaces of revolution and contain integral cooling channels fabricated by improved method involving combination of welding and bulge forming. Devised to make rocket nozzles; also useful in fabrication of heat exchangers, stationary combustion chambers, and chemical-reactor vessels. Advantages include easier fabrication and greater flexibility of design.

The Potential of Different Concepts of Fast Breeder Reactor for the French Fleet Renewal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Massara, Simone; Tetart, Philippe; Lecarpentier, David

2006-07-01

The performances of different concepts of Fast Breeder Reactor (Na-cooled, He-cooled and Pb-cooled FBR) for the current French fleet renewal are analyzed in the framework of a transition scenario to a 100% FBR fleet at the end of the 21. century. Firstly, the modeling of these three FBR types by means of a semi-analytical approach in TIRELIRE - STRATEGIE, the EDF fuel cycle simulation code, is presented, together with some validation elements against ERANOS, the French reference code system for neutronic FBR analysis (CEA). Afterwards, performances comparisons are made in terms of maximum deployable power, natural uranium consumption and wastemore » production. The results show that the FBR maximum deployable capacity, independently from the FBR technology, is highly sensitive to the fuel cycle options, like the spent nuclear fuel cooling time or the Minor Actinides management strategy. Thus, some of the key parameters defining the dynamic of FBR deployment are highlighted, to inform the orientation of R and D in the development and optimization of these systems. (authors)« less

Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Y.A.; Feltus, M.A.

1995-07-01

Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specificmore » MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company`s Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates.« less

Gas-cooled nuclear reactor

DOEpatents

Peinado, Charles O.; Koutz, Stanley L.

1985-01-01

A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trond Bjornard; John Hockert

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 reactormore » 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 summarized in this report, namely the development of guidance to support the consideration of MC&A in the design of both pebble-bed and prismatic-fueled HTGRs. The objective is to identify and incorporate design features into the facility design that will cost effectively aid in making MC&A more effective and efficient, with minimum impact on operations. The theft of nuclear material is addressed through both MC&A and physical protection, while the threat of sabotage is addressed principally through physical protection.« less

Small Modular Reactors: The Army’s Secure Source of Energy?

DTIC Science & Technology

2012-03-21

significant advantages of SMRs is the minimal amount of carbon dioxide (greenhouse gases) that is released in conjunction with the lifecycle operations...moderator in these reactors as well as the cooling agent and the means by which heat is removed to produce steam for turning the turbines of the...separate water system to generate steam to turn a turbine which then produces electricity. In the second type of light water reactors, the boiling water

Computer modeling and simulators as part of university training for NPP operating personnel

NASA Astrophysics Data System (ADS)

Volman, M.

2017-01-01

This paper considers aspects of a program for training future nuclear power plant personnel developed by the NPP Department of Ivanovo State Power Engineering University. Computer modeling is used for numerical experiments on the kinetics of nuclear reactors in Mathcad. Simulation modeling is carried out on the computer and full-scale simulator of water-cooled power reactor for the simulation of neutron-physical reactor measurements and the start-up - shutdown process.

Jet pump-drive system for heat removal

NASA Technical Reports Server (NTRS)

French, James R. (Inventor)

1987-01-01

The invention does away with the necessity of moving parts such as a check valve in a nuclear reactor cooling system. Instead, a jet pump, in combination with a TEMP, is employed to assure safe cooling of a nuclear reactor after shutdown. A main flow exists for a reactor coolant. A point of withdrawal is provided for a secondary flow. A TEMP, responsive to the heat from said coolant in the secondary flow path, automatically pumps said withdrawn coolant to a higher pressure and thus higher velocity compared to the main flow. The high velocity coolant is applied as a driver flow for the jet pump which has a main flow chamber located in the main flow circulation pump. Upon nuclear shutdown and loss of power for the main reactor pumping system, the TEMP/jet pump combination continues to boost the coolant flow in the direction it is already circulating. During the decay time for the nuclear reactor, the jet pump keeps running until the coolant temperature drops to a lower and safe temperature where the heat is no longer a problem. At this lower temperature, the TEMP/jet pump combination ceases its circulation boosting operation. When the nuclear reactor is restarted and the coolant again exceeds the lower temperature setting, the TEMP/jet pump automatically resumes operation. The TEMP/jet pump combination is thus automatic, self-regulating and provides an emergency pumping system free of moving parts.

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ariani, Menik; Su'ud, Zaki; Waris, Abdul

2012-06-06

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it ismore » shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.« less

Convective cooling in a pool-type research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sipaun, Susan, E-mail: susan@nm.gov.my; Usman, Shoaib, E-mail: usmans@mst.edu

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to passmore » through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.« less

Decommissioning of German Nuclear Research Facilities under the Governance of the Federal Ministry of Education and Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weigl, M.

2008-07-01

Since the announcement of the first nuclear program in 1956, nuclear R and D in Germany has been supported by the Federal Government under four nuclear programs and later on under more general energy R and D programs. The original goal was to help German industry to achieve safe, low-cost generation of energy and self-sufficiency in the various branches of nuclear technology, including the fast breeder reactor and the fuel cycle. Several national research centers were established to host or operate experimental and demonstration plants. These are mainly located at the sites of the national research centers at Juelich andmore » Karlsruhe. In the meantime, all these facilities were shut down and most of them are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. For two other projects the return to 'green field' sites will be reached by the end of this decade. These are the dismantling of the Multi-Purpose Research Reactor (MZFR) and the Compact Sodium Cooled Reactor (KNK) both located at the Forschungszentrum Karlsruhe. Within these projects a lot of new solutions und innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). For example, high performance underwater cutting technologies like plasma arc cutting and contact arc metal cutting. (authors)« less

Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, Steven A.; Sanchez, Travis

2005-02-06

The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion componentsmore » such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)« less

Dismantlement of the TSF-SNAP Reactor Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

Natural Convection Heat Transfer in a Rectangular Liquid Metal Pool With Bottom Heating and Top Cooling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Il S.; Yu, Yong H.; Son, Hyoung M.

2006-07-01

An experimental study is performed to investigate the natural convection heat transfer characteristics with subcooled coolant to create engineering database for basic applications in a lead alloy cooled reactor. Tests are performed in the ALTOS (Applied Liquid-metal Thermal Operation Study) apparatus as part of MITHOS (Metal Integrated Thermo Hydrodynamic Operation System). A relationship is determined between the Nusselt number Nu and the Rayleigh number Ra in the liquid metal rectangular pool. Results are compared with correlations and experimental data in the literature. Given the similar Ra condition, the present test results for Nu of the liquid metal pool with topmore » subcooling are found to be similar to those predicted by the existing correlations or experiments. The current test results are utilized to develop natural convection heat transfer correlations applicable to low Prandtl number Pr fluids that are heated from below and cooled by the external coolant above. Results from this study are slated to be used in designing BORIS (Battery Optimized Reactor Integral System), a small lead cooled modular fast reactor for deployment at remote sites cycled with MOBIS (Modular Optimized Brayton Integral System) for electricity generation, tied with NAVIS (Naval Application Vessel Integral System) for ship propulsion, joined with THAIS (Thermochemical Hydrogen Acquisition Integral System) for hydrogen production, and coupled with DORIS (Desalination Optimized Reactor Integral System) for seawater desalination. Tests are performed with Wood's metal (Pb-Bi-Sn-Cd) filling a rectangular pool whose lower surface is heated and upper surface cooled by forced convection of water. The test section is 20 cm long, 11.3 cm high and 15 cm wide. The simulant has a melting temperature of 78 deg. C. The constant temperature and heat flux condition was realized for the bottom heating once the steady state had been met. The test parameters include the heated bottom surface temperature of the liquid metal pool, the input power to the bottom surface of the section, and the coolant temperature. (authors)« less

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

2008-06-23

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.« less

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Per F.

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 ofmore » refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.« less

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Koenig, D. R.

1978-01-01

One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

STEEL FOR PRESSURE VESSELS FOR POWER REACTORS (in German)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zastrow, E.

1960-11-01

Both gas-cooled and water-cooled reactors place on the steel pressure vessel rigid requirements with respect to the design, radiation stability, gamma -induced internal stresses, and inability to, or difficulty in, repairing the vessel once it is installed. The factors to be considered in the selection of a given steel for a pressure vessel are reviewed, and the properties of steels previously used for this purpose are tabulated. The studies being raade at present to improve the desirable properties of steels for pressure vessels are briefly summarized. The corrosion stability and irradiation stability of steel are discussed. Neutron activation of themore » steel is also briefly reviewed. (J.S.R.)« less

Impact of nuclear data on sodium-cooled fast reactor calculations

NASA Astrophysics Data System (ADS)

Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

2016-03-01

Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

21. View from the work area of the front face ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

21. View from the work area of the front face of the pile in the 105 building, in this case at the F Reactor in February 1945. The 2,004 pigtails and process tube nozzles are neatly aligned in rows and columns across the face of the pile. The cooling water risers stand at the left and right of the pile and the distribution crossheaders run across its face. The pipes running vertically at the bottom of the pile carry cooling water to the thermal shield. The low railing along the floor in front of the face prevented workers from accidentally falling into the charging elevator pit. D-8326 - B Reactor, Richland, Benton County, WA

Silicon Carbide as a tritium permeation barrier in tungsten plasma-facing components

NASA Astrophysics Data System (ADS)

Wright, G. M.; Durrett, M. G.; Hoover, K. W.; Kesler, L. A.; Whyte, D. G.

2015-03-01

The control of tritium inventory is of great importance in future fusion reactors, not only from a safety standpoint but also to maximize a reactor's efficiency. Due to the high mobility of hydrogenic species in tungsten (W) one concern is the loss of tritium from the system via permeation through the tungsten plasma-facing components (PFC). This can lead to loss of tritium through the cooling channels of the wall thereby mandating tritium monitoring and recovery methods for the cooling system of the first wall. The permeated tritium is then out of the fuel cycle and cannot contribute to energy production until it is recovered and recycled into the system.

Temperature Swing Adsorption Compressor Development

NASA Technical Reports Server (NTRS)

Finn, John E.; Mulloth, Lila M.; Affleck, Dave L.

2001-01-01

Closing the oxygen loop in an air revitalization system based on four-bed molecular sieve and Sabatier reactor technology requires a vacuum pump-compressor that can take the low-pressure CO, from the 4BMS and compress and store for use by a Sabatier reactor. NASA Ames Research Center proposed a solid-state temperature-swing adsorption (TSA) compressor that appears to meet performance requirements, be quiet and reliable, and consume less power than a comparable mechanical compressor/accumulator combination. Under this task, TSA compressor technology is being advanced through development of a complete prototype system. A liquid-cooled TSA compressor has been partially tested, and the rest of the system is being fabricated. An air-cooled TSA compressor is also being designed.

The Fukushima Nuclear Disaster and the U.S. Customs and Border Protection Response

NASA Astrophysics Data System (ADS)

McCormick, Kathy

2013-10-01

On 3/11/11, the reactors at the Fukushima Nuclear Plant in Japan were damaged by a magnitude 9.0 earthquake. Of the six reactors at the site, three were in operation prior to the event, and were automatically shut-down during the earthquake. Emergency cooling systems came online and were subsequently destroyed by a tsunami generated by the earthquake. For the operating reactors, all the reactor cores were exposed, resulting in overheating and the release of steam and hydrogen gas to the containment vessels, several of which subsequently exploded, releasing radioactivity into the atmosphere. The cores of the operating reactors melted down, and radioactive water was released to the ocean in cooling efforts. The primary radiation concerns in the United States from the disaster were radioactive plumes driven by westerly winds and contaminated commercial products and travelers. In the United States, one of the primary governmental organizations to respond to the disaster was U.S. Customs and Border Protection (CBP), which has responsibility to oversee the safety and security of cargo and travelers entering the United States. This talk will describe the various types of radioactive commodities and events encountered by CBP in the U.S. from the Fukushima disaster. Thanks to the CBP Teleforensics Center for their assistance with this presentation.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.