Trench fast reactor design using the microcomputer

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

Rohach, A.F.; Sankoorikal, J.T.; Schmidt, R.R.

1987-01-01

This project is a study of alternative liquid-metal-cooled fast power reactor system concepts. Specifically, an unconventional primary system is being conceptually designed and evaluated. The project design is based primarily on microcomputer analysis through the use of computational modules. The reactor system concept is a long, narrow pool with a long, narrow reactor called a trench-type pool reactor in it. The reactor consists of five core-blanket modules in a line. Specific power is to be modest, permitting long fuel residence time. Two fuel cycles are currently being considered. The reactor design philosophy is that of the inherently safe concept. Thismore » requires transient analysis dependent on reactivity coefficients: prompt fuel, including Doppler and expansion, fuel expansion, sodium temperature and void, and core expansion. Conceptual reactor design is done on a microcomputer. A part of the trench reactor project is to develop a microcomputer-based system that can be used by the user for scoping studies and design. Current development includes the neutronics and fuel management aspects of the design. Thermal-hydraulic analysis and economics are currently being incorporated into the microcomputer system. The system is menu-driven including preparation of program input data and of output data for displays in graphics form.« less

Demand driven salt clean-up in a molten salt fast reactor – Defining a priority list

PubMed Central

Litskevich, D.; Gregg, R.; Mount, A. R.

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified. PMID:29494604

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

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Experiences in utilization of research reactors in Yugoslavia

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

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro envisage the role of research reactors in the promotion of nuclear power programs in relation to the size of the program, the competence of domestic industries and the degree of independence where fuel supply is concerned. (author)« less

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

Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W'smore » proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)« less

CHEMICAL ENGINEERING DIVISION SUMMARY REPORT, OCTOBER, NOVEMBER, DECEMBER 1960

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

None

1961-03-01

Chemical-metallurgical processing studies were made of pyrometallurgical development snd research, and fuel processing facilities for EBR-II. Fuel-cycle applications of fluidization and volatility techniques included laboratory investigations of fluoride volatility processes, engineeringscale development, and conversion of UF/sub 6/ to UO/sub 2/. Reactor safety studies consisted of metal oxidation and ignition kinetics, and metal-water reactions. Reactor chemistry investigations were conducted to determine nuclear constants and suitable reactor decontamination methods. Routine operations are summarized for the high-level gammairradiation facillty and waste processing. (B.O.G.)

Summary of NR Program Prometheus Efforts

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

J Ashcroft; C Eshelman

2006-02-08

The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

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

Tan, Lizhen; Yang, Ying; Sridharan, K.

2015-12-01

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less

Liquid fuel molten salt reactors for thorium utilization

DOE PAGES

Gehin, Jess C.; Powers, Jeffrey J.

2016-04-08

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

Small space reactor power systems for unmanned solar system exploration missions

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

NASA-EPA automotive thermal reactor technology program

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Hibbard, R. R.

1972-01-01

The status of the NASA-EPA automotive thermal reactor technology program is summarized. This program is concerned primarily with materials evaluation, reactor design, and combustion kinetics. From engine dynamometer tests of candidate metals and coatings, two ferritic iron alloys (GE 1541 and Armco 18-SR) and a nickel-base alloy (Inconel 601) offer promise for reactor use. None of the coatings evaluated warrant further consideration. Development studies on a ceramic thermal reactor appear promising based on initial vehicle road tests. A chemical kinetic study has shown that gas temperatures of at least 900 K to 1000 K are required for the effective cleanup of carbon monoxide and hydrocarbons, but that higher temperatures require shorter combustion times and thus may permit smaller reactors.

Advanced Demonstration and Test Reactor Options Study

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

Petti, David Andrew; Hill, R.; Gehin, J.

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercializationmore » of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy”. Advanced reactors are defined in this study as reactors that use coolants other than water. Advanced reactor technologies have the potential to expand the energy applications, enhance the competitiveness, and improve the sustainability of nuclear energy.« less

Small and medium power reactors 1987

NASA Astrophysics Data System (ADS)

1987-12-01

This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

Simulation models and designs for advanced Fischer-Tropsch technology

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

Choi, G.N.; Kramer, S.J.; Tam, S.S.

1995-12-31

Process designs and economics were developed for three grass-roots indirect Fischer-Tropsch coal liquefaction facilities. A baseline and an alternate upgrading design were developed for a mine-mouth plant located in southern Illinois using Illinois No. 6 coal, and one for a mine-mouth plane located in Wyoming using Power River Basin coal. The alternate design used close-coupled ZSM-5 reactors to upgrade the vapor stream leaving the Fischer-Tropsch reactor. ASPEN process simulation models were developed for all three designs. These results have been reported previously. In this study, the ASPEN process simulation model was enhanced to improve the vapor/liquid equilibrium calculations for themore » products leaving the slurry bed Fischer-Tropsch reactors. This significantly improved the predictions for the alternate ZSM-5 upgrading design. Another model was developed for the Wyoming coal case using ZSM-5 upgrading of the Fischer-Tropsch reactor vapors. To date, this is the best indirect coal liquefaction case. Sensitivity studies showed that additional cost reductions are possible.« less

Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

NASA Technical Reports Server (NTRS)

Weinstein, H.; Lavan, Z.

1975-01-01

Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

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

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

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

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

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

Gehin, Jess C.; Powers, Jeffrey J.

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

Performance of compact fast pyrolysis reactor with Auger-type modules for the continuous liquid biofuel production

NASA Astrophysics Data System (ADS)

Nishimura, Shun; Ebitani, Kohki

2018-01-01

Development of a compact fast pyrolysis reactor constructed using Auger-type technology to afford liquid biofuel with high yield has been an interesting concept in support of local production for local consumption. To establish a widely useable module package, details of the performance of the developing compact module reactor were investigated. This study surveyed the properties of as-produced pyrolysis oil as a function of operation time, and clarified the recent performance of the developing compact fast pyrolysis reactor. Results show that after condensation in the scrubber collector, e.g. approx. 10 h for a 25 kg/h feedstock rate, static performance of pyrolysis oil with approximately 20 MJ/kg (4.8 kcal/g) calorific values were constantly obtained after an additional 14 h. The feeding speed of cedar chips strongly influenced the time for oil condensation process: i.e. 1.6 times higher feeding speed decreased the condensation period by half (approx. 5 h in the case of 40 kg/h). Increasing the reactor throughput capacity is an important goal for the next stage in the development of a compact fast pyrolysis reactor with Auger-type modules.

Update on ORNL TRANSFORM Tool: Simulating Multi-Module Advanced Reactor with End-to-End I&C

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

Hale, Richard Edward; Fugate, David L.; Cetiner, Sacit M.

2015-05-01

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the fourth year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled reactor) concepts, including the use of multiple coupled reactors at a single site. The focus of this report is the development of a steam generator and drum system model that includes the complex dynamics of typical steam drum systems, the development of instrumentation and controls for the steam generator with drum system model, and the development of multi-reactor module models that reflect the full power reactormore » innovative small module design concept. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor models; ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface technical area; and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the TRANSFORM tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the Advanced Reactors Technology program; (2) developing a library of baseline component modules that can be assembled into full plant models using available geometry, design, and thermal-hydraulic data; (3) defining modeling conventions for interconnecting component models; and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.« less

Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

NASA Astrophysics Data System (ADS)

Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

2018-04-01

1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

MYRRHA: A multipurpose nuclear research facility

NASA Astrophysics Data System (ADS)

Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

2014-12-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

Performance of Self-developing Radiography Films in LVR-15's Neutron Beams

NASA Astrophysics Data System (ADS)

Soltes, Jaroslav; Viererbl, Ladislav; Klupak, Vit; Vins, Miroslav; Michalcova, Bozena

In the search for a suitable detector for demonstration neutron radiography measurements on the zero-power VR-1 training reactor at the Czech Technical University in Prague, some options were considered. Due to the reactor's low power and spatial limitations, an easy and practical solution had to be found. Self-developing films represent a flexible detection tool in x-ray imaging. Therefore, the goal of this study was to evaluate their potential for neutron detection. For this purpose, bare and converter covered films were studied in the thermal and epithermal neutron beams at the LVR-15 research reactor in Rez, Czech Republic.

Progress and challenges of nuclear science development in Vietnam - an outlook on the occassion of the 10-th anniversary of the Dalat Nuclear Research Reactor

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

Hien, P.D.

1994-12-31

Over ten years since the commissioning of the Dalat nuclear research reactor a number of nuclear techniques have been developed and applied in Vietnam Manufacturing of radioisotopes and nuclear instruments, development of isotope tracer and nuclear analytical techniques for environmental studies, exploitation of filtered neutron beams, ... have been major activities of reactor utilizations. Efforts made during ten years of reactor operation have resulted also in establishing and sustaining the applications of nuclear techniques in medicine, industry, agriculture, etc. The successes achieved and lessons teamed over the past ten years are discussed illustrating the approaches taken for developing the nuclearmore » science in the conditions of a country having a very low national income and experiencing a transition from a centrally planned to a market-oriented economic system.« less

Metabolic modeling of synthesis gas fermentation in bubble column reactors.

PubMed

Chen, Jin; Gomez, Jose A; Höffner, Kai; Barton, Paul I; Henson, Michael A

2015-01-01

A promising route to renewable liquid fuels and chemicals is the fermentation of synthesis gas (syngas) streams to synthesize desired products such as ethanol and 2,3-butanediol. While commercial development of syngas fermentation technology is underway, an unmet need is the development of integrated metabolic and transport models for industrially relevant syngas bubble column reactors. We developed and evaluated a spatiotemporal metabolic model for bubble column reactors with the syngas fermenting bacterium Clostridium ljungdahlii as the microbial catalyst. Our modeling approach involved combining a genome-scale reconstruction of C. ljungdahlii metabolism with multiphase transport equations that govern convective and dispersive processes within the spatially varying column. The reactor model was spatially discretized to yield a large set of ordinary differential equations (ODEs) in time with embedded linear programs (LPs) and solved using the MATLAB based code DFBAlab. Simulations were performed to analyze the effects of important process and cellular parameters on key measures of reactor performance including ethanol titer, ethanol-to-acetate ratio, and CO and H2 conversions. Our computational study demonstrated that mathematical modeling provides a complementary tool to experimentation for understanding, predicting, and optimizing syngas fermentation reactors. These model predictions could guide future cellular and process engineering efforts aimed at alleviating bottlenecks to biochemical production in syngas bubble column reactors.

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

Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

NASA Technical Reports Server (NTRS)

Hall, Michael L.; Doster, Joseph M.

1986-01-01

Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

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

Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.

2014-08-21

In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and representmore » the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.« less

Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

NASA Astrophysics Data System (ADS)

Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

2014-08-01

In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

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.

Skyshine study for next generation of fusion devices

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

Gohar, Y.; Yang, S.

1987-02-01

A shielding analysis for next generation of fusion devices (ETR/INTOR) was performed to study the dose equivalent outside the reactor building during operation including the contribution from neutrons and photons scattered back by collisions with air nuclei (skyshine component). Two different three-dimensional geometrical models for a tokamak fusion reactor based on INTOR design parameters were developed for this study. In the first geometrical model, the reactor geometry and the spatial distribution of the deuterium-tritium neutron source were simplified for a parametric survey. The second geometrical model employed an explicit representation of the toroidal geometry of the reactor chamber and themore » spatial distribution of the neutron source. The MCNP general Monte Carlo code for neutron and photon transport was used to perform all the calculations. The energy distribution of the neutron source was used explicitly in the calculations with ENDF/B-V data. The dose equivalent results were analyzed as a function of the concrete roof thickness of the reactor building and the location outside the reactor building.« less

A systematic reactor design approach for the synthesis of active pharmaceutical ingredients.

PubMed

Emenike, Victor N; Schenkendorf, René; Krewer, Ulrike

2018-05-01

Today's highly competitive pharmaceutical industry is in dire need of an accelerated transition from the drug development phase to the drug production phase. At the heart of this transition are chemical reactors that facilitate the synthesis of active pharmaceutical ingredients (APIs) and whose design can affect subsequent processing steps. Inspired by this challenge, we present a model-based approach for systematic reactor design. The proposed concept is based on the elementary process functions (EPF) methodology to select an optimal reactor configuration from existing state-of-the-art reactor types or can possibly lead to the design of novel reactors. As a conceptual study, this work summarizes the essential steps in adapting the EPF approach to optimal reactor design problems in the field of API syntheses. Practically, the nucleophilic aromatic substitution of 2,4-difluoronitrobenzene was analyzed as a case study of pharmaceutical relevance. Here, a small-scale tubular coil reactor with controlled heating was identified as the optimal set-up reducing the residence time by 33% in comparison to literature values. Copyright © 2017 Elsevier B.V. All rights reserved.

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

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.

Research reactor loading pattern optimization using estimation of distribution algorithms

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

Jiang, S.; Ziver, K.; AMCG Group, RM Consultants, Abingdon

2006-07-01

A new evolutionary search based approach for solving the nuclear reactor loading pattern optimization problems is presented based on the Estimation of Distribution Algorithms. The optimization technique developed is then applied to the maximization of the effective multiplication factor (K{sub eff}) of the Imperial College CONSORT research reactor (the last remaining civilian research reactor in the United Kingdom). A new elitism-guided searching strategy has been developed and applied to improve the local convergence together with some problem-dependent information based on the 'stand-alone K{sub eff} with fuel coupling calculations. A comparison study between the EDAs and a Genetic Algorithm with Heuristicmore » Tie Breaking Crossover operator has shown that the new algorithm is efficient and robust. (authors)« less

X-ray digital industrial radiography (DIR) for local liquid velocity (VLL) measurement in trickle bed reactors (TBRs): Validation of the technique

NASA Astrophysics Data System (ADS)

Mohd Salleh, Khairul Anuar; Rahman, Mohd Fitri Abdul; Lee, Hyoung Koo; Al Dahhan, Muthanna H.

2014-06-01

Local liquid velocity measurements in Trickle Bed Reactors (TBRs) are one of the essential components in its hydrodynamic studies. These measurements are used to effectively determine a reactor's operating condition. This study was conducted to validate a newly developed technique that combines Digital Industrial Radiography (DIR) with Particle Tracking Velocimetry (PTV) to measure the Local Liquid Velocity (VLL) inside TBRs. Three millimeter-sized Expanded Polystyrene (EPS) beads were used as packing material. Three validation procedures were designed to test the newly developed technique. All procedures and statistical approaches provided strong evidence that the technique can be used to measure the VLL within TBRs.

X-ray digital industrial radiography (DIR) for local liquid velocity (V(LL)) measurement in trickle bed reactors (TBRs): validation of the technique.

PubMed

Mohd Salleh, Khairul Anuar; Rahman, Mohd Fitri Abdul; Lee, Hyoung Koo; Al Dahhan, Muthanna H

2014-06-01

Local liquid velocity measurements in Trickle Bed Reactors (TBRs) are one of the essential components in its hydrodynamic studies. These measurements are used to effectively determine a reactor's operating condition. This study was conducted to validate a newly developed technique that combines Digital Industrial Radiography (DIR) with Particle Tracking Velocimetry (PTV) to measure the Local Liquid Velocity (V(LL)) inside TBRs. Three millimeter-sized Expanded Polystyrene (EPS) beads were used as packing material. Three validation procedures were designed to test the newly developed technique. All procedures and statistical approaches provided strong evidence that the technique can be used to measure the V(LL) within TBRs.

Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

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

Kim, T. K.; Grandy, C.; Natesan, K.

The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treatedmore » separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s, and the remaining critical paths and R&D needs are generally related to the completion of qualification of fuel and structural materials, validation of reactor design codes and methods, and support of the licensing frameworks. The LFR’s technology is instead less-mature compared to the SFR’s, and will be at the engineering demonstration stage by the early 2030s. Key LFR technology development activities will focus on resolving remaining design challenges and demonstrating the viability of systems and components in the integral system, which will be done in parallel with addressing the gaps shared with SFR technology. The approach and timeline presented here assume that, for the first module demonstration, vendors would pursue a two-step licensing process based on 10CFR Part 50.« less

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

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

D-He-3 spherical torus fusion reactor system study

NASA Astrophysics Data System (ADS)

Macon, William A., Jr.

1992-04-01

This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

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.

Treatment of low strength industrial cluster wastewater by anaerobic hybrid reactor.

PubMed

Kumar, Amit; Yadav, Asheesh Kumar; Sreekrishnan, T R; Satya, Santosh; Kaushik, C P

2008-05-01

The study was aimed at treating the complex, combined wastewater generated in Mangolpuri industrial cluster. It was considered as a low strength wastewater with respect to its organic content. Anaerobic treatment of this wastewater was studied using an anaerobic hybrid reactor (AHR) which combined the best features of both the upflow anaerobic sludge blanket (UASB) reactor and anaerobic fluidized bed rector (AFBR). The performance of the reactor under different organic and hydraulic loading rates were studied. The COD removal reached 94% at an organic loading rate (OLR) of 2.08 kg COD m(-3)d(-1) at an hydraulic retention time (HRT) of 6.0 h. The granules developed were characterized in terms of their diameter and terminal settling velocity.

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)

Employing ISRU Models to Improve Hardware Design

NASA Technical Reports Server (NTRS)

Linne, Diane L.

2010-01-01

An analytical model for hydrogen reduction of regolith was used to investigate the effects of several key variables on the energy and mass performance of reactors for a lunar in-situ resource utilization oxygen production plant. Reactor geometry, reaction time, number of reactors, heat recuperation, heat loss, and operating pressure were all studied to guide hardware designers who are developing future prototype reactors. The effects of heat recuperation where the incoming regolith is pre-heated by the hot spent regolith before transfer was also investigated for the first time. In general, longer reaction times per batch provide a lower overall energy, but also result in larger and heavier reactors. Three reactors with long heat-up times results in similar energy requirements as a two-reactor system with all other parameters the same. Three reactors with heat recuperation results in energy reductions of 20 to 40 percent compared to a three-reactor system with no heat recuperation. Increasing operating pressure can provide similar energy reductions as heat recuperation for the same reaction times.

Mars power system concept definition study. Volume 1: Study results

NASA Technical Reports Server (NTRS)

Littman, Franklin D.

1994-01-01

A preliminary top level study was completed to define power system concepts applicable to Mars surface applications. This effort included definition of power system requirements and selection of power systems with the potential for high commonality. These power systems included dynamic isotope, Proton Exchange Membrane (PEM) regenerative fuel cell, sodium sulfur battery, photovoltaic, and reactor concepts. Design influencing factors were identified. Characterization studies were then done for each concept to determine system performance, size/volume, and mass. Operations studies were done to determine emplacement/deployment maintenance/servicing, and startup/shutdown requirements. Technology development roadmaps were written for each candidate power system (included in Volume 2). Example power system architectures were defined and compared on a mass basis. The dynamic isotope power system and nuclear reactor power system architectures had significantly lower total masses than the photovoltaic system architectures. Integrated development and deployment time phasing plans were completed for an example DIPS and reactor architecture option to determine the development strategies required to meet the mission scenario requirements.

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

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

Woods, Brian; Gutowska, Izabela; Chiger, Howard

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion tomore » ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.« less

Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

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

S. Bragg-Sitton; J. Bess; J. Werner

2011-09-01

Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al.,more » 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).« less

COST FUNCTION STUDIES FOR POWER REACTORS

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

Heestand, J.; Wos, L.T.

1961-11-01

A function to evaluate the cost of electricity produced by a nuclear power reactor was developed. The basic equation, revenue = capital charges + profit + operating expenses, was expanded in terms of various cost parameters to enable analysis of multiregion nuclear reactors with uranium and/or plutonium for fuel. A corresponding IBM 704 computer program, which will compute either the price of electricity or the value of plutonium, is presented in detail. (auth)

Multiphysics Object-Oriented Simulation Environment (MOOSE)

ScienceCinema

None

2017-12-09

Nuclear reactor operators can expand safety margins with more precise information about how materials behave inside operating reactors. INL's new simulation platform makes such studies easier & more informative by letting researchers "plug-n-play" their mathematical models, skipping years of computer code development.

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.

Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

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

Moe, Wayne L.

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.« less

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

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

Moe, Wayne Leland

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.« less

Advanced Instrumentation for Transient Reactor Testing

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

Corradini, Michael L.; Anderson, Mark; Imel, George

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and designmore » increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).« less

Waste tyre pyrolysis: modelling of a moving bed reactor.

PubMed

Aylón, E; Fernández-Colino, A; Murillo, R; Grasa, G; Navarro, M V; García, T; Mastral, A M

2010-12-01

This paper describes the development of a new model for waste tyre pyrolysis in a moving bed reactor. This model comprises three different sub-models: a kinetic sub-model that predicts solid conversion in terms of reaction time and temperature, a heat transfer sub-model that calculates the temperature profile inside the particle and the energy flux from the surroundings to the tyre particles and, finally, a hydrodynamic model that predicts the solid flow pattern inside the reactor. These three sub-models have been integrated in order to develop a comprehensive reactor model. Experimental results were obtained in a continuous moving bed reactor and used to validate model predictions, with good approximation achieved between the experimental and simulated results. In addition, a parametric study of the model was carried out, which showed that tyre particle heating is clearly faster than average particle residence time inside the reactor. Therefore, this fast particle heating together with fast reaction kinetics enables total solid conversion to be achieved in this system in accordance with the predictive model. Copyright © 2010 Elsevier Ltd. All rights reserved.

A Theoretical Investigation of Oxidation Efficiency of a Volatile Removal Assembly Reactor Under Microgravity Conditions

NASA Technical Reports Server (NTRS)

Guo, Boyun

2005-01-01

Volatile Removal Assembly (VRA) is a subsystem of the Closed Environment Life Support System (CELSS) installed in the International Space Station. It is used for removing contaminants (volatile organics) in the wastewater produced by the space station crews. The major contaminants are formic acid, ethanol, and propylene glycol. The VRA contains a slim packbed reactor (3.5 cm diameter and four 28 cm long tubes in series) to perform catalyst oxidation of wastewater at elevated pressure and temperature under microgravity conditions. In the reactor, the contaminants are burned with oxygen gas (O2) to form water and carbon dioxide (CO2) that dissolves in the water stream. Optimal design of the reactor requires a thorough understanding about how the reactor performs under microgravity conditions. The objective of this study was to develop a mathematical model to interpret experimental data obtained from normal and microgravity conditions, and to predict the performance of VRA reactor under microgravity conditions. Catalyst oxidation kinetics and the total oxygen-water contact area control the efficiency of catalyst oxidation for mass transfer, which depends on oxygen gas holdup and distribution in the reactor. The process involves bubbly flow in porous media with chemical reactions in microgravity environment. This presents a unique problem in fluid dynamics that has not been studied. Guo et al. (2004) developed a mathematical model that predicts oxygen holdup in the VRA reactor. No mathematical model has been found in the literature that can be used to predict the efficiency of catalyst oxidation under microgravity conditions.

Microbial dynamics in anaerobic digestion reactors for treating organic urban residues during the start-up process.

PubMed

Alcántara-Hernández, R J; Taş, N; Carlos-Pinedo, S; Durán-Moreno, A; Falcón, L I

2017-06-01

Anaerobic digestion of organic residues offers economic benefits via biogas production, still methane (CH 4 ) yield relies on the development of a robust microbial consortia for adequate substrate degradation, among other factors. In this study, we monitor biogas production and changes in the microbial community composition in two semi-continuous stirred tank reactors during the setting process under mesophilic conditions (35°C) using a 16S rDNA high-throughput sequencing method. Reactors were initially inoculated with anaerobic granular sludge from a brewery wastewater treatment plant, and gradually fed organic urban residues (4·0 kg VS m -3  day -1 ) . The inocula and biomass samples showed changes related to adaptations of the community to urban organic wastes including a higher relative proportion of Clostridiales, with Ruminococcus spp. and Syntrophomonas spp. as recurrent species. Candidatus Cloacamonas spp. (Spirochaetes) also increased from ~2·2% in the inoculum to >10% in the reactor biomass. The new community consolidated the cellulose degradation and the propionate and amino acids fermentation processes. Acetoclastic methanogens were more abundant in the reactor, where Methanosaeta spp. was found as a key player. This study demonstrates a successful use of brewery treatment plant granular sludge to obtain a robust consortium for methane production from urban organic solid waste in Mexico. This study describes the selection of relevant bacteria and archaea in anaerobic digesters inoculated with anaerobic granular sludge from a brewery wastewater treatment plant. Generally, these sludge granules are used to inoculate reactors digesting organic urban wastes. Though, it is still not clearly understood how micro-organisms respond to substrate variations during the reactor start-up process. After feeding two reactors with organic urban residues, it was found that a broader potential for cellulose degradation was developed including Bacteroidetes, Firmicutes and Spirochaetes. These results clarify the bacterial processes behind new reactors establishment for treating organic wastes in urban areas. © 2017 The Society for Applied Microbiology.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

breeder reactors should become...the primary type of reactors . In developing breeder reactors , we should follow the path of using metal fuel. Breeder reactors give us more time to...first reactor used for power generation was a fast reactor : the " Breeder 1" reactor at the Idaho National Reactor Test Center which was used to

Analysis of decommissioning costs for the AFRRI TRIGA reactor facility. Technical report

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

Forsbacka, M.; Moore, M.

1989-12-01

This report provides a cost analysis for decommissioning the Armed Forces Radiobiology Research Institute (AFRRI) TRIGA reactor facility. AFRRI is not suggesting that the AFRRI TRIGA reactor facility be decommissioned. This report was prepared in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations, which requires that funding for the decommissioning of reactor facilities be available when licensed activities cease. The planned method of decommissioning is complete decontamination (DECON) of the AFRRI TRIGA reactor site to allow for restoration of the site to full public access. The cost of DECON in 1990 dollars is estimated to be $3,200,000.more » The anticipated ancillary costs of facility site demobilization and spent fuel shipment will be an additional $600,000. Thus, the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for developing this cost estimate was a study of the decommissioning costs of similar reactor facility performed by Battelle Pacific Northwest Laboratory, as provided in U.S. Nuclear Regulatory Commission publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA reactor facility.« less

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.

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

Development of the Technology of Vortex Diagnostics to Improve the Safety of Operation of Nuclear Reactors

NASA Astrophysics Data System (ADS)

Mitrofanova, O. V.; Ivlev, O. A.; Pozdeeva, I. G.; Urtenov, D. S.

2017-11-01

The results of studies are aimed at developing theoretical foundations and instrumentation system to ensure a technology of vortex diagnostics of the state of flows of fluids for nuclear power installations with power water reactors and fast neutrons reactors with liquid-metal coolants. The technology of vortex diagnostics is based on the study of acoustic, magneto-hydrodynamic and resonant effects related to the formation of stable vortex structures. For creation a system of monitoring and diagnostics of the crisis phenomena due to hydrodynamics of the flow, it is proposed to use acoustic method to record the radiation of elastic waves in the fluids caused by the dynamic local rearrangement of its structure.

Study of carbon dioxide gas treatment based on equations of kinetics in plasma discharge reactor

NASA Astrophysics Data System (ADS)

Abedi-Varaki, Mehdi

2017-08-01

Carbon dioxide (CO2) as the primary greenhouse gas, is the main pollutant that is warming earth. CO2 is widely emitted through the cars, planes, power plants and other human activities that involve the burning of fossil fuels (coal, natural gas and oil). Thus, there is a need to develop some method to reduce CO2 emission. To this end, this study investigates the behavior of CO2 in dielectric barrier discharge (DBD) plasma reactor. The behavior of different species and their reaction rates are studied using a zero-dimensional model based on equations of kinetics inside plasma reactor. The results show that the plasma reactor has an effective reduction on the CO2 density inside the reactor. As a result of reduction in the temporal variations of reaction rate, the speed of chemical reactions for CO2 decreases and very low concentration of CO2 molecules inside the plasma reactor is generated. The obtained results are compared with the existing experimental and simulation findings in the literature.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

NASA Astrophysics Data System (ADS)

Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

2015-09-01

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF2-ThF4-233UF4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

THE ARMOUR DUST FUELED REACTOR (ADFR)

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

Krucoff, D.

1958-01-01

The A-DFR is based on the use of a fissionable dust carried in a gas. This fuel ferm offers promise of a major economic advance through the use of 2,000 to 3,000 F operating temperatures and a low cost fuel cycle. The development program is described that was initiated to investigate experimentally the proposed fuel and study analytically other reactor characteristics. A brief review of the reactor concept is presented. (W.D.M.)

Multiphase organic synthesis in microchannel reactors.

PubMed

Kobayashi, Juta; Mori, Yuichiro; Kobayashi, Shū

2006-07-17

"Miniaturization" is one of the most important aspects in today's technology. Organic chemistry is no exception. The search for highly effective, controllable, environmentally friendly methods for preparing products is of prime importance. The development of multiphase organic reactions in microchannel reactors has gained significant importance in recent years to allow novel reactivity, and has led to many fruitful results that are not attainable in conventional reactors. This Focus Review aims to shed light on how effectively multiphase organic reactions can be conducted with microchannel reactors by providing examples of recent remarkable studies, which have been grouped on the basis of the phases involved.

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos supported the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for active and passive reactor scrams using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following active-system scrams.

Determination of parameters of a nuclear reactor through noise measurements

DOEpatents

Cohn, C.E.

1975-07-15

A method of measuring parameters of a nuclear reactor by noise measurements is described. Noise signals are developed by the detectors placed in the reactor core. The polarity coincidence between the noise signals is used to develop quantities from which various parameters of the reactor can be calculated. (auth)

The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

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

Mann, A.; Herrick, R.; Gunn, J.

2007-07-01

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal ofmore » the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)« less

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

PubMed

Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

2016-04-01

A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system. Copyright © 2016 Elsevier Ltd. All rights reserved.

The RERTR Program : a status report.

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

Travelli, A.

1998-10-19

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners since its inception in 1978. A brief summary of the results that the program had attained by the end of 1997 is followed by a detailed review of the major events, findings, and activities that took place in 1998. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. Four additional shipments of spent fuel from foreign research reactors were accepted by the U.S. Altogether, 2,231 spent fuel assemblies from foreignmore » research reactors have been received by the U.S. under the acceptance policy. Fuel development activities began to yield solid results. Irradiations of the first two batches of microplates were completed. Preliminary postirradiation examinations of these microplates indicate excellent irradiation behavior of some of the fuel materials that were tested. These materials hold the promise of achieving the pro am goal of developing LEU research reactor fuels with uranium density in the 8-9 g /cm{sup 3} range. Progress was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian-supplied research reactors to LEU fuels. Feasibility studies for converting to LEU fuel four Russian-designed research reactors (IR-8 in Russia, Budapest research reactor in Hungary, MARIA in Poland, and WWR-SM in Uzbekistan) were completed. A new program activity began to study the feasibility of converting three Russian plutonium production reactors to the use of low-enriched U0{sub 2}-Al dispersion fuel, so that they can continue to produce heat and electricity without producing significant amounts of plutonium. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, the transient performance of the core under hypothetical accident conditions. A major milestone was accomplished in the development of a process to produce molybdenum-99 from fission targets utilizing LEU instead of HEU. Targets containing LEU metal foils were irradiated in the RAS-GAS reactor at BATAN, Indonesia, and molybdenum-99 was successfully extracted through the ensuing process. These are exciting times for the program and for all those involved in it, and last year's successes augur well for the future. However, as in the past, the success of the RERTR program will depend on the international friendship and cooperation that have always been its trademark.« less

Economics and Environmental Compatibility of Fusion Reactors —Its Analysis and Coming Issues— 4.Economic Effect of Fusion in Energy Market 4.2Various Externalities of Energy Systems and the Integrated Evaluation

NASA Astrophysics Data System (ADS)

Ito, Keishiro

The primacy of a nuclear fusion reactor in a competitive energy market remarkably depends on to what extent the reactor contributes to reduce the externalities of energy. The reduction effects are classified into two effects, which have quite dissimilar characteristics. One is an effect of environmental dimensions. The other is related to energy security. In this study I took up the results of EC's Extern Eproject studies as are presentative example of the former effect. Concerning the latter effect, I clarified the fundamental characteristics of externalities related to energy security and the conceptual framework for the purpose of evaluation. In the socio-economical evaluation of research into and development investments in nuclear fusions reactors, the public will require the development of integrated evaluation systems to support the cost-effect analysis of how well the reduction effects of externalities have been integrated with the effects of technological innovation, learning, spillover, etc.

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

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

Cao, Yan; Gohar, Yousry

2015-11-01

In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate themore » dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.« less

Anaerobic sequencing batch reactor in pilot scale for treatment of tofu industry wastewater

NASA Astrophysics Data System (ADS)

Rahayu, Suparni Setyowati; Purwanto, Budiyono

2015-12-01

The small industry of tofu production process releases the waste water without being processed first, and the wastewater is directly discharged into water. In this study, Anaerobic Sequencing Batch Reactor in Pilot Scale for Treatment of Tofu Industry was developed through an anaerobic process to produce biogas as one kind of environmentally friendly renewable energy which can be developed into the countryside. The purpose of this study was to examine the fundamental characteristics of organic matter elimination of industrial wastewater with small tofu effective method and utilize anaerobic active sludge with Anaerobic Sequencing Bath Reactor (ASBR) to get rural biogas as an energy source. The first factor is the amount of the active sludge concentration which functions as the decomposers of organic matter and controlling selectivity allowance to degrade organic matter. The second factor is that HRT is the average period required substrate to react with the bacteria in the Anaerobic Sequencing Bath Reactor (ASBR).The results of processing the waste of tofu production industry using ASBR reactor with active sludge additions as starter generates cumulative volume of 5814.4 mL at HRT 5 days so that in this study it is obtained the conversion 0.16 L of CH4/g COD and produce biogas containing of CH4: 81.23% and CO2: 16.12%. The wastewater treatment of tofu production using ASBR reactor is able to produce renewable energy that has economic value as well as environmentally friendly by nature.

Enrichment of acetogenic bacteria in high rate anaerobic reactors under mesophilic and thermophilic conditions.

PubMed

Ryan, P; Forbes, C; McHugh, S; O'Reilly, C; Fleming, G T A; Colleran, E

2010-07-01

The objective of the current study was to expand the knowledge of the role of acetogenic Bacteria in high rate anaerobic digesters. To this end, acetogens were enriched by supplying a variety of acetogenic growth supportive substrates to two laboratory scale high rate upflow anaerobic sludge bed (UASB) reactors operated at 37 degrees C (R1) and 55 degrees C (R2). The reactors were initially fed a glucose/acetate influent. Having achieved high operational performance and granular sludge development and activity, both reactors were changed to homoacetogenic bacterial substrates on day 373 of the trial. The reactors were initially fed with sodium vanillate as a sole substrate. Although % COD removal indicated that the 55 degrees C reactor out performed the 37 degrees C reactor, effluent acetate levels from R2 were generally higher than from R1, reaching values as high as 5023 mg l(-1). Homoacetogenic activity in both reactors was confirmed on day 419 by specific acetogenic activity (SAA) measurement, with higher values obtained for R2 than R1. Sodium formate was introduced as sole substrate to both reactors on day 464. It was found that formate supported acetogenic activity at both temperatures. By the end of the trial, no specific methanogenic activity (SMA) was observed against acetate and propionate indicating that the methane produced was solely by hydrogenotrophic Archaea. Higher SMA and SAA values against H(2)/CO(2) suggested development of a formate utilising acetogenic population growing in syntrophy with hydrogenotrophic methanogens. Throughout the formate trial, the mesophilic reactor performed better overall than the thermophilic reactor. Copyright 2010 Elsevier Ltd. All rights reserved.

Pre-Combustion Carbon Dioxide Capture by a New Dual Phase Ceramic-Carbonate Membrane Reactor

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

Lin, Jerry Y. S.

2015-01-31

This report documents synthesis, characterization and carbon dioxide permeation and separation properties of a new group of ceramic-carbonate dual-phase membranes and results of a laboratory study on their application for water gas shift reaction with carbon dioxide separation. A series of ceramic-carbonate dual phase membranes with various oxygen ionic or mixed ionic and electronic conducting metal oxide materials in disk, tube, symmetric, and asymmetric geometric configurations was developed. These membranes, with the thickness of 10 μm to 1.5 mm, show CO 2 permeance in the range of 0.5-5×10 -7 mol·m -2·s -1·Pa -1 in 500-900°C and measured CO 2/N 2more » selectivity of up to 3000. CO 2 permeation mechanism and factors that affect CO 2 permeation through the dual-phase membranes have been identified. A reliable CO 2 permeation model was developed. A robust method was established for the optimization of the microstructures of ceramic-carbonate membranes. The ceramic-carbonate membranes exhibit high stability for high temperature CO 2 separations and water gas shift reaction. Water gas shift reaction in the dual-phase membrane reactors was studied by both modeling and experiments. It is found that high temperature syngas water gas shift reaction in tubular ceramic-carbonate dual phase membrane reactor is feasible even without catalyst. The membrane reactor exhibits good CO 2 permeation flux, high thermal and chemical stability and high thermal shock resistance. Reaction and separation conditions in the membrane reactor to produce hydrogen of 93% purity and CO 2 stream of >95% purity, with 90% CO 2 capture have been identified. Integration of the ceramic-carbonate dual-phase membrane reactor with IGCC process for carbon dioxide capture was analyzed. A methodology was developed to identify optimum operation conditions for a membrane tube of given dimensions that would treat coal syngas with targeted performance. The calculation results show that the dual-phase membrane reactor could improve IGCC process efficiency but the cost of the membrane reactor with membranes having current CO 2 permeance is high. Further research should be directed towards improving the performance of the membranes and developing cost-effective, scalable methods for fabrication of dual-phase membranes and membrane reactors.« less

The Simulator Development for RDE Reactor

NASA Astrophysics Data System (ADS)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Catalyst Residence Time Distributions in Riser Reactors for Catalytic Fast Pyrolysis. Part 2: Pilot-Scale Simulations and Operational Parameter Study

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

Foust, Thomas D.; Ziegler, Jack L.; Pannala, Sreekanth

2017-02-21

Here, wsing the validated simulation model developed in part one of this study for biomass catalytic fast pyrolysis (CFP), we assess the functional utility of using this validated model to assist in the development of CFP processes in fluidized catalytic cracking (FCC) reactors to a commercially viable state. Specifically, we examine the effects of mass flow rates, boundary conditions (BCs), pyrolysis vapor molecular weight variation, and the impact of the chemical cracking kinetics on the catalyst residence times. The factors that had the largest impact on the catalyst residence time included the feed stock molecular weight and the degree ofmore » chemical cracking as controlled by the catalyst activity. Lastly, because FCC reactors have primarily been developed and utilized for petroleum cracking, we perform a comparison analysis of CFP with petroleum and show the operating regimes are fundamentally different.« less

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

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

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-01-14

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-05-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

NASA Technical Reports Server (NTRS)

Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

2012-01-01

Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

Evaluation of power density on the bioethanol production using mesoscale oscillatory baffled reactor and stirred tank reactor

NASA Astrophysics Data System (ADS)

Yussof, H. W.; Bahri, S. S.; Mazlan, N. A.

2018-03-01

A recent development in oscillatory baffled reactor technology is down-scaling the reactor, so that it can be used for production of small-scale bioproduct. In the present study, a mesoscale oscillatory baffled reactor (MOBR) with central baffle system was developed. The reactor performance of the MOBR was compared with conventional stirred tank reactor (STR) to evaluate the performance of bioethanol fermentation using Saccharomyces cerevisiae. Evaluation was made at similar power density of 24.21, 57.38, 112.35 and 193.67 Wm-3 by varying frequency (f), amplitude (xo) and agitation speed (rpm). It was found that the MOBR improved the mixing intensity resulted in lower glucose concentration (0.988 gL-1) and higher bioethanol concentration (38.98 gL-1) after 12 hours fermentation at power density of 193.67 Wm-3. Based on the results, the bioethanol yield obtained using MOBR was 39% higher than the maximum achieved in STR. Bioethanol production using MOBR proved to be feasible as it is not only able to compete with conventional STR but also offers advantages of straight-forward scale-up, whereas it is complicated and difficult in STR. Overall, MOBR offers great prospective over the conventional STR.

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

None

>Fundamental Alloying. Studies of crystal structures, reactions at metal surfaces, spectroscopy of molten salts, mechanical deformation, and alloy theory are reported. Long-Range Applied Metallurgy. A thermal comparator is described and the characteristic temperature of U0/sub 2/ determined. Sintering studies were carried out on ThO/sub 2/. The diffusion of fission products in fuel and of Al/sup 26/ and Mn/sup 54/ in Al and the reaction of Be with UC were studied. Transformation and oxidation data were obtained for a number of Zr alloys. Reactor Metallurgy. A large number of ceramic technology projects are described. Some corrosion data are given for metalsmore » exposed to impure He and molten fluorides. Studies were made of the fission-gas-retention Properties of ceramic fuel bodies. A large number of materials compatibility studies are described. The mechanical properties of some reactor materials were studied. Fabrication work was conducted to develop materials for application in low-, medium-, and high-temperature reactors or systems. A large number of new metallographic and nondestructive testing techniques are reported. Studies were carried out on the oxidation, carburization, and stability of alloys. Equipment for postirradiation examination is described. Preparation of some alloys and dispersion fuels by powder metallurgy methods was studied. The development of welding and brazing techniques for reactor materials is described. (D.L.C.)« less

A Practical Approach to Starting Fission Surface Power Development

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

The Prometheus Power and Propulsion Program has been reformulated to address NASA needs relative to lunar and Mars exploration. Emphasis has switched from the Jupiter Icy Moons Orbiter (JIMO) flight system development to more generalized technology development addressing Fission Surface Power (FSP) and Nuclear Thermal Propulsion (NTP). Current NASA budget priorities and the deferred mission need date for nuclear systems prohibit a fully funded reactor Flight Development Program. However, a modestly funded Advanced Technology Program can and should be conducted to reduce the risk and cost of future flight systems. A potential roadmap for FSP technology development leading to possible flight applications could include three elements: 1) Conceptual Design Studies, 2) Advanced Component Technology, and 3) Non-Nuclear System Testing. The Conceptual Design Studies would expand on recent NASA and DOE analyses while increasing the depth of study in areas of greatest uncertainty such as reactor integration and human-rated shielding. The Advanced Component Technology element would address the major technology risks through development and testing of reactor fuels, structural materials, primary loop components, shielding, power conversion, heat rejection, and power management and distribution (PMAD). The Non-Nuclear System Testing would provide a modular, technology testbed to investigate and resolve system integration issues.

Potential of Electric Power Production from Microbial Fuel Cell (MFC) in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

NASA Astrophysics Data System (ADS)

Zaman, Badrus; Wardhana, Irawan Wisnu

2018-02-01

Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media). Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day) operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

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

Pan, Paul Y

2010-12-10

An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

Minimizing mixing intensity to improve the performance of rice straw anaerobic digestion via enhanced development of microbe-substrate aggregates.

PubMed

Kim, Moonkyung; Kim, Byung-Chul; Choi, Yongju; Nam, Kyoungphile

2017-12-01

The aim of this work was to study the effect of the differential development of microbe-substrate aggregates at different mixing intensities on the performance of anaerobic digestion of rice straw. Batch and semi-continuous reactors were operated for up to 50 and 300days, respectively, under different mixing intensities. In both batch and semi-continuous reactors, minimal mixing conditions exhibited maximum methane production and lignocellulose biodegradability, which both had strong correlations with the development of microbe-substrate aggregates. The results implied that the aggregated microorganisms on the particulate substrate played a key role in rice straw hydrolysis, determining the performance of anaerobic digestion. Increasing the mixing speed from 50 to 150rpm significantly reduced the methane production rate by disintegrating the microbe-substrate aggregates in the semi-continuous reactor. A temporary stress of high-speed mixing fundamentally affected the microbial communities, increasing the possibility of chronic reactor failure. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

Tan, Lizhen; Yang, Ying; Chen, Tianyi

Advanced nuclear reactors as well as the life extension of light water reactors require advanced alloys capable of satisfactory operation up to neutron damage levels approaching 200 displacements per atom (dpa). Extensive studies, including fundamental theories, have demonstrated the superior resistance to radiation-induced swelling in ferritic steels, primarily inherited from their body-centered cubic (bcc) structure. This study aims at developing nanoprecipitates strengthened advanced ferritic alloys for advanced nuclear reactor applications. To be more specific, this study aims at enhancing the amorphization ability of some precipitates, such as Laves phase and other types of intermetallic phases, through smart alloying strategy, andmore » thereby promote the crystalline®amorphous transformation of these precipitates under irradiation.« less

Fischer-Tropsch Slurry Reactor modeling

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

Soong, Y.; Gamwo, I.K.; Harke, F.W.

1995-12-31

This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas,more » solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.« less

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

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

Bess, J. D.; Briggs, J. B.; Gulliford, J.

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energymore » Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

Performance Assessment of the Commercial CFD Software for the Prediction of the Reactor Internal Flow

NASA Astrophysics Data System (ADS)

Lee, Gong Hee; Bang, Young Seok; Woo, Sweng Woong; Kim, Do Hyeong; Kang, Min Ku

2014-06-01

As the computer hardware technology develops the license applicants for nuclear power plant use the commercial CFD software with the aim of reducing the excessive conservatism associated with using simplified and conservative analysis tools. Even if some of CFD software developer and its user think that a state of the art CFD software can be used to solve reasonably at least the single-phase nuclear reactor problems, there is still limitation and uncertainty in the calculation result. From a regulatory perspective, Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of the commercial CFD software for nuclear reactor problems. In this study, in order to examine the validity of the results of 1/5 scaled APR+ (Advanced Power Reactor Plus) flow distribution tests and the applicability of CFD in the analysis of reactor internal flow, the simulation was conducted with the two commercial CFD software (ANSYS CFX V.14 and FLUENT V.14) among the numerous commercial CFD software and was compared with the measurement. In addition, what needs to be improved in CFD for the accurate simulation of reactor core inlet flow was discussed.

Dynamic analysis of gas-core reactor system

NASA Technical Reports Server (NTRS)

Turner, K. H., Jr.

1973-01-01

A heat transfer analysis was incorporated into a previously developed model CODYN to obtain a model of open-cycle gaseous core reactor dynamics which can predict the heat flux at the cavity wall. The resulting model was used to study the sensitivity of the model to the value of the reactivity coefficients and to determine the system response for twenty specified perturbations. In addition, the model was used to study the effectiveness of several control systems in controlling the reactor. It was concluded that control drums located in the moderator region capable of inserting reactivity quickly provided the best control.

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.).

SP-100 program developments

NASA Technical Reports Server (NTRS)

Schnyer, A. D.; Sholtis, J. A., Jr.; Wahlquist, E. J.; Verga, R. L.; Wiley, R. L.

1985-01-01

An update is provided on the status of the Sp-100 Space Reactor Power Program. The historical background that led to the program is reviewed and the overall program objectives and development approach are discussed. The results of the mission studies identify applications for which space nuclear power is desirable and even essential. Results of a series of technology feasibility experiments are expected to significantly improve the earlier technology data base for engineering development. The conclusion is reached that a nuclear reactor space power system can be developed by the early 1990s to meet emerging mission performance requirements.

Development of a Software Safety Process and a Case Study of Its Use

NASA Technical Reports Server (NTRS)

Knight, J. C.

1996-01-01

Research in the year covered by this reporting period has been primarily directed toward: continued development of mock-ups of computer screens for operator of a digital reactor control system; development of a reactor simulation to permit testing of various elements of the control system; formal specification of user interfaces; fault-tree analysis including software; evaluation of formal verification techniques; and continued development of a software documentation system. Technical results relating to this grant and the remainder of the principal investigator's research program are contained in various reports and papers.

Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

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

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

1984-06-01

ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Componentmore » Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.« less

Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

NASA Astrophysics Data System (ADS)

Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

2015-11-01

High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

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

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« less

Reactor application of an improved bundle divertor

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

Yang, T.F.; Ruck, G.W.; Lee, A.Y.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supportedmore » by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW.« less

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

Reflector and Protections in a Sodium-cooled Fast Reactor: Modelling and Optimization

NASA Astrophysics Data System (ADS)

Blanchet, David; Fontaine, Bruno

2017-09-01

The ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Generation IV nuclear reactor concept under development in France [1]. In this frame, studies are underway to optimize radial reflectors and protections. Considering radial protections made in natural boron carbide, this study is conducted to assess the neutronic performances of the MgO as the reference choice for reflector material, in comparison with other possible materials including a more conventional stainless steel. The analysis is based upon a simplified 1-D and 2-D deterministic modelling of the reactor, providing simplified interfaces between core, reflector and protections. Such models allow examining detailed reaction rate distributions; they also provide physical insights into local spectral effects occurring at the Core-Reflector and at the Reflector-Protection interfaces.

Reactor monitoring using antineutrino detectors

NASA Astrophysics Data System (ADS)

Bowden, N. S.

2011-08-01

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

Versatile in situ gas analysis apparatus for nanomaterials reactors.

PubMed

Meysami, Seyyed Shayan; Snoek, Lavina C; Grobert, Nicole

2014-09-02

We report a newly developed technique for the in situ real-time gas analysis of reactors commonly used for the production of nanomaterials, by showing case-study results obtained using a dedicated apparatus for measuring the gas composition in reactors operating at high temperature (<1000 °C). The in situ gas-cooled sampling probe mapped the chemistry inside the high-temperature reactor, while suppressing the thermal decomposition of the analytes. It thus allows a more accurate study of the mechanism of progressive thermocatalytic cracking of precursors compared to previously reported conventional residual gas analyses of the reactor exhaust gas and hence paves the way for the controlled production of novel nanomaterials with tailored properties. Our studies demonstrate that the composition of the precursors dynamically changes as they travel inside of the reactor, causing a nonuniform growth of nanomaterials. Moreover, mapping of the nanomaterials reactor using quantitative gas analysis revealed the actual contribution of thermocatalytic cracking and a quantification of individual precursor fragments. This information is particularly important for quality control of the produced nanomaterials and for the recycling of exhaust residues, ultimately leading toward a more cost-effective continuous production of nanomaterials in large quantities. Our case study of multiwall carbon nanotube synthesis was conducted using the probe in conjunction with chemical vapor deposition (CVD) techniques. Given the similarities of this particular CVD setup to other CVD reactors and high-temperature setups generally used for nanomaterials synthesis, the concept and methodology of in situ gas analysis presented here does also apply to other systems, making it a versatile and widely applicable method across a wide range of materials/manufacturing methods, catalysis, as well as reactor design and engineering.

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.

Assessing thermal conductivity of composting reactor with attention on varying thermal resistance between compost and the inner surface.

PubMed

Wang, Yongjiang; Niu, Wenjuan; Ai, Ping

2016-12-01

Dynamic estimation of heat transfer through composting reactor wall was crucial for insulating design and maintaining a sanitary temperature. A model, incorporating conductive, convective and radiative heat transfer mechanisms, was developed in this paper to provide thermal resistance calculations for composting reactor wall. The mechanism of thermal transfer from compost to inner surface of structural layer, as a first step of heat loss, was important for improving insulation performance, which was divided into conduction and convection and discussed specifically in this study. It was found decreasing conductive resistance was responsible for the drop of insulation between compost and reactor wall. Increasing compost porosity or manufacturing a curved surface, decreasing the contact area of compost and the reactor wall, might improve the insulation performance. Upon modeling of heat transfers from compost to ambient environment, the study yielded a condensed and simplified model that could be used to conduct thermal resistance analysis for composting reactor. With theoretical derivations and a case application, the model was applicable for both dynamic estimation and typical composting scenario. Copyright © 2016 Elsevier Ltd. All rights reserved.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

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

Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id; Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 datamore » library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.« less

Virtual environments simulation in research reactor

NASA Astrophysics Data System (ADS)

Muhamad, Shalina Bt. Sheik; Bahrin, Muhammad Hannan Bin

2017-01-01

Virtual reality based simulations are interactive and engaging. It has the useful potential in improving safety training. Virtual reality technology can be used to train workers who are unfamiliar with the physical layout of an area. In this study, a simulation program based on the virtual environment at research reactor was developed. The platform used for virtual simulation is 3DVia software for which it's rendering capabilities, physics for movement and collision and interactive navigation features have been taken advantage of. A real research reactor was virtually modelled and simulated with the model of avatars adopted to simulate walking. Collision detection algorithms were developed for various parts of the 3D building and avatars to restrain the avatars to certain regions of the virtual environment. A user can control the avatar to move around inside the virtual environment. Thus, this work can assist in the training of personnel, as in evaluating the radiological safety of the research reactor facility.

Neutronics Analysis of SMART Small Modular Reactor using SRAC 2006 Code

NASA Astrophysics Data System (ADS)

Ramdhani, Rahmi N.; Prastyo, Puguh A.; Waris, Abdul; Widayani; Kurniadi, Rizal

2017-07-01

Small modular reactors (SMRs) are part of a new generation of nuclear reactor being developed worldwide. One of the advantages of SMR is the flexibility to adopt the advanced design concepts and technology. SMART (System integrated Modular Advanced ReacTor) is a small sized integral type PWR with a thermal power of 330 MW that has been developed by KAERI (Korea Atomic Energy Research Institute). SMART core consists of 57 fuel assemblies which are based on the well proven 17×17 array that has been used in Korean commercial PWRs. SMART is soluble boron free, and the high initial reactivity is mainly controlled by burnable absorbers. The goal of this study is to perform neutronics evaluation of SMART core with UO2 as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2006 code with JENDL 3.3 as nuclear data library.

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

Development costs for a nuclear electric propulsion stage.

NASA Technical Reports Server (NTRS)

Mondt, J. F.; Prickett, W. Z.

1973-01-01

Development costs are presented for an unmanned nuclear electric propulsion (NEP) stage based upon a liquid metal cooled, in-core thermionic reactor. A total of 120 kWe are delivered to the thrust subsystem which employs mercury ion engines for electric propulsion. This study represents the most recent cost evaluation of the development of a reactor power system for a wide range of nuclear space power applications. These include geocentric, and outer planet and other deep space missions. The development program is described for the total NEP stage, based upon specific development programs for key NEP stage components and subsystems.

Metrics for the technical performance evaluation of light water reactor accident-tolerant fuel

DOE PAGES

Bragg-Sitton, Shannon M.; Todosow, Michael; Montgomery, Robert; ...

2017-03-26

The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident-tolerant fuel (ATF) for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, andmore » economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+), fuels with enhanced accident tolerance would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. Research and development of ATF in the United States is conducted under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Advanced Fuels Campaign. The DOE is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This study summarizes the technical evaluation methodology proposed in the United States to aid in the optimization and prioritization of candidate ATF designs.« less

Neutron flux and power in RTP core-15

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

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE PAGES

Collette, R.; King, J.; Buesch, C.; ...

2016-04-01

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

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

Collette, R.; King, J.; Buesch, C.

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Evaluation of the resilience of a full-scale down-flow hanging sponge reactor to long-term outages at a sewage treatment plant in India.

PubMed

Onodera, Takashi; Takayama, Daisuke; Ohashi, Akiyoshi; Yamaguchi, Takashi; Uemura, Shigeki; Harada, Hideki

2016-10-01

Resilience to process outages is an essential requirement for sustainable wastewater treatment systems in developing countries. In this study, we evaluated the ability of a full-scale down-flow hanging sponge (DHS) reactor to recover after a 10-day outage. The DHS tested in this study uses polyurethane sponge as packing material. This full-scale DHS reactor has been tested over a period of about 4 years in India with a flow rate of 500 m(3)/day. Water was not supplied to the DHS reactor that was subjected to the 10-day outage; however, the biomass did not dry out because the sponge was able to retain enough water. Soon after the reactor was restarted, a small quantity of biomass, amounting to only 0.1% of the total retained biomass, was eluted. The DHS effluent achieved satisfactory removal of suspended solids, chemical oxygen demand, and ammonium nitrogen within 90, 45, and 90 min, respectively. Conversely, fecal coliforms in the DHS effluent did not reach satisfactory levels within 540 min; instead, the normal levels of fecal coliforms were achieved within 3 days. Overall, the tests demonstrated that the DHS reactor was sufficiently robust to withstand long-term outages and achieved steady state soon after restart. This reinforces the suitability of this technology for developing countries. Copyright © 2016 Elsevier Ltd. All rights reserved.

Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

NASA Astrophysics Data System (ADS)

Ródenas, José

2017-11-01

All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

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.

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. .

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Astrophysics Data System (ADS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-09-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Technical Reports Server (NTRS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-01-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Design of an ammonia closed-loop storage system in a CSP power plant with a power tower cavity receiver

NASA Astrophysics Data System (ADS)

Abdiwe, Ramadan; Haider, Markus

2017-06-01

In this study the thermochemical system using ammonia as energy storage carrier is investigated and a transient mathematical model using MATLAB software was developed to predict the behavior of the ammonia closed-loop storage system including but not limited to the ammonia solar reactor and the ammonia synthesis reactor. The MATLAB model contains transient mass and energy balances as well as chemical equilibrium model for each relevant system component. For the importance of the dissociation and formation processes in the system, a Computational Fluid Dynamics (CFD) simulation on the ammonia solar and synthesis reactors has been performed. The CFD commercial package FLUENT is used for the simulation study and all the important mechanisms for packed bed reactors are taken into account, such as momentum, heat and mass transfer, and chemical reactions. The FLUENT simulation reveals the profiles inside both reactors and compared them with the profiles from the MATLAB code.

A brief history of design studies on innovative nuclear reactors

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2014-09-01

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

Two-phase anaerobic digestion of vegetable market waste fraction of municipal solid waste and development of improved technology for phase separation in two-phase reactor.

PubMed

Majhi, Bijoy Kumar; Jash, Tushar

2016-12-01

Biogas production from vegetable market waste (VMW) fraction of municipal solid waste (MSW) by two-phase anaerobic digestion system should be preferred over the single-stage reactors. This is because VMW undergoes rapid acidification leading to accumulation of volatile fatty acids and consequent low pH resulting in frequent failure of digesters. The weakest part in the two-phase anaerobic reactors was the techniques applied for solid-liquid phase separation of digestate in the first reactor where solubilization, hydrolysis and acidogenesis of solid organic waste occur. In this study, a two-phase reactor which consisted of a solid-phase reactor and a methane reactor was designed, built and operated with VMW fraction of Indian MSW. A robust type filter, which is unique in its implementation method, was developed and incorporated in the solid-phase reactor to separate the process liquid produced in the first reactor. Experiments were carried out to assess the long term performance of the two-phase reactor with respect to biogas production, volatile solids reduction, pH and number of occurrence of clogging in the filtering system or choking in the process liquid transfer line. The system performed well and was operated successfully without the occurrence of clogging or any other disruptions throughout. Biogas production of 0.86-0.889m 3 kg -1 VS, at OLR of 1.11-1.585kgm -3 d -1 , were obtained from vegetable market waste, which were higher than the results reported for similar substrates digested in two-phase reactors. The VS reduction was 82-86%. The two-phase anaerobic digestion system was demonstrated to be stable and suitable for the treatment of VMW fraction of MSW for energy generation. Copyright © 2016 Elsevier Ltd. All rights reserved.

Precise Nuclear Data Measurements Possible with the NIFFTE fissionTPC for Advanced Reactor Designs

NASA Astrophysics Data System (ADS)

Towell, Rusty; Niffte Collaboration

2015-10-01

The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) Collaboration has applied the proven technology of Time Projection Chambers (TPC) to the task of precisely measuring fission cross sections. With the NIFFTE fission TPC, precise measurements have been made during the last year at the Los Alamos Neutron Science Center from both U-235 and Pu-239 targets. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow systematics to be controlled at the level of 1%. The fissionTPC performance will be presented. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors such as Liquid Fluoride Thorium Reactors over uranium-fueled reactors. These advantages include improved reactor safety, minimizing radioactive waste, improved reactor efficiency, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium-fueled reactors will also be discussed.

Research and Engineering Operation, Irradiation Processing Department monthly record report, May 1965

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

Ambrose, T.W.

1965-06-04

Process and development activities reported include: depleted uranium irradiations, thoria irradiation, and hot die sizing. Reactor engineering activities include: brittle fracture of 190-C tanks, increased graphite temperature limits for the F reactor, VSR channel caulking, K reactor downcomer flow, zircaloy hydriding, and ribbed zircaloy process tubes. Reactor physics activities include: thoria irradiations, E-D irradiations, boiling protection with the high speed scanner, and in-core flux monitoring. Radiological engineering activities include: radiation control, classification, radiation occurrences, effluent activity data, and well car shielding. Process standards are listed, along with audits, and fuel failure experience. Operational physics and process physics studies are presented.more » Lastly, testing activities are detailed.« less

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

Update on reactors and reactor instruments in Asia

NASA Astrophysics Data System (ADS)

Rao, K. R.

1991-10-01

The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

Multiscale Simulations of ALD in Cross Flow Reactors

DOE PAGES

Yanguas-Gil, Angel; Libera, Joseph A.; Elam, Jeffrey W.

2014-08-13

In this study, we have developed a multiscale simulation code that allows us to study the impact of surface chemistry on the coating of large area substrates with high surface area/high aspect-ratio features. Our code, based on open-source libraries, takes advantage of the ALD surface chemistry to achieve an extremely efficient two-way coupling between reactor and feature length scales, and it can provide simulated quartz crystal microbalance and mass spectrometry data at any point of the reactor. By combining experimental surface characterization with simple analysis of growth profiles in a tubular cross flow reactor, we are able to extract amore » minimal set of reactions to effectively model the surface chemistry, including the presence of spurious CVD, to evaluate the impact of surface chemistry on the coating of large, high surface area substrates.« less

Vibro-acoustic Imaging at the Breazeale Reactor

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

Smith, James Arthur; Jewell, James Keith; Lee, James Edwin

2016-09-01

The INL is developing Vibro-acoustic imaging technology to characterize microstructure in fuels and materials in spent fuel pools and within reactor vessels. A vibro-acoustic development laboratory has been established at the INL. The progress in developing the vibro-acoustic technology at the INL is the focus of this report. A successful technology demonstration was performed in a working TRIGA research reactor. Vibro-acoustic imaging was performed in the reactor pool of the Breazeale reactor in late September of 2015. A confocal transducer driven at a nominal 3 MHz was used to collect the 60 kHz differential beat frequency induced in a spentmore » TRIGA fuel rod and empty gamma tube located in the main reactor water pool. Data was collected and analyzed with the INLDAS data acquisition software using a short time Fourier transform.« less

Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

ERIC Educational Resources Information Center

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-01-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

Development of a simultaneous partial nitrification, anaerobic ammonia oxidation and denitrification (SNAD) bench scale process for removal of ammonia from effluent of a fertilizer industry.

PubMed

Keluskar, Radhika; Nerurkar, Anuradha; Desai, Anjana

2013-02-01

A simultaneous partial nitrification, anammox and denitrification (SNAD) process was developed for the treatment of ammonia laden effluent of a fertilizer industry. Autotrophic aerobic and anaerobic ammonia oxidizing biomass was enriched and their ammonia removal ability was confirmed in synthetic effluent system. Seed consortium developed from these was applied in the treatment of effluent in an oxygen limited bench scale SNAD type (1L) reactor run at ambient temperature (∼30°C). Around 98.9% ammonia removal was achieved with ammonia loading rate 0.35kgNH(4)(+)-N/m(3)day in the presence of 46.6mg/L COD at 2.31days hydraulic retention time. Qualitative and quantitative analysis of the biomass from upper and lower zone of the reactor revealed presence of autotrophic ammonia oxidizing bacteria (AOB), Planctomycetes and denitrifiers as the dominant bacteria carrying out anoxic oxidation of ammonia in the reactor. Physiological and molecular studies strongly indicate presence of anammox bacteria in the anoxic zone of the SNAD reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

Study of parameters affecting the conversion in a plug flow reactor for reactions of the type 2A→B

NASA Astrophysics Data System (ADS)

Beltran-Prieto, Juan Carlos; Long, Nguyen Huynh Bach Son

2018-04-01

Modeling of chemical reactors is an important tool to quantify reagent conversion, product yield and selectivity towards a specific compound and to describe the behavior of the system. Proposal of differential equations describing the mass and energy balance are among the most important steps required during the modeling process as they play a special role in the design and operation of the reactor. Parameters governing transfer of heat and mass have a strong relevance in the rate of the reaction. Understanding this information is important for the selection of reactor and operating regime. In this paper we studied the irreversible gas-phase reaction 2A→B. We model the conversion that can be achieved as function of the reactor volume and feeding temperature. Additionally, we discuss the effect of activation energy and the heat of reaction on the conversion achieved in the tubular reactor. Furthermore, we considered that dimerization occurs instantaneously in the catalytic surface to develop equations for the determination of rate of reaction per unit area of three different catalytic surface shapes. This data can be combined with information about the global rate of conversion in the reactor to improve regent conversion and yield of product.

Technical Requirements For Reactors To Be Deployed Internationally For the Global Nuclear Energy Partnership

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

Ingersoll, Daniel T

2007-01-01

Technical Requirements For Reactors To Be Deployed Internationally For the Global Nuclear Energy Partnership Robert Price U.S. Department of Energy, 1000 Independence Ave, SW, Washington, DC 20585, Daniel T. Ingersoll Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6162, INTRODUCTION The Global Nuclear Energy Partnership (GNEP) seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy. Fully meeting the GNEP vision may require the deployment of thousands of reactors in scores of countries, many of which do not use nuclear energy currently. Some of these needs will be met by large-scalemore » Generation III and III+ reactors (>1000 MWe) and Generation IV reactors when they are available. However, because many developing countries have small and immature electricity grids, the currently available Generation III(+) reactors may be unsuitable since they are too large, too expensive, and too complex. Therefore, GNEP envisions new types of reactors that must be developed for international deployment that are "right sized" for the developing countries and that are based on technologies, designs, and policies focused on reducing proliferation risk. The first step in developing such systems is the generation of technical requirements that will ensure that the systems meet both the GNEP policy goals and the power needs of the recipient countries. REQUIREMENTS Reactor systems deployed internationally within the GNEP context must meet a number of requirements similar to the safety, reliability, economics, and proliferation goals established for the DOE Generation IV program. Because of the emphasis on deployment to nonnuclear developing countries, the requirements will be weighted differently than with Generation IV, especially regarding safety and non-proliferation goals. Also, the reactors should be sized for market conditions in developing countries where energy demand per capita, institutional maturity and industrial infrastructure vary considerably, and must utilize fuel that is compatible with the fuel recycle technologies being developed by GNEP. Arrangements are already underway to establish Working Groups jointly with Japan and Russia to develop requirements for reactor systems. Additional bilateral and multilateral arrangements are expected as GNEP progresses. These Working Groups will be instrumental in establishing an international consensus on reactor system requirements. GNEP CERTIFICATION After establishing an accepted set of requirements for new reactors that are deployed internationally, a mechanism is needed that allows capable countries to continue to market their reactor technologies and services while assuring that they are compatible with GNEP goals and technologies. This will help to preserve the current system of open, commercial competition while steering the international community to meet common policy goals. The proposed vehicle to achieve this is the concept of GNEP Certification. Using objective criteria derived from the technical requirements in several key areas such as safety, security, non-proliferation, and safeguards, reactor designs could be evaluated and then certified if they meet the criteria. This certification would ensure that reactor designs meet internationally approved standards and that the designs are compatible with GNEP assured fuel services. SUMMARY New "right sized" power reactor systems will need to be developed and deployed internationally to fully achieve the GNEP vision of an expanded use of nuclear energy world-wide. The technical requirements for these systems are being developed through national and international Working Groups. The process is expected to culminate in a new GNEP Certification process that enables commercial competition while ensuring that the policy goals of GNEP are adequately met.« less

DEVELOPMENT OF AGENTS AND PROCEDURES FOR DECONTAMINATION OF THE YANKEE REACTOR PRIMARY COOLANT SYSTEM

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

Watkins, R.M.

1959-03-01

Developments relative to decontamination achieved under the Yankee Reasearch and Development program are reported. The decontamination of a large test loop which had been used to conduct corrosion rate studies for the Yankee reactor program is described. The basic permanganate-citrate decontamination procedure suggested for application in Yankee reactor primary system cleanup was used. A study of the chemistry of this decontamination operation is presented, together with conclusions pertaining to the effectiveness of the solutions under the conditions studied. In an attempt to further improve the efficiency of the procedure, an additional series of static and dynamic tests was performcd usingmore » contaminated sections of stainless steel tubing from the original SlW steam generator. Survival variables in the process (reagent composition, contact time, temperature, and flow velocity) were studied. The changes in decontamination efficiency produced by these variations are discussed and compared with results obtained throughthe use of similar procedures. Based on the observations made, conclusions are drawn concerning the optimum conditions for this cleanup process, a new set of suggested basic permanganate-citrate decontamination instructions is presented, and recommendations are made concerning future studies involving this procedure. (auth)« less

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

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

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

Short- and long-term responses to molybdenum-99 shortages in nuclear medicine.

PubMed

Ballinger, J R

2010-11-01

Most nuclear medicine studies use (99)Tc(m), which is the decay product of (99)Mo. The world supply of (99)Mo comes from only five nuclear research reactors and availability has been much reduced in recent times owing to problems at the largest reactors. In the short-term there are limited actions that can be taken owing to capacity issues on alternative imaging modalities. In the long-term, stability of (99)Mo supply will rely on a combination of replacing conventional reactors and developing new technologies.

Short- and long-term responses to molybdenum-99 shortages in nuclear medicine

PubMed Central

Ballinger, J R

2010-01-01

Most nuclear medicine studies use 99Tcm, which is the decay product of 99Mo. The world supply of 99Mo comes from only five nuclear research reactors and availability has been much reduced in recent times owing to problems at the largest reactors. In the short-term there are limited actions that can be taken owing to capacity issues on alternative imaging modalities. In the long-term, stability of 99Mo supply will rely on a combination of replacing conventional reactors and developing new technologies. PMID:20965898

Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

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

Snoj, L.; Sklenka, L.; Rataj, J.

2012-07-01

The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three differentmore » research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)« less

Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

PubMed

Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

2018-01-04

A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

Energy from the Atom. A Basic Teaching Unit on Energy. Revised.

ERIC Educational Resources Information Center

McDermott, Hugh, Ed.; Scharmann, Larry, Ed.

Recommended for grades 9-12 social studies and/or physical science classes, this 4-8 day unit focuses on four topics: (1) the background and history of atomic development; (2) two common types of nuclear reactors (boiling water and pressurized water reactors); (3) disposal of radioactive waste; and (4) the future of nuclear energy. Each topic…

Nuclear system that burns its own wastes shows promise

NASA Technical Reports Server (NTRS)

Atchison, K.

1975-01-01

A nuclear fission energy system, capable of eliminating a significant amount of its radioactive wastes by burning them, is described. A theoretical investigation of this system conducted by computer analysis, is based on use of gaseous fuel nuclear reactors. Gaseous core reactors using a uranium plasma fuel are studied along with development for space propulsion.

Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

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

Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.

A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

Immobilized lysozyme for the continuous lysis of lactic bacteria in wine: Bench-scale fluidized-bed reactor study.

PubMed

Cappannella, Elena; Benucci, Ilaria; Lombardelli, Claudio; Liburdi, Katia; Bavaro, Teodora; Esti, Marco

2016-11-01

Lysozyme from hen egg white (HEWL) was covalently immobilized on spherical supports based on microbial chitosan in order to develop a system for the continuous, efficient and food-grade enzymatic lysis of lactic bacteria (Oenococcus oeni) in white and red wine. The objective is to limit the sulfur dioxide dosage required to control malolactic fermentation, via a cell concentration typical during this process. The immobilization procedure was optimized in batch mode, evaluating the enzyme loading, the specific activity, and the kinetic parameters in model wine. Subsequently, a bench-scale fluidized-bed reactor was developed, applying the optimized process conditions. HEWL appeared more effective in the immobilized form than in the free one, when the reactor was applied in real white and red wine. This preliminary study suggests that covalent immobilization renders the enzyme less sensitive to the inhibitory effect of wine flavans. Copyright © 2016 Elsevier Ltd. All rights reserved.

Reactor Operations Monitoring System

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

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

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…

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

Rahayu, Suparni Setyowati, E-mail: suparnirahayu@yahoo.co.id; Department of Mechanical Engineering, State Polytechnic of Semarang, Semarang Indonesia; Purwanto,, E-mail: p.purwanto@che.undip.ac.id

The small industry of tofu production process releases the waste water without being processed first, and the wastewater is directly discharged into water. In this study, Anaerobic Sequencing Batch Reactor in Pilot Scale for Treatment of Tofu Industry was developed through an anaerobic process to produce biogas as one kind of environmentally friendly renewable energy which can be developed into the countryside. The purpose of this study was to examine the fundamental characteristics of organic matter elimination of industrial wastewater with small tofu effective method and utilize anaerobic active sludge with Anaerobic Sequencing Bath Reactor (ASBR) to get rural biogasmore » as an energy source. The first factor is the amount of the active sludge concentration which functions as the decomposers of organic matter and controlling selectivity allowance to degrade organic matter. The second factor is that HRT is the average period required substrate to react with the bacteria in the Anaerobic Sequencing Bath Reactor (ASBR).The results of processing the waste of tofu production industry using ASBR reactor with active sludge additions as starter generates cumulative volume of 5814.4 mL at HRT 5 days so that in this study it is obtained the conversion 0.16 L of CH{sub 4}/g COD and produce biogas containing of CH{sub 4}: 81.23% and CO{sub 2}: 16.12%. The wastewater treatment of tofu production using ASBR reactor is able to produce renewable energy that has economic value as well as environmentally friendly by nature.« less

Synchronized fusion development considering physics, materials and heat transfer

NASA Astrophysics Data System (ADS)

Wong, C. P. C.; Liu, Y.; Duan, X. R.; Xu, M.; Li, Q.; Feng, K. M.; Zheng, G. Y.; Li, Z. X.; Wang, X. Y.; Li, B.; Zhang, G. S.

2017-12-01

Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q  =  10. ITER will demonstrate D-T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m-2 and an average heat flux of 0.1 MW m-2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like W f/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

Modeling residence-time distribution in horizontal screw hydrolysis reactors

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

Sievers, David A.; Stickel, Jonathan J.

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Current and future trends for biofilm reactors for fermentation processes.

PubMed

Ercan, Duygu; Demirci, Ali

2015-03-01

Biofilms in the environment can both cause detrimental and beneficial effects. However, their use in bioreactors provides many advantages including lesser tendencies to develop membrane fouling and lower required capital costs, their higher biomass density and operation stability, contribution to resistance of microorganisms, etc. Biofilm formation occurs naturally by the attachment of microbial cells to the support without use of any chemicals agent in biofilm reactors. Biofilm reactors have been studied and commercially used for waste water treatment and bench and pilot-scale production of value-added products in the past decades. It is important to understand the fundamentals of biofilm formation, physical and chemical properties of a biofilm matrix to run the biofilm reactor at optimum conditions. This review includes the principles of biofilm formation; properties of a biofilm matrix and their roles in the biofilm formation; factors that improve the biofilm formation, such as support materials; advantages and disadvantages of biofilm reactors; and industrial applications of biofilm reactors.

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

Modeling residence-time distribution in horizontal screw hydrolysis reactors

DOE PAGES

Sievers, David A.; Stickel, Jonathan J.

2017-10-12

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Design and development of a prototype wet oxidation system for the reclamation of water and the disposition of waste residues onboard space vehicles

NASA Technical Reports Server (NTRS)

Jagow, R. B.

1972-01-01

Laboratory investigations to define optimum process conditions for oxidation of fecal/urine slurries were conducted in a one-liter batch reactor. The results of these tests formed the basis for the design, fabrication, and testing of an initial prototype system, including a 100-hour design verification test. Areas of further development were identified during this test. Development of a high pressure slurry pump, materials corrosion studies, oxygen supply trade studies, comparison of salt removal water recovery devices, ammonia removal investigation, development of a solids grinder, reactor design studies and bearing life tests, and development of shutoff valves and a back pressure regulator were undertaken. The development work has progressed to the point where a prototype system suitable for manned chamber testing can be fabricated and tested with a high degree of confidence of success.

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.

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

A Single-Granule-Level Approach Reveals Ecological Heterogeneity in an Upflow Anaerobic Sludge Blanket Reactor

PubMed Central

Mei, Ran; Narihiro, Takashi; Bocher, Benjamin T. W.; Yamaguchi, Takashi; Liu, Wen-Tso

2016-01-01

Upflow anaerobic sludge blanket (UASB) reactor has served as an effective process to treat industrial wastewater such as purified terephthalic acid (PTA) wastewater. For optimal UASB performance, balanced ecological interactions between syntrophs, methanogens, and fermenters are critical. However, much of the interactions remain unclear because UASB have been studied at a “macro”-level perspective of the reactor ecosystem. In reality, such reactors are composed of a suite of granules, each forming individual micro-ecosystems treating wastewater. Thus, typical approaches may be oversimplifying the complexity of the microbial ecology and granular development. To identify critical microbial interactions at both macro- and micro- level ecosystem ecology, we perform community and network analyses on 300 PTA–degrading granules from a lab-scale UASB reactor and two full-scale reactors. Based on MiSeq-based 16S rRNA gene sequencing of individual granules, different granule-types co-exist in both full-scale reactors regardless of granule size and reactor sampling depth, suggesting that distinct microbial interactions occur in different granules throughout the reactor. In addition, we identify novel networks of syntrophic metabolic interactions in different granules, perhaps caused by distinct thermodynamic conditions. Moreover, unseen methanogenic relationships (e.g. “Candidatus Aminicenantes” and Methanosaeta) are observed in UASB reactors. In total, we discover unexpected microbial interactions in granular micro-ecosystems supporting UASB ecology and treatment through a unique single-granule level approach. PMID:27936088

Using the surface charge profiler for in-line monitoring of doping concentration in silicon epitaxial wafer manufacturing

NASA Astrophysics Data System (ADS)

Tower, Joshua P.; Kamieniecki, Emil; Nguyen, M. C.; Danel, Adrien

1999-08-01

The Surface Charge Profiler (SCP) has been introduced for monitoring and development of silicon epitaxial processes. The SCP measures the near-surface doping concentration and offers advantages that lead to yield enhancement in several ways. First, non-destructive measurement technology enables in-line process monitoring, eliminating the need to sacrifice production wafers for resistivity measurements. Additionally, the full-wafer mapping capability helps in development of improved epitaxial growth processes and early detection of reactor problems. As examples, we present the use of SCP to study the effects of susceptor degradation in barrel reactors and to study autodoping for development of improved dopant uniformity.

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

Development of a new plasma reactor for propene removal

NASA Astrophysics Data System (ADS)

Oukacine, Linda; Tatibouët, Jean-Michel

2008-10-01

The purpose of the study is to develop a new plasma reactor being applied to gas phase pollution abatement, involving a surface dielectric barrier discharge (SDBD) at atmospheric pressure. Propene was chosen as a model pollutant. The system can associate a SDBD with a volume dielectric barrier discharge (VDBD). A specific catalyst can be placed in post-plasma site in order to destroy the residual ozone after use it as a strong oxidant for total oxidation of propene and by-products formed by the plasma reactor. A comparative study has been established between the propene removal efficiency of these two plasma geometries. The results demonstrate that SDBD is a promising system for gas cleaning. The experiments show that ozone production depends on plasma system configuration and indicate the effectiveness of combining SDBD and VDBD. The NOx formation remains very low, whereas ozone formation is the highest for the SDBD. The influence of some materials on the propene removal and the ozone production were studied.

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.

A brief history of design studies on innovative nuclear reactors

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

Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com

2014-09-30

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USAmore » and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.« less

Low-temperature catalytic gasification of food processing wastes. 1995 topical report

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

Elliott, D.C.; Hart, T.R.

The catalytic gasification system described in this report has undergone continuing development and refining work at Pacific Northwest National Laboratory (PNNL) for over 16 years. The original experiments, performed for the Gas Research Institute, were aimed at developing kinetics information for steam gasification of biomass in the presence of catalysts. From the fundamental research evolved the concept of a pressurized, catalytic gasification system for converting wet biomass feedstocks to fuel gas. Extensive batch reactor testing and limited continuous stirred-tank reactor tests provided useful design information for evaluating the preliminary economics of the process. This report is a follow-on to previousmore » interim reports which reviewed the results of the studies conducted with batch and continuous-feed reactor systems from 1989 to 1994, including much work with food processing wastes. The discussion here provides details of experiments on food processing waste feedstock materials, exclusively, that were conducted in batch and continuous- flow reactors.« less

Hodoscope Cineradiography Of Nuclear Fuel Destruction Experiments

NASA Astrophysics Data System (ADS)

De Volpi, A.

1983-08-01

Nuclear reactor safety studies have applied cineradiographic techniques to achieve key information regarding the durability of fuel elements that are subjected to destructive transients in test reactors. Beginning with its development in 1963, the fast-neutron hodoscope has recorded data at the TREAT reactor in the United States of America. Consisting of a collimator instrumented with several hundred parallel channels of detectors and associated instrumentation, the hodoscope measures fuel motion that takes place within thick-walled steel test containers. Fuel movement is determined by detecting the emission of fast neutrons induced in the test capsule by bursts of the test reactor that last from 0.3 to 30 s. The system has been designed so as to achieve under certain typical conditions( horizontal) spatial resolution less than lmm, time resolution close to lms, mass resolution below 0.1 g, with adequate dynamic range and recording duration. A variety of imaging forms have been developed to display the results of processing and analyzing recorded data.*

A probabilistic safety analysis of incidents in nuclear research reactors.

PubMed

Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

2012-06-01

This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

Initial conceptual design study of self-critical nuclear pumped laser systems

NASA Technical Reports Server (NTRS)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS Advanced Reactor is a 640-MW(e) pressurized-water reactor developed by Asea Brown Boveri. A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity normally is controlled by the boron concentration in the coolant and the temperature of the moderator coolant. Analyses of five initiating events have been completed on the basis of calculations performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. The initiating events analyzed are (1) reactor scram, (2) loss of off-site power (3) main steam-line break, (4) small-break loss of coolant, and (5) large-break loss of coolant. Inmore » addition to the baseline calculation for each sequence, sensitivity studies were performed to explore the response of the PIUS reactor to severe off-normal conditions having a very low probability of occurrence. The sensitivity studies provide insights into the robustness of the design.« less

Biogasification of community-derived biomass and solid wastes in a pilot-scale SOLCON reactor

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

Srivastava, V.J.; Biljetina, R.; Isaacson, H.R.

1988-01-01

The Institute of Gas Technology has developed a novel, solids- concentrating (SOLCON) bioreactor to convert a variety of individual or mixed feedstocks (biomass and wastes) to methane at higher rates and efficiencies than those obtained from conventional high-rate anaerobic digesters. The biogasification studies are being conducted in a pilot-scale experimental test unit (ETU) located in the Walt Disney World Resort Complex, Orlando, Florida. This paper describes the ETU facility, the logistics of feedstock integration, the SOLCON reactor design and operating techniques, and the results obtained during 4 years of stable, uninterrupted operation with different feedstocks. The SOLCON reactor consistently outperformedmore » the conventional stirred-tank reactor by 20% to 50%.« less

Anaerobic treatability of wastewater contaminated with propylene glycol.

PubMed

Sezgin, Naim; Tonuk, Gulseven Ubay

2013-09-01

The purpose of this study was to investigate the biodegradability of propylene glycol in anaerobic conditions by using methanogenic culture. A master reactor was set up to develop a culture that would be acclimated to propylene glycol. After reaching steady-state, culture was transferred to serum bottles. Three reactors with same initial conditions were run for consistency. Propylene glycol was completely biodegradable under anaerobic methanogenic conditions. Semi-continuous reactors operated at a temperature of 35°C had consistently achieved a propylene glycol removal of higher than 95 % based on chemical oxygen demand (COD). It was found that in semi-continuous reactors, anaerobic treatment of propylene glycol at concentrations higher than 1,500 mg COD m(-3) day(-1) was not convenient due to instable effluent COD.

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

Nuclear energy.

PubMed

Grandin, Karl; Jagers, Peter; Kullander, Sven

2010-01-01

Nuclear energy can play a role in carbon free production of electrical energy, thus making it interesting for tomorrow's energy mix. However, several issues have to be addressed. In fission technology, the design of so-called fourth generation reactors show great promise, in particular in addressing materials efficiency and safety issues. If successfully developed, such reactors may have an important and sustainable part in future energy production. Working fusion reactors may be even more materials efficient and environmental friendly, but also need more development and research. The roadmap for development of fourth generation fission and fusion reactors, therefore, asks for attention and research in these fields must be strengthened.

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

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

Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Yazid, Putranto Ilham

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With themore » tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.« less

Radiation Shielding Design and Orientation Considerations for a 1 kWe Heat Pipe Cooled Reactor Utilized to Bore Through the Ice Caps of Mars

NASA Astrophysics Data System (ADS)

Fensin, Michael L.; Elliott, John O.; Lipinski, Ronald J.; Poston, David I.

2006-01-01

The goal in designing any space power system is to develop a system able to meet the mission requirements for success while minimizing the overall costs. The mission requirements for the this study was to develop a reactor (with Stirling engine power conversion) and shielding configuration able to fit, along with all the other necessary science equipment, in a Cryobot 3 m high with ~0.5 m diameter hull, produce 1 kWe for 5yrs, and not adversely affect the mission science by keeping the total integrated dose to the science equipment below 150 krad. Since in most space power missions the overall system mass dictates the mission cost, the shielding designs in this study incorporated Martian water extracted at the startup site in order to minimize the tungsten and LiH mass loading at launch. Different reliability and mass minimization concerns led to three design configuration evolutions. With the help of implementing Martian water and configuring the reactor as far from the science equipment as possible, the needed tungsten and LiH shield mass was minimized. This study further characterizes the startup dose and the necessary mission requirements in order to ensure integrity of the surface equipment during reactor startup phase.

Automatic reactor for solid-phase synthesis of molecularly imprinted polymeric nanoparticles (MIP NPs) in water.

PubMed

Poma, Alessandro; Guerreiro, Antonio; Caygill, Sarah; Moczko, Ewa; Piletsky, Sergey

We report the development of an automated chemical reactor for solid-phase synthesis of MIP NPs in water. Operational parameters are under computer control, requiring minimal operator intervention. In this study, "ready for use" MIP NPs with sub-nanomolar affinity are prepared against pepsin A, trypsin and α-amylase in only 4 hours.

Automatic reactor for solid-phase synthesis of molecularly imprinted polymeric nanoparticles (MIP NPs) in water

PubMed Central

Poma, Alessandro; Guerreiro, Antonio; Caygill, Sarah; Moczko, Ewa; Piletsky, Sergey

2015-01-01

We report the development of an automated chemical reactor for solid-phase synthesis of MIP NPs in water. Operational parameters are under computer control, requiring minimal operator intervention. In this study, “ready for use” MIP NPs with sub-nanomolar affinity are prepared against pepsin A, trypsin and α-amylase in only 4 hours. PMID:26722622

Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

NASA Astrophysics Data System (ADS)

Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

2018-01-01

The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

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

Heeger, Karsten M.

2014-09-13

This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zeromore » $$\\theta_{13}$$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.« less

Example study for granular bioreactor stratification: Three-dimensional evaluation of a sulfate-reducing granular bioreactor

PubMed Central

Hao, Tian-wei; Luo, Jing-hai; Su, Kui-zu; Wei, Li; Mackey, Hamish R.; Chi, Kun; Chen, Guang-Hao

2016-01-01

Recently, sulfate-reducing granular sludge has been developed for application in sulfate-laden water and wastewater treatment. However, little is known about biomass stratification and its effects on the bioprocesses inside the granular bioreactor. A comprehensive investigation followed by a verification trial was therefore conducted in the present work. The investigation focused on the performance of each sludge layer, the internal hydrodynamics and microbial community structures along the height of the reactor. The reactor substratum (the section below baffle 1) was identified as the main acidification zone based on microbial analysis and reactor performance. Two baffle installations increased mixing intensity but at the same time introduced dead zones. Computational fluid dynamics simulation was employed to visualize the internal hydrodynamics. The 16S rRNA gene of the organisms further revealed that more diverse communities of sulfate-reducing bacteria (SRB) and acidogens were detected in the reactor substratum than in the superstratum (the section above baffle 1). The findings of this study shed light on biomass stratification in an SRB granular bioreactor to aid in the design and optimization of such reactors. PMID:27539264

AGC 2 Irradiated Material Properties Analysis

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

Rohrbaugh, David Thomas

2017-05-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core componentsmore » within a commercial reactor, a complete 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 accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

AGC 2 Irradiation Creep Strain Data Analysis

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

Windes, William E.; Rohrbaugh, David T.; Swank, W. David

2016-08-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within amore » commercial reactor, a complete 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 accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

Development of Cross Section Library and Application Programming Interface (API)

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

Lee, C. H.; Marin-Lafleche, A.; Smith, M. A.

2014-04-09

The goal of NEAMS neutronics is to develop a high-fidelity deterministic neutron transport code termed PROTEUS for use on all reactor types of interest, but focused primarily on sodium-cooled fast reactors. While PROTEUS-SN has demonstrated good accuracy for homogeneous fast reactor problems and partially heterogeneous fast reactor problems, the simulation results were not satisfactory when applied on fully heterogeneous thermal problems like the Advanced Test Reactor (ATR). This is mainly attributed to the quality of cross section data for heterogeneous geometries since the conventional cross section generation approach does not work accurately for such irregular and complex geometries. Therefore, onemore » of the NEAMS neutronics tasks since FY12 has been the development of a procedure to generate appropriate cross sections for a heterogeneous geometry core.« less

Early Program Development

NASA Image and Video Library

1963-01-01

This artist's concept from 1963 shows a proposed NERVA (Nuclear Engine for Rocket Vehicle Application) incorporating the NRX-A1, the first NERVA-type cold flow reactor. The NERVA engine, based on Kiwi nuclear reactor technology, was intended to power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which Marshall Space Flight Center had development responsibility.

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

Kim, Yeon Soo; Jeong, G. Y.; Sohn, D. -S.

U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data setmore » of full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model.« less

Interim waste storage for the Integral Fast Reactor

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

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

Alternative nuclear technologies

NASA Astrophysics Data System (ADS)

Schubert, E.

1981-10-01

The lead times required to develop a select group of nuclear fission reactor types and fuel cycles to the point of readiness for full commercialization are compared. Along with lead times, fuel material requirements and comparative costs of producing electric power were estimated. A conservative approach and consistent criteria for all systems were used in estimates of the steps required and the times involved in developing each technology. The impact of the inevitable exhaustion of the low- or reasonable-cost uranium reserves in the United States on the desirability of completing the breeder reactor program, with its favorable long-term result on fission fuel supplies, is discussed. The long times projected to bring the most advanced alternative converter reactor technologies the heavy water reactor and the high-temperature gas-cooled reactor into commercial deployment when compared to the time projected to bring the breeder reactor into equivalent status suggest that the country's best choice is to develop the breeder. The perceived diversion-proliferation problems with the uranium plutonium fuel cycle have workable solutions that can be developed which will enable the use of those materials at substantially reduced levels of diversion risk.

Radiation chemistry for modern nuclear energy development

NASA Astrophysics Data System (ADS)

Chmielewski, Andrzej G.; Szołucha, Monika M.

2016-07-01

Radiation chemistry plays a significant role in modern nuclear energy development. Pioneering research in nuclear science, for example the development of generation IV nuclear reactors, cannot be pursued without chemical solutions. Present issues related to light water reactors concern radiolysis of water in the primary circuit; long-term storage of spent nuclear fuel; radiation effects on cables and wire insulation, and on ion exchangers used for water purification; as well as the procedures of radioactive waste reprocessing and storage. Radiation effects on materials and enhanced corrosion are crucial in current (II/III/III+) and future (IV) generation reactors, and in waste management, deep geological disposal and spent fuel reprocessing. The new generation of reactors (III+ and IV) impose new challenges for radiation chemists due to their new conditions of operation and the usage of new types of coolant. In the case of the supercritical water-cooled reactor (SCWR), water chemistry control may be the key factor in preventing corrosion of reactor structural materials. This paper mainly focuses on radiation effects on long-term performance and safety in the development of nuclear power plants.

Dynamic Modeling and Control Studies of a Two-Stage Bubbling Fluidized Bed Adsorber-Reactor for Solid-Sorbent CO{sub 2} Capture

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

Modekurti, Srinivasarao; Bhattacharyya, Debangsu; Zitney, Stephen E.

2013-07-31

A one-dimensional, non-isothermal, pressure-driven dynamic model has been developed for a two-stage bubbling fluidized bed (BFB) adsorber-reactor for solid-sorbent carbon dioxide (CO{sub 2}) capture using Aspen Custom Modeler® (ACM). The BFB model for the flow of gas through a continuous phase of downward moving solids considers three regions: emulsion, bubble, and cloud-wake. Both the upper and lower reactor stages are of overflow-type configuration, i.e., the solids leave from the top of each stage. In addition, dynamic models have been developed for the downcomer that transfers solids between the stages and the exit hopper that removes solids from the bottom ofmore » the bed. The models of all auxiliary equipment such as valves and gas distributor have been integrated with the main model of the two-stage adsorber reactor. Using the developed dynamic model, the transient responses of various process variables such as CO{sub 2} capture rate and flue gas outlet temperatures have been studied by simulating typical disturbances such as change in the temperature, flowrate, and composition of the incoming flue gas from pulverized coal-fired power plants. In control studies, the performance of a proportional-integral-derivative (PID) controller, feedback-augmented feedforward controller, and linear model predictive controller (LMPC) are evaluated for maintaining the overall CO{sub 2} capture rate at a desired level in the face of typical disturbances.« less

Development of a Model and Computer Code to Describe Solar Grade Silicon Production Processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

The program aims at developing mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon. The major interest is in collecting silicon as a liquid on the reactor walls and other collection surfaces. Two reactor systems are of major interest, a SiCl4/Na reactor in which Si(l) is collected on the flow tube reactor walls and a reactor in which Si(l) droplets formed by the SiCl4/Na reaction are collected by a jet impingement method. During this quarter the following tasks were accomplished: (1) particle deposition routines were added to the boundary layer code; and (2) Si droplet sizes in SiCl4/Na reactors at temperatures below the dew point of Si are being calculated.

Catalyst and process development for synthesis gas conversion to isobutylene. Quarterly report, October 1, 1992--December 31, 1992

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

Anthony, R.G.; Akgerman, A.

1993-02-01

The objectives of this project are to develop a new catalyst, the kinetics for this catalyst, reactor models for trickle bed, slurry and fixed bed reactors, and simulate the performance of fixed bed trickle flow reactors, slurry flow reactors, and fixed bed gas phase reactors for conversion of a hydrogen lean synthesis gas to isobutylene. The goals for the quarter include: (1) Conduct experiments using a trickle bed reactor to determine the effect of reactor type on the product distribution. (2) Use spherical pellets of silica as a support for zirconia for the purpose of increasing surface, area and performancemore » of the catalysts. (3) Conduct exploratory experiments to determine the effect of super critical drying of the catalyst on the catalyst surface area and performance. (4) Prepare a ceria/zirconia catalyst by the precipitation method.« less

Producing Hydrogen With Sunlight

NASA Technical Reports Server (NTRS)

Biddle, J. R.; Peterson, D. B.; Fujita, T.

1987-01-01

Costs high but reduced by further research. Producing hydrogen fuel on large scale from water by solar energy practical if plant costs reduced, according to study. Sunlight attractive energy source because it is free and because photon energy converts directly to chemical energy when it breaks water molecules into diatomic hydrogen and oxygen. Conversion process low in efficiency and photochemical reactor must be spread over large area, requiring large investment in plant. Economic analysis pertains to generic photochemical processes. Does not delve into details of photochemical reactor design because detailed reactor designs do not exist at this early stage of development.

Development of an attached growth reactor for NH₄-N removal at a drinking water supply system in Kathmandu Valley, Nepal.

PubMed

Khanitchaidecha, Wilawan; Shakya, Maneesha; Nakano, Yuichi; Tanaka, Yasuhiro; Kazama, Futaba

2012-01-01

Higher concentrations of ammonium (NH(4)-N) and iron (Fe) than a standard for drinking are typical characteristics of groundwater in the study area. To remove NH(4)-N and Fe, the drinking water supply system in this study consists of a series of treatment units (i.e., aeration and sedimentation, filtration, and chlorination); however, NH(4)-N in treated water is higher than a standard for drinking (i.e., <1.5 mg NH(4)-N/L). The objective of this study, therefore, is to develop an attached growth system containing a fiber carrier for reducing NH(4)-N concentration within a safe level in the treated water. To avoid the need of air supply for nitrification, groundwater was continuously dripped through the reactor. It made the system simple operation and energy efficient. Effects of reactor design (reactor length and carrier area) were studied to achieve a high NH(4)-N removal efficiency. In accordance with raw groundwater characteristics in the area, effects of low inorganic carbon (IC) and phosphate (PO(4)-P) and high Fe on the removal efficiency were also investigated. The results showed a significant increase in NH(4)-N removal efficiency with reactor length and carrier area. A low IC and PO(4)-P had no effect on NH(4)-N removal, whereas a high Fe decreased the efficiency significantly. The first 550 days operation of a pilot-scale reactor installed in the drinking water supply system showed a gradual increase in the efficiency, reaching to 95-100%, and stability in the performance even with increased flow rate from 210 to 860 L/day. The high efficiency of the present work was indicated because only less than 1 mg of NH(4)-N/L was left over in the treated water.

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development

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

George W. Griffith

2011-10-01

A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the fibers and the matrix allows formore » ductile behavior. The SiC CMC has relatively high strength at high reactor accident temperatures when compared to metallic cladding. SiC also has a very low chemical reactivity and doesn't react exothermically with the reactor cooling water. The radiation behavior of SiC has also been studied extensively as structural fusion system components. The SiC CMC technology is in the early stages of development and will need to mature before confidence in the developed designs can created. The advanced SiC CMC materials do offer the potential for greatly improved safety because of their high temperature strength, chemical stability and reduced hydrogen generation.« less

Hybrid plasmachemical reactor

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

Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V.

2015-04-15

A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

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

GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography

NASA Astrophysics Data System (ADS)

Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.

2015-09-01

An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.

In-air and pressurized water reactor environment fatigue experiments of 316 stainless steel to study the effect of environment on cyclic hardening

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

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurindranath

Argonne National Laboratory (ANL), under the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316SS) material which is widely used in the US reactors. Contrary to the conventional S~N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening)more » under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In a second paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.« less

Current status of SPINNORs designs

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

Su'ud, Zaki

2010-06-22

This study discuss about the SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) and the VSPINNOR (Very Small Power Reactor, Indonesia, No On-site Refuelling) which are small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for more than 10 or 15 years without on-site refuelling. They are based on the concept of a long-life core reactor developed in Indonesia since early 1990 in collaboration with the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (RLNR TITech). The reactor cores are designed to have near zero (less then one effective delayed neutron fraction)more » burn-up reactivity swing during the whole course of their operation to avoid a possibility of prompt criticality accident. The basic concept is that central region of the reactor core is filled with fertile (blanket) material. During the reactor operation fissile material accumulates in this central region, which helps to compensate fissile material loss in the peripheral core region and also contributes to negative coolant loss reactivity effect. A concept of high fuel volume fraction in the core is applied to achieve smaller size of a critical reactor. In this paper we consider to add Np-237 to the fuel to enhance non proliferation characteristics of the systems. The effect of Np-237 amount variation is discussed.« less

Aging management program of the reactor building concrete at Point Lepreau Generating Station

NASA Astrophysics Data System (ADS)

Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

2011-04-01

In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

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

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal yearmore » 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.« 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%.

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

Assessing the degree of plug flow in oxidation flow reactors (OFRs): a study on a potential aerosol mass (PAM) reactor

NASA Astrophysics Data System (ADS)

Mitroo, Dhruv; Sun, Yujian; Combest, Daniel P.; Kumar, Purushottam; Williams, Brent J.

2018-03-01

Oxidation flow reactors (OFRs) have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate). While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs) inside the Washington University Potential Aerosol Mass (WU-PAM) reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS) model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD) simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study).

Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

NASA Astrophysics Data System (ADS)

Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

2016-01-01

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

1976-01-01

The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

Co-Production of Electricity and Hydrogen Using a Novel Iron-based Catalyst

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

Hilaly, Ahmad; Georgas, Adam; Leboreiro, Jose

2011-09-30

The primary objective of this project was to develop a hydrogen production technology for gasification applications based on a circulating fluid-bed reactor and an attrition resistant iron catalyst. The work towards achieving this objective consisted of three key activities: Development of an iron-based catalyst suitable for a circulating fluid-bed reactor; Design, construction, and operation of a bench-scale circulating fluid-bed reactor system for hydrogen production; Techno-economic analysis of the steam-iron and the pressure swing adsorption hydrogen production processes. This report describes the work completed in each of these activities during this project. The catalyst development and testing program prepared and iron-basedmore » catalysts using different support and promoters to identify catalysts that had sufficient activity for cyclic reduction with syngas and steam oxidation and attrition resistance to enable use in a circulating fluid-bed reactor system. The best performing catalyst from this catalyst development program was produced by a commercial catalyst toll manufacturer to support the bench-scale testing activities. The reactor testing systems used during material development evaluated catalysts in a single fluid-bed reactor by cycling between reduction with syngas and oxidation with steam. The prototype SIP reactor system (PSRS) consisted of two circulating fluid-bed reactors with the iron catalyst being transferred between the two reactors. This design enabled demonstration of the technical feasibility of the combination of the circulating fluid-bed reactor system and the iron-based catalyst for commercial hydrogen production. The specific activities associated with this bench-scale circulating fluid-bed reactor systems that were completed in this project included design, construction, commissioning, and operation. The experimental portion of this project focused on technical demonstration of the performance of an iron-based catalyst and a circulating fluid-bed reactor system for hydrogen production. Although a technology can be technically feasible, successful commercial deployment also requires that a technology offer an economic advantage over existing commercial technologies. To effective estimate the economics of this steam-iron process, a techno-economic analysis of this steam iron process and a commercial pressure swing adsorption process were completed. The results from this analysis described in this report show the economic potential of the steam iron process for integration with a gasification plant for coproduction of hydrogen and electricity.« less

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

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

Farmer, Mitchell T.; Bunt, R.; Corradini, M.

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affectmore » reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).« less

Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

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

None

The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

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.

On-Site Determination and Monitoring of Real-Time Fluence Delivery for an Operating UV Reactor Based on a True Fluence Rate Detector.

PubMed

Li, Mengkai; Li, Wentao; Qiang, Zhimin; Blatchley, Ernest R

2017-07-18

At present, on-site fluence (distribution) determination and monitoring of an operating UV system represent a considerable challenge. The recently developed microfluorescent silica detector (MFSD) is able to measure the approximate true fluence rate (FR) at a fixed position in a UV reactor that can be compared with a FR model directly. Hence it has provided a connection between model calculation and real-time fluence determination. In this study, an on-site determination and monitoring method of fluence delivery for an operating UV reactor was developed. True FR detectors, a UV transmittance (UVT) meter, and a flow rate meter were used for fundamental measurements. The fluence distribution, as well as reduction equivalent fluence (REF), 10th percentile dose in the UV fluence distribution (F 10 ), minimum fluence (F min ), and mean fluence (F mean ) of a test reactor, was calculated in advance by the combined use of computational fluid dynamics and FR field modeling. A field test was carried out on the test reactor for disinfection of a secondary water supply. The estimated real-time REF, F 10 , F min , and F mean decreased 73.6%, 71.4%, 69.6%, and 72.9%, respectively, during a 6-month period, which was attributable to lamp output attenuation and sleeve fouling. The results were analyzed with synchronous data from a previously developed triparameter UV monitoring system and water temperature sensor. This study allowed demonstration of an accurate method for on-site, real-time fluence determination which could be used to enhance the security and public confidence of UV-based water treatment processes.

POWER-BURST FACILITY (PBF) CONCEPTUAL DESIGN

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

Wasserman, A.A.; Johnson, S.O.; Heffner, R.E.

1963-06-21

A description is presented of the conceptual design of a high- performance, pulsed reactor called the Power Burst Facility (PBF). This reactor is designed to generate power bursts with initial asymptotic periods as short as 1 msec, producing energy releases large enough to destroy entire fuel subassemblies placed in a capsule or flow loop mounted in the reactor, all without damage to the reactor itself. It will be used primarily to evaluate the consequences and hazards of very rapid destructive accidents in reactors representing the entire range of current nuclear technology as applied to power generation, propulsion, and testing. Itmore » will also be used to carry out detailed studies of nondestructive reactivity feedback mechanisms in the shortperiod domain. The facility was designed to be sufficiently flexible to accommodate future cores of even more advanced design. The design for the first reactor core is based upon proven technology; hence, completion of the final design of this core will involve no significant development delays. Construction of the PBF is proposed to begin in September 1984, and is expected to take approximately 20 months to complete. (auth)« less

Gas-phase optical fiber photocatalytic reactors for indoor air application: a preliminary study on performance indicators

NASA Astrophysics Data System (ADS)

Palmiste, Ü.; Voll, H.

2017-10-01

The development of advanced air cleaning technologies aims to reduce building energy consumption by reduction of outdoor air flow rates while keeping the indoor air quality at an acceptable level by air cleaning. Photocatalytic oxidation is an emerging technology for gas-phase air cleaning that can be applied in a standalone unit or a subsystem of a building mechanical ventilation system. Quantitative information on photocatalytic reactor performance is required to evaluate the technical and economic viability of the advanced air cleaning by PCO technology as an energy conservation measure in a building air conditioning system. Photocatalytic reactors applying optical fibers as light guide or photocatalyst coating support have been reported as an approach to address the current light utilization problems and thus, improve the overall efficiency. The aim of the paper is to present a preliminary evaluation on continuous flow optical fiber photocatalytic reactors based on performance indicators commonly applied for air cleaners. Based on experimental data, monolith-type optical fiber reactor performance surpasses annular-type optical fiber reactors in single-pass removal efficiency, clean air delivery rate and operating cost efficiency.

Demonstration of Robustness and Integrated Operation of a Series-Bosch System

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, Matthew J.; Stanley, Christine; Barnett, Bill; Junaedi, Christian; Vilekar, Saurabh A.; Ryan, Kent

2016-01-01

Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a NASA-designed RWGS reactor designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith® technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with NASA's RWGS reactor. The test results will be provided in this paper. Separately, in 2015, a semi-continuous CF reactor was designed and fabricated at NASA based on the results from batch CF reactor testing. The batch CF reactor and the semi-continuous CF reactor were individually integrated with an upstream RWGS reactor to demonstrate the system operation and to evaluate performance. Here, we compare the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level.

Space station prototype Sabatier reactor design verification testing

NASA Technical Reports Server (NTRS)

Cusick, R. J.

1974-01-01

A six-man, flight prototype carbon dioxide reduction subsystem for the SSP ETC/LSS (Space Station Prototype Environmental/Thermal Control and Life Support System) was developed and fabricated for the NASA-Johnson Space Center between February 1971 and October 1973. Component design verification testing was conducted on the Sabatier reactor covering design and off-design conditions as part of this development program. The reactor was designed to convert a minimum of 98 per cent hydrogen to water and methane for both six-man and two-man reactant flow conditions. Important design features of the reactor and test conditions are described. Reactor test results are presented that show design goals were achieved and off-design performance was stable.

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

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-06-10

scale pressurized water reactors suitable for destroyer-sized vessels or for alternative nuclear power systems using thorium liquid salt technology...or to design a new reactor type potentially using a thorium liquid salt reactor developed for maritime use. The committee recommends an increase of...either using a pressurized water reactor or a thorium liquid salt reactor . (Page 158) Senate The Senate Armed Services Committee, in its report

Impact of non-ionic surfactant on the long-term development of lab-scale-activated sludge bacterial communities.

PubMed

Lozada, Mariana; Basile, Laura; Erijman, Leonardo

2007-01-01

The development of bacterial communities in replicate lab-scale-activated sludge reactors degrading a non-ionic surfactant was evaluated by statistical analysis of denaturing gradient gel electrophoresis (DGGE) fingerprints. Four sequential batch reactors were fed with synthetic sewage, two of which received, in addition, 0.01% of nonylphenol ethoxylates (NPE). The dynamic character of bacterial community structure was confirmed by the differences in species composition among replicate reactors. Measurement of similarities between reactors was obtained by pairwise similarity analysis using the Bray Curtis coefficient. The group of NPE-amended reactors exhibited the highest similarity values (Sjk=0.53+/-0.03), indicating that the bacterial community structure of NPE-amended reactors was better replicated than control reactors (Sjk=0.36+/-0.04). Replicate NPE-amended reactors taken at different times of operation clustered together, whereas analogous relations within the control reactor cluster were not observed. The DGGE pattern of isolates grown in conditioned media prepared with media taken at the end of the aeration cycle grouped separately from other conditioned and synthetic media regardless of the carbon source amendment, suggesting that NPE degradation residuals could have a role in the shaping of the community structure.

Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

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

Fourmentel, D.; Radulovic, V.; Barbot, L.

Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development atmore » the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to the delayed gamma rays. In this paper we describe experiments in each of the three reactors and how we estimate delayed gamma rays with MIC measurements. The results and perspectives are discussed. (authors)« less

A Potential NASA Research Reactor to Support NTR Development

NASA Technical Reports Server (NTRS)

Eades, Michael; Gerrish, Harold; Hardin, Leroy

2013-01-01

In support of efforts for research into the design and development of a man rated Nuclear Thermal Rocket (NTR) engine, the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed research reactor. The proposed reactor would be licensed by NASA and operated jointly by NASA and university partners. The purpose of this reactor would be to perform further research into the technologies and systems needed for a successful NTR project and promote nuclear training and education.

A coupled nuclear reactor thermal energy storage system for enhanced load following operation

NASA Astrophysics Data System (ADS)

Alameri, Saeed A.

Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES to absorb the decay heat of the reactor fuel while cooling the PAHTR after an emergency shutdown. The simulated reactivity insertion accident assessment determined the maximum allowable reactivity insertion to the PAHTR as a function of shutdown response times.

Electrochemical regeneration of phenol-saturated activated carbon - proposal of a reactor.

PubMed

Zanella, Odivan; Bilibio, Denise; Priamo, Wagner Luiz; Tessaro, Isabel Cristina; Féris, Liliana Amaral

2017-03-01

An electrochemical process was used to investigate the activated carbon regeneration efficiency (RE) saturated with aromatics. For this purpose, an electrochemical reactor was developed and the operational conditions of this equipment were investigated, which is applied in activated carbon regeneration process. The influence of regeneration parameters such as processing time, the current used, the polarity and the processing fluid (electrolyte) were studied. The performance of electrochemical regeneration was evaluated by adsorption tests, using phenol as adsorbate. The increase in current applied and the process time was found to enhance the RE. Another aspect that indicated a better reactor performance was the type of electrolyte used, showing best results for NaCl. The polarity showed the highest influence on the process, when the cathodic regeneration was more efficient. The electrochemical regeneration process developed in this study presented regeneration capacities greater than 100% when the best process conditions were used, showing that this form of regeneration for activated carbon saturated with aromatics is very promising.

Direct numerical simulation of reactor two-phase flows enabled by high-performance computing

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

Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.

Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent researchmore » progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.« less

Development of a Reactor Model for Chemical Conversion of Lunar Regolith

NASA Technical Reports Server (NTRS)

Hegde, U.; Balasubramaniam, R.; Gokoglu, S.

2009-01-01

Lunar regolith will be used for a variety of purposes such as oxygen and propellant production and manufacture of various materials. The design and development of chemical conversion reactors for processing lunar regolith will require an understanding of the coupling among the chemical, mass and energy transport processes occurring at the length and time scales of the overall reactor with those occurring at the corresponding scales of the regolith particles. To this end, a coupled transport model is developed using, as an example, the reduction of ilmenite-containing regolith by a continuous flow of hydrogen in a flow-through reactor. The ilmenite conversion occurs on the surface and within the regolith particles. As the ilmenite reduction proceeds, the hydrogen in the reactor is consumed, and this, in turn, affects the conversion rate of the ilmenite in the particles. Several important quantities are identified as a result of the analysis. Reactor scale parameters include the void fraction (i.e., the fraction of the reactor volume not occupied by the regolith particles) and the residence time of hydrogen in the reactor. Particle scale quantities include the time for hydrogen to diffuse into the pores of the regolith particles and the chemical reaction time. The paper investigates the relationships between these quantities and their impact on the regolith conversion. Application of the model to various chemical reactor types, such as fluidized-bed, packed-bed, and rotary-bed configurations, are discussed.

Development of a Reactor Model for Chemical Conversion of Lunar Regolith

NASA Technical Reports Server (NTRS)

Hedge, uday; Balasubramaniam, R.; Gokoglu, S.

2007-01-01

Lunar regolith will be used for a variety of purposes such as oxygen and propellant production and manufacture of various materials. The design and development of chemical conversion reactors for processing lunar regolith will require an understanding of the coupling among the chemical, mass and energy transport processes occurring at the length and time scales of the overall reactor with those occurring at the corresponding scales of the regolith particles. To this end, a coupled transport model is developed using, as an example, the reduction of ilmenite-containing regolith by a continuous flow of hydrogen in a flow-through reactor. The ilmenite conversion occurs on the surface and within the regolith particles. As the ilmenite reduction proceeds, the hydrogen in the reactor is consumed, and this, in turn, affects the conversion rate of the ilmenite in the particles. Several important quantities are identified as a result of the analysis. Reactor scale parameters include the void fraction (i.e., the fraction of the reactor volume not occupied by the regolith particles) and the residence time of hydrogen in the reactor. Particle scale quantities include the time for hydrogen to diffuse into the pores of the regolith particles and the chemical reaction time. The paper investigates the relationships between these quantities and their impact on the regolith conversion. Application of the model to various chemical reactor types, such as fluidized-bed, packed-bed, and rotary-bed configurations, are discussed.

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

Removal of sulphates acidity and iron from acid mine drainage in a bench scale biochemical treatment system.

PubMed

Prasad, D; Henry, J G

2009-02-01

The focus of this study was to develop a simple biochemical system to treat acid mine drainage for its safe disposal. Recovery and reuse of the metals removed were not considered. A three-step process for the treatment of acid mine drainage (AMD), proposed earlier, separates sulphate reducing activity from metal precipitation units and from a pH control system. Following our earlier work on the first step (biological reactor), this paper examines the second step (i.e. chemical reactor). The objectives of this study were: (1) to determine the increase in pH and the reduction of iron in the chemical reactor for different proportions of simulated AMD, and (2) to assess the capability of the chemical reactor. A series of experiments was conducted to study the effects of addition of alkaline sulphidogenic liquor (ASL) derived from a batch sulphidogenic biological reactor (operating with activated sludge and a COD/SO4 ratio of 1.6) on the simulated AMD characteristics. At 60-minute contact time, addition of 30% ASL (pH of 7.60-7.76) to the chemical reactor with 70% AMD (pH of 1.65-2.02), increased the pH of the AMD to 6.57 and alkalinity from 0 to 485 mg l(-1) as CaCO3, respectively and precipitated about 97% of the iron present in the simulated AMD. Others have demonstrated that metals in mine drainage can be precipitated by bacterial sulphate reduction. In this study, iron, a common and major component of mine drainage was used as a surrogate for metals in general. The results indicate the feasibility of treating AMD by an engineered sulphidogenic anaerobic reactor followed by a chemical reactor and that our three-step biochemical process has important advantages over other conventional AMD treatment systems.

Nuclear power in the 21st century: Challenges and possibilities.

PubMed

Horvath, Akos; Rachlew, Elisabeth

2016-01-01

The current situation and possible future developments for nuclear power--including fission and fusion processes--is presented. The fission nuclear power continues to be an essential part of the low-carbon electricity generation in the world for decades to come. There are breakthrough possibilities in the development of new generation nuclear reactors where the life-time of the nuclear waste can be reduced to some hundreds of years instead of the present time-scales of hundred thousand of years. Research on the fourth generation reactors is needed for the realisation of this development. For the fast nuclear reactors, a substantial research and development effort is required in many fields--from material sciences to safety demonstration--to attain the envisaged goals. Fusion provides a long-term vision for an efficient energy production. The fusion option for a nuclear reactor for efficient production of electricity has been set out in a focussed European programme including the international project of ITER after which a fusion electricity DEMO reactor is envisaged.

Heuristic optimization of a continuous flow point-of-use UV-LED disinfection reactor using computational fluid dynamics.

PubMed

Jenny, Richard M; Jasper, Micah N; Simmons, Otto D; Shatalov, Max; Ducoste, Joel J

2015-10-15

Alternative disinfection sources such as ultraviolet light (UV) are being pursued to inactivate pathogenic microorganisms such as Cryptosporidium and Giardia, while simultaneously reducing the risk of exposure to carcinogenic disinfection by-products (DBPs) in drinking water. UV-LEDs offer a UV disinfecting source that do not contain mercury, have the potential for long lifetimes, are robust, and have a high degree of design flexibility. However, the increased flexibility in design options will add a substantial level of complexity when developing a UV-LED reactor, particularly with regards to reactor shape, size, spatial orientation of light, and germicidal emission wavelength. Anticipating that LEDs are the future of UV disinfection, new methods are needed for designing such reactors. In this research study, the evaluation of a new design paradigm using a point-of-use UV-LED disinfection reactor has been performed. ModeFrontier, a numerical optimization platform, was coupled with COMSOL Multi-physics, a computational fluid dynamics (CFD) software package, to generate an optimized UV-LED continuous flow reactor. Three optimality conditions were considered: 1) single objective analysis minimizing input supply power while achieving at least (2.0) log10 inactivation of Escherichia coli ATCC 11229; and 2) two multi-objective analyses (one of which maximized the log10 inactivation of E. coli ATCC 11229 and minimized the supply power). All tests were completed at a flow rate of 109 mL/min and 92% UVT (measured at 254 nm). The numerical solution for the first objective was validated experimentally using biodosimetry. The optimal design predictions displayed good agreement with the experimental data and contained several non-intuitive features, particularly with the UV-LED spatial arrangement, where the lights were unevenly populated throughout the reactor. The optimal designs may not have been developed from experienced designers due to the increased degrees of freedom offered by using UV-LEDs. The results of this study revealed that the coupled optimization routine with CFD was effective at significantly decreasing the engineer's design decision space and finding a potentially near-optimal UV-LED reactor solution. Published by Elsevier Ltd.

Start-up of an anaerobic fluidized bed reactor treating synthetic carbohydrate rich wastewater.

PubMed

Yeshanew, Martha M; Frunzo, Luigi; Luongo, Vincenzo; Pirozzi, Francesco; Lens, Piet N L; Esposito, Giovanni

2016-12-15

The present work studied the start-up process of a mesophilic (37 ± 2 °C) anaerobic fluidized bed reactor (AFBR) operated at a hydraulic retention time (HRT) of 20 days using synthetic carbohydrate rich wastewater. Anox Kaldness-K1 carriers were used as biofilm carrier material. The reactor performance and biofilm formation were evaluated during the process. The start-up process at lower liquid recirculation flow rate enhanced the biofilm formation and reactor performance. The organic substrate composition had a major impact on early colonization of methanogenic archaea onto the surface of the Kaldness carriers during the start-up process. Specific organic substrates favouring the growth of methanogenic archaea, such as acetate, are preferred in order to facilitate the subsequent biofilm formation and AFBR start-up. The supply of 'bio-available' nutrients and trace elements, in particular iron, had an important role on optimal methanogenic activity and speeding-up of the biofilm development on the Kaldness carriers. This paper provides possible strategies to optimize the various operational parameters that influence the initial biofilm formation and development in an AFBR and similar high rate anaerobic reactors, hence can be used to reduce the long time required for process start-up. Copyright © 2016 Elsevier Ltd. All rights reserved.

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

CFD Simulation of flow pattern in a bubble column reactor for forming aerobic granules and its development.

PubMed

Fan, Wenwen; Yuan, LinJiang; Li, Yonglin

2018-06-22

The flow pattern is considered to play an important role in the formation of aerobic granular sludge in a bubble column reactor; therefore, it is necessary to understand the behavior of the flow in the reactor. A three-dimensional computational fluid dynamics (CFD) simulation for bubble column reactor was established to visualize the flow patterns of two-phase air-liquid flow and three-phase air-liquid-sludge flow under different ratios of height to diameter (H/D ratio) and superficial gas upflow velocities (SGVs). Moreover, a simulation of the three-phase flow pattern at the same SGV and different characteristics of the sludge was performed in this study. The results show that not only SGV but also properties of sludge involve the transformation of flow behaviors and relative velocity between liquid and sludge. For the original activated sludge floc to cultivate aerobic granules, the flow pattern has nothing to do with sludge, but is influenced by SGV, and the vortices is occurred and the relative velocity is increased with an increase in SGV; the two-phase flow can simplify the three-phase flow that predicts the flow pattern development in bubble column reactor (BCR) for aerobic granulation. For the aerobic granules, the liquid flow behavior developed from the symmetrical circular flow to numbers and small-size vortices with an increase in the sludge diameter, the relative velocity is amount up to u r  = 5.0, it is 29.4 times of original floc sludge.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, David I.; McClure, Patrick

2017-01-01

The development of NASAs Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, Dave I.; McClure, Patrick

2017-01-01

The development of NASA's Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

Molecular characterization of anaerobic sulfur-oxidizing microbial communities in up-flow anaerobic sludge blanket reactor treating municipal sewage.

PubMed

Aida, Azrina A; Hatamoto, Masashi; Yamamoto, Masamitsu; Ono, Shinya; Nakamura, Akinobu; Takahashi, Masanobu; Yamaguchi, Takashi

2014-11-01

A novel wastewater treatment system consisting of an up-flow anaerobic sludge blanket (UASB) reactor and a down-flow hanging sponge (DHS) reactor with sulfur-redox reaction was developed for treatment of municipal sewage under low-temperature conditions. In the UASB reactor, a novel phenomenon of anaerobic sulfur oxidation occurred in the absence of oxygen, nitrite and nitrate as electron acceptors. The microorganisms involved in anaerobic sulfur oxidation have not been elucidated. Therefore, in this study, we studied the microbial communities existing in the UASB reactor that probably enhanced anaerobic sulfur oxidation. Sludge samples collected from the UASB reactor before and after sulfur oxidation were used for cloning and terminal restriction fragment length polymorphism (T-RFLP) analysis of the 16S rRNA genes of the bacterial and archaeal domains. The microbial community structures of bacteria and archaea indicated that the genus Smithella and uncultured bacteria within the phylum Caldiserica were the dominant bacteria groups. Methanosaeta spp. was the dominant group of the domain archaea. The T-RFLP analysis, which was consistent with the cloning results, also yielded characteristic fingerprints for bacterial communities, whereas the archaeal community structure yielded stable microbial community. From these results, it can be presumed that these major bacteria groups, genus Smithella and uncultured bacteria within the phylum Caldiserica, probably play an important role in sulfur oxidation in UASB reactors. Copyright © 2014 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors

NASA Astrophysics Data System (ADS)

Weathered, Matthew Thomas

The thermal hydraulic properties of liquid sodium make it an attractive coolant for use in Generation IV reactors. The liquid metal's high thermal conductivity and low Prandtl number increases efficiency in heat transfer at fuel rods and heat exchangers, but can also cause features such as high magnitude temperature oscillations and gradients in the coolant. Currently, there exists a knowledge gap in the mechanisms which may create these features and their effect on mechanical structures in a sodium fast reactor. Two of these mechanisms include thermal striping and thermal stratification. Thermal striping is the oscillating temperature field created by the turbulent mixing of non-isothermal flows. Usually this occurs at the reactor core outlet or in piping junctions and can cause thermal fatigue in mechanical structures. Meanwhile, thermal stratification results from large volumes of non-isothermal sodium in a pool type reactor, usually caused by a loss of coolant flow accident. This stratification creates buoyancy driven flow transients and high temperature gradients which can also lead to thermal fatigue in reactor structures. In order to study these phenomena in sodium, a novel method for the deployment of optical fiber temperature sensors was developed. This method promotes rapid thermal response time and high spatial temperature resolution in the fluid. The thermal striping and stratification behavior in sodium may be experimentally analyzed with these sensors with greater fidelity than ever before. Thermal striping behavior at a junction of non-isothermal sodium was fully characterized with optical fibers. An experimental vessel was hydrodynamically scaled to model thermal stratification in a prototypical sodium reactor pool. Novel auxiliary applications of the optical fiber temperature sensors were developed throughout the course of this work. One such application includes local convection coefficient determination in a vessel with the corollary application of level sensing. Other applications were cross correlation velocimetry to determine bulk sodium flow rate and the characterization of coherent vortical structures in sodium with temperature frequency data. The data harvested, instrumentation developed and techniques refined in this work will help in the design of more robust reactors as well as validate computational models for licensing sodium fast reactors.

Application of two component biodegradable carriers in a particle-fixed biofilm airlift suspension reactor: development and structure of biofilms.

PubMed

Hille, Andrea; He, Mei; Ochmann, Clemens; Neu, Thomas R; Horn, Harald

2009-01-01

Two component biodegradable carriers for biofilm airlift suspension (BAS) reactors were investigated with respect to development of biofilm structure and oxygen transport inside the biofilm. The carriers were composed of PHB (polyhydroxybutyrate), which is easily degradable and PCL (caprolactone), which is less easily degradable by heterotrophic microorganisms. Cryosectioning combined with classical light microscopy and CLSM was used to identify the surface structure of the carrier material over a period of 250 days of biofilm cultivation in an airlift reactor. Pores of 50 to several hundred micrometers depth are formed due to the preferred degradation of PHB. Furthermore, microelectrode studies show the transport mechanism for different types of biofilm structures, which were generated under different substrate conditions. At high loading rates, the growth of a rather loosely structured biofilm with high penetration depths of oxygen was found. Strong changes of substrate concentration during fed-batch mode operation of the reactor enhance the growth of filamentous biofilms on the carriers. Mass transport in the outer regions of such biofilms was mainly driven by advection.

Multi-phase CFD modeling of solid sorbent carbon capture system

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

Ryan, E. M.; DeCroix, D.; Breault, R.

2013-07-01

Computational fluid dynamics (CFD) simulations are used to investigate a low temperature post-combustion carbon capture reactor. The CFD models are based on a small scale solid sorbent carbon capture reactor design from ADA-ES and Southern Company. The reactor is a fluidized bed design based on a silica-supported amine sorbent. CFD models using both Eulerian–Eulerian and Eulerian–Lagrangian multi-phase modeling methods are developed to investigate the hydrodynamics and adsorption of carbon dioxide in the reactor. Models developed in both FLUENT® and BARRACUDA are presented to explore the strengths and weaknesses of state of the art CFD codes for modeling multi-phase carbon capturemore » reactors. The results of the simulations show that the FLUENT® Eulerian–Lagrangian simulations (DDPM) are unstable for the given reactor design; while the BARRACUDA Eulerian–Lagrangian model is able to simulate the system given appropriate simplifying assumptions. FLUENT® Eulerian–Eulerian simulations also provide a stable solution for the carbon capture reactor given the appropriate simplifying assumptions.« less

Multi-Phase CFD Modeling of Solid Sorbent Carbon Capture System

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

Ryan, Emily M.; DeCroix, David; Breault, Ronald W.

2013-07-30

Computational fluid dynamics (CFD) simulations are used to investigate a low temperature post-combustion carbon capture reactor. The CFD models are based on a small scale solid sorbent carbon capture reactor design from ADA-ES and Southern Company. The reactor is a fluidized bed design based on a silica-supported amine sorbent. CFD models using both Eulerian-Eulerian and Eulerian-Lagrangian multi-phase modeling methods are developed to investigate the hydrodynamics and adsorption of carbon dioxide in the reactor. Models developed in both FLUENT® and BARRACUDA are presented to explore the strengths and weaknesses of state of the art CFD codes for modeling multi-phase carbon capturemore » reactors. The results of the simulations show that the FLUENT® Eulerian-Lagrangian simulations (DDPM) are unstable for the given reactor design; while the BARRACUDA Eulerian-Lagrangian model is able to simulate the system given appropriate simplifying assumptions. FLUENT® Eulerian-Eulerian simulations also provide a stable solution for the carbon capture reactor given the appropriate simplifying assumptions.« less

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Improving High-Temperature Measurements in Nuclear Reactors with Mo/Nb Thermocouples

NASA Astrophysics Data System (ADS)

Villard, J.-F.; Fourrez, S.; Fourmentel, D.; Legrand, A.

2008-10-01

Many irradiation experiments performed in research reactors are used to assess the effects of nuclear radiations on material or fuel sample properties, and are therefore a crucial stage in most qualification and innovation studies regarding nuclear technologies. However, monitoring these experiments requires accurate and reliable instrumentation. Among all measurement systems implemented in irradiation devices, temperature—and more particularly high-temperature (above 1000°C)—is a major parameter for future experiments related, for example, to the Generation IV International Forum (GIF) Program or the International Thermonuclear Experimental Reactor (ITER) Project. In this context, the French Commissariat à l’Energie Atomique (CEA) develops and qualifies innovative in-pile instrumentation for its irradiation experiments in current and future research reactors. Logically, a significant part of these research and development programs concerns the improvement of in-pile high-temperature measurements. This article describes the development and qualification of innovative high-temperature thermocouples specifically designed for in-pile applications. This key study has been achieved with technical contributions from the Thermocoax Company. This new kind of thermocouple is based on molybdenum and niobium thermoelements, which remain nearly unchanged by thermal neutron flux even under harsh nuclear environments, whereas typical high-temperature thermocouples such as Type C or Type S are altered by significant drifts caused by material transmutations under the same conditions. This improvement has a significant impact on the temperature measurement capabilities for future irradiation experiments. Details of the successive stages of this development are given, including the results of prototype qualification tests and the manufacturing process.

Design, scale-up, Six Sigma in processing different feedstocks in a fixed bed downdraft biomass gasifier

NASA Astrophysics Data System (ADS)

Boravelli, Sai Chandra Teja

This thesis mainly focuses on design and process development of a downdraft biomass gasification processes. The objective is to develop a gasifier and process of gasification for a continuous steady state process. A lab scale downdraft gasifier was designed to develop the process and obtain optimum operating procedure. Sustainable and dependable sources such as biomass are potential sources of renewable energy and have a reasonable motivation to be used in developing a small scale energy production plant for countries such as Canada where wood stocks are more reliable sources than fossil fuels. This thesis addresses the process of thermal conversion of biomass gasification process in a downdraft reactor. Downdraft biomass gasifiers are relatively cheap and easy to operate because of their design. We constructed a simple biomass gasifier to study the steady state process for different sizes of the reactor. The experimental part of this investigation look at how operating conditions such as feed rate, air flow, the length of the bed, the vibration of the reactor, height and density of syngas flame in combustion flare changes for different sizes of the reactor. These experimental results also compare the trends of tar, char and syngas production for wood pellets in a steady state process. This study also includes biomass gasification process for different wood feedstocks. It compares how shape, size and moisture content of different feedstocks makes a difference in operating conditions for the gasification process. For this, Six Sigma DMAIC techniques were used to analyze and understand how each feedstock makes a significant impact on the process.

Electrogenerative gold recovery from cyanide solutions using a flow-through cell with activated reticulated vitreous carbon.

PubMed

Yap, Chin Yean; Mohamed, Norita

2008-10-01

An electrogenerative flow-through reactor with an activated reticulated vitreous carbon cathode was developed. The influence of palladium-tin activation of the cathode towards gold deposition was studied by cyclic voltammetry. The reactor proved to be efficient in recovering more than 99% of gold within 4 h of operation. The performance of the reactor was evaluated with initial gold concentrations of 10, 100 and 500 mg L-1 and various electrolyte flow rates. Gold recovery was found to be strongly dependent on electrolyte flow rate and initial gold concentration in the cyanide solution under the experimental conditions used.

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for a loss of offsite power initiator using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following amore » loss of offsite power.« less

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

Steiner, J.L.; Harmony, S.C.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS Supplement design were performed for a large-break loss-of-coolant (LBLOCA) initiator using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following anmore » LBLOCA.« less

Developing the European Center of Competence on VVER-type nuclear power reactors

NASA Astrophysics Data System (ADS)

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-09-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for VVER-technology is focused on master's degree programmes. The specifics of a systematic approach to training in the area of VVER-type nuclear power reactors technologies are analysed. This paper discusses enhancement of the training opportunities of the European Center that have arisen from advances in methodology and distance education. With a special attention paid to the European Nuclear Education Network (ENEN), the possibilities of further development of the international cooperation between European countries and educational institutions are examined.

How much does a tokamak reactor cost?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

2017-10-01

The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

Preliminary Analysis of High-Flux RSG-GAS to Transmute Am-241 of PWR’s Spent Fuel in Asian Region

NASA Astrophysics Data System (ADS)

Budi Setiawan, M.; Kuntjoro, S.

2018-02-01

A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-GAS research reactor was performed, aiming at an optimal transmutation loading for present nuclear energy development. The MA selected in the analysis includes Am-241 discharged from pressurized water reactors (PWRs) in Asian region. Until recently, studies have been undertaken in various methods to reduce radiotoxicity from actinides in high-level waste. From the cell calculation using computer code SRAC2006, it is obtained that the target Am-241 which has a cross section of the thermal energy absorption in the region (group 8) is relatively large; it will be easily burned in the RSG-GAS reactor. Minor actinides of Am-241 which can be inserted in the fuel (B/T fuel) is 2.5 kg which is equivalent to Am-241 resulted from the partition of spent fuel from 2 units power reactors PWR with power 1000MW(th) operated for one year.

Integration of a photocatalytic multi-tube reactor for indoor air purification in HVAC systems: a feasibility study.

PubMed

van Walsem, Jeroen; Roegiers, Jelle; Modde, Bart; Lenaerts, Silvia; Denys, Siegfried

2018-04-24

This work is focused on an in-depth experimental characterization of multi-tube reactors for indoor air purification integrated in ventilation systems. Glass tubes were selected as an excellent photocatalyst substrate to meet the challenging requirements of the operating conditions in a ventilation system in which high flow rates are typical. Glass tubes show a low-pressure drop which reduces the energy demand of the ventilator, and additionally, they provide a large exposed surface area to allow interaction between indoor air contaminants and the photocatalyst. Furthermore, the performance of a range of P25-loaded sol-gel coatings was investigated, based on their adhesion properties and photocatalytic activities. Moreover, the UV light transmission and photocatalytic reactor performance under various operating conditions were studied. These results provide vital insights for the further development and scaling up of multi-tube reactors in ventilation systems which can provide a better comfort, improved air quality in indoor environments, and reduced human exposure to harmful pollutants.

ANALYTICAL CHEMISTRY DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1961

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

None

1962-02-01

Research and development progress is reported on analytlcal instrumentation, dlssolver-solution analyses, special research problems, reactor projects analyses, x-ray and spectrochemical analyses, mass spectrometry, optical and electron microscopy, radiochemical analyses, nuclear analyses, inorganic preparations, organic preparations, ionic analyses, infrared spectral studies, anodization of sector coils for the Analog II Cyclotron, quality control, process analyses, and the Thermal Breeder Reactor Projects Analytical Chemistry Laboratory. (M.C.G.)

Rapid solar-thermal decarbonization of methane

NASA Astrophysics Data System (ADS)

Dahl, Jaimee Kristen

Due to the ever-increasing demand for energy and the concern over the environmental impact of continuing to produce energy using current methods, there is interest in developing a hydrogen economy. Hydrogen is a desirable energy source because it is abundant in nature and burns cleanly. One method for producing hydrogen is to utilize a renewable energy source to obtain high enough temperatures to decompose a fossil fuel into its elements. This thesis work is directed at developing a solar-thermal aerosol flow reactor to dissociate methane to carbon black and hydrogen. The technology is intended as a "bridge" between current hydrogen production methods, such as conventional steam-methane reformers, and future "zero emission" technology for producing hydrogen, such as dissociating water using a renewable heating source. A solar furnace is used to heat a reactor to temperatures in excess of 2000 K. The final reactor design studied consists of three concentric vertical tubes---an outer quartz protection tube, a middle solid graphite heating tube, and an inner porous graphite reaction tube. A "fluid-wall" is created on the inside wall of the porous reaction tube in order to prevent deposition of the carbon black co-product on the reactor tube wall. The amorphous carbon black produced aids in heating the gas stream by absorbing radiation from the reactor wall. Conversions of 90% are obtained at a reactor wall temperature of 2100 K and an average residence time of 0.01 s. Computer modeling is also performed to study the gas flow and temperature profiles in the reactor as well as the kinetics of the methane dissociation reaction. The simulations indicate that there is little flow of the fluid-wall gas through the porous wall in the hot zone region, but this can be remedied by increasing the inlet temperature of the fluid-wall gas and/or increasing the tube permeability only in the hot zone region of the wall. The following expression describes the kinetics of methane dissociation in a solar-thermal fluid-wall reactor: dXdt=5.8x108 exp-155,600RT 1-X 7.2s-1. The experimental and theoretical work reported in this thesis is the groundwork that will be utilized in scaling up the reactor to produce hydrogen in distributed or centralized facilities.

Strategies for the startup of methanogenic inverse fluidized-bed reactors using colonized particles.

PubMed

Alvarado-Lassman, A; Sandoval-Ramos, A; Flores-Altamirano, M G; Vallejo-Cantú, N A; Méndez-Contreras, J M

2010-05-01

One of the inconveniences in the startup of methanogenic inverse fluidized-bed reactors (IFBRs) is the long period required for biofilm formation and stabilization of the system. Previous researchers have preferred to start up in batch mode to shorten stabilization times. Much less work has been done with continuous-mode startup for the IFBR configuration of reactors. In this study, we prepared two IFBRs with similar characteristics to compare startup times for batch- and continuous-operation modes. The reactors were inoculated with a small quantity of colonized particles and run for a period of 3 months, to establish the optimal startup strategy using synthetic media as a substrate (glucose as a source of carbon). After the startup stage, the continuous- and batch-mode reactors removed more than 80% of the chemical oxygen demand (COD) in 51 and 60 days of operation, respectively; however, at the end of the experiments, the continuous-mode reactor had more biomass attached to the support media than the batch-mode reactor. Both reactors developed fully covered support media, but only the continuous-mode reactor had methane yields close to the theoretical value that is typical of stable reactors. Then, a combined startup strategy was proposed, with industrial wastewater as the substrate, using a sequence of batch cycles followed by continuous operation, which allows stable operation at an organic loading rate of 20 g COD/L x d in 15 days. Using a fraction of colonized support as an inoculum presents advantages, with respect to previously reported strategies.

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

Preliminary design studies on a nuclear seawater desalination system

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

Wibisono, A. F.; Jung, Y. H.; Choi, J.

2012-07-01

Seawater desalination is one of the most promising technologies to provide fresh water especially in the arid region. The most used technology in seawater desalination are thermal desalination (MSF and MED) and membrane desalination (RO). Some developments have been done in the area of coupling the desalination plant with a nuclear reactor to reduce the cost of energy required in thermal desalination. The coupling a nuclear reactor to a desalination plant can be done either by using the co-generation or by using dedicated heat from a nuclear system. The comparison of the co-generation nuclear reactor with desalination plant, dedicated nuclearmore » heat system, and fossil fueled system will be discussed in this paper using economical assessment with IAEA DEEP software. A newly designed nuclear system dedicated for the seawater desalination will also be suggested by KAIST (Korea Advanced Inst. of Science and Technology) research team and described in detail within this paper. The suggested reactor system is using gas cooled type reactor and in this preliminary study the scope of design will be limited to comparison of two cases in different operating temperature ranges. (authors)« less

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

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

M. L. Grossbeck J-P.A. Renier Tim Bigelow

2003-09-30

Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility andmore » cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.« less

Experimental study on the instability of Pressure Balance Injection System (PBIS)

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

Okamoto, Koji; Teshima, Hideyuki; Madarame, Haruki

1996-06-01

The Passive Safety Reactor has been developed to reduce the construction cost and to improve the safety. Japan Atomic Energy Research institute (JAERI) proposed the System-Integrated Pressurized Water Reactor (SPWR) as a Passive Safety Reactor. In the SPWR design, the Pressure Balanced Injection System (PBIS) was introduced for the passive safety concept. The water with boron in a containment vessel were passively injected into the core by the pressure difference between the containment vessel and reactor vessel at a severe accidental condition. However there are few studies on the thermo-hydraulic characteristics of the PBIS. In this study, the thermal hydraulicsmore » of the PBIS are experimentally investigated using the small scale model. The instability of the injected flow was observed in the adiabatic experiment. The instability was caused by the pressure balance between the two vessels. The mechanism of the instability are discussed, resulting in the good agreement with the experimental results. In the steam experiment, another instability was observed, which was caused by the heat balance in the main tank.« less

CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0

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

Toth, Sandor; Legradi, Gabor; Aszodi, Attila

2006-07-01

From the aspect of planning the power upgrading of nuclear reactors - including the VVER-440 type reactor - it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960more » mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper. (authors)« less

Development and application of kinetic model on biological anoxic/aerobic filter.

PubMed

Kim, Youngnoh; Tanaka, Kazuhiro; Lee, Yong-Woo; Chung, Jinwook

2008-01-01

An up-flow biological anoxic filter (BANF) has been developed to achieve high removal performance of suspended solids and BOD removal as well as nitrogen. With a view to understand treatment mechanisms, we developed a filtration model that incorporates filtration, deposit scoring and biological reactions simultaneously. The biological reactions consist of four types of reaction; dissolution of organic particles; utilization of dissolved organic matter; denitrification; and self-degradation of bacteria. Whereas the reactor is generally assumed to be a plug flow reactor in the filtration model, it is assumed a continuous-flow stirred tank reactor (CSTR) in the model of biological reactions. The hydrodynamics is supposed that the filter bottom (the portion sludge settled) is a CSTR and the filter bed (the portion filled with filter media) consists of number of CSTR of equal size arranged in series. The model obtained in this study was verified and simulated using experimental results taken from a pilot-scale plant and predicted the experimental data well, applying to design and operate BANF.

Fusion Studies in Japan

NASA Astrophysics Data System (ADS)

Ogawa, Yuichi

2016-05-01

A new strategic energy plan decided by the Japanese Cabinet in 2014 strongly supports the steady promotion of nuclear fusion development activities, including the ITER project and the Broader Approach activities from the long-term viewpoint. Atomic Energy Commission (AEC) in Japan formulated the Third Phase Basic Program so as to promote an experimental fusion reactor project. In 2005 AEC has reviewed this Program, and discussed on selection and concentration among many projects of fusion reactor development. In addition to the promotion of ITER project, advanced tokamak research by JT-60SA, helical plasma experiment by LHD, FIREX project in laser fusion research and fusion engineering by IFMIF were highly prioritized. Although the basic concept is quite different between tokamak, helical and laser fusion researches, there exist a lot of common features such as plasma physics on 3-D magnetic geometry, high power heat load on plasma facing component and so on. Therefore, a synergetic scenario on fusion reactor development among various plasma confinement concepts would be important.

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

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

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

Bernard, J.A.

1989-09-01

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power onmore » spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.« less

A Performance-Based Training Qualification Guide/Checklist Developed for Reactor Operators at the High Flux Beam Reactor at Brookhaven National Laboratory.

ERIC Educational Resources Information Center

McNair, Robert C.

A Performance-Based Training (PBT) Qualification Guide/Checklist was developed that would enable a trainee to attain the skills, knowledge, and attitude required to operate the High Flux Beam Reactor at Brookhaven National Laboratory. Design of this guide/checklist was based on the Instructional System Design Model. The needs analysis identified…

Closed Brayton cycle power conversion systems for nuclear reactors :

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

Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.

2006-04-01

This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors,more » reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.« less

Technology development for iron Fischer-Tropsch catalysts. Technical progress report No. 8, July 1, 1992--September 30, 1992

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

Frame, R.R.; Gala, H.B.

1992-12-31

The objectives of this contract are to develop a technology for the production of active and stable iron Fischer-Tropsch catalysts for use in slurry-phase synthesis reactors and to develop a scaleup procedure for large-scale synthesis of such catalysts for process development and long-term testing in slurry bubble-column reactors. With a feed containing hydrogen and carbon monoxide in the molar ratio of 0.5 to 1.0 to the slurry bubble-column reactor, the catalyst performance target is 88% CO + H{sub 2} conversion at a minimum space velocity of 2.4 NL/hr/gFe. The desired sum of methane and ethane selectivities is no more thanmore » 4%, and the conversion loss per week is not to exceed 1%. Contract Tasks are as follows: 1.0--Catalyst development, 1.1--Technology assessment, 1.2--Precipitated catalyst preparation method development, 1.3--Novel catalyst preparation methods investigation, 1.4--Catalyst pretreatment, 1.5--Catalyst characterization, 2.0--Catalyst testing, 3.0--Catalyst aging studies, and 4.0--Preliminary design and cost estimate of a catalyst synthesis facility. This paper reports progress made on Task 1.« less

DEVELOPMENT AND APPLICATION OF MATERIALS PROPERTIES FOR FLAW STABILITY ANALYSIS IN EXTREME ENVIRONMENT SERVICE

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

Sindelar, R; Ps Lam, P; Andrew Duncan, A

Discovery of aging phenomena in the materials of a structure may arise after its design and construction that impact its structural integrity. This condition can be addressed through a demonstration of integrity with the material-specific degraded conditions. Two case studies of development of fracture and crack growth property data, and their application in development of in-service inspection programs for nuclear structures in the defense complex are presented. The first case study covers the development of fracture toughness properties in the form of J-R curves for rolled plate Type 304 stainless steel with Type 308 stainless steel filler in the applicationmore » to demonstrate the integrity of the reactor tanks of the heavy water production reactors at the Savannah River Site. The fracture properties for the base, weld, and heat-affected zone of the weldments irradiated at low temperatures (110-150 C) up to 6.4 dpa{sub NRT} and 275 appm helium were developed. An expert group provided consensus for application of the irradiated properties for material input to acceptance criteria for ultrasonic examination of the reactor tanks. Dr. Spencer H. Bush played a lead advisory role in this work. The second case study covers the development of fracture toughness for A285 carbon steel in high level radioactive waste tanks. The approach in this case study incorporated a statistical experimental design for material testing to address metallurgical factors important to fracture toughness. Tolerance intervals were constructed to identify the lower bound fracture toughness for material input to flaw disposition through acceptance by analysis.« less

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

Wichman, K.; Tsao, J.; Mayfield, M.

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRCmore » and the International Piping Integrity Research Group is also briefly summarized.« less

NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

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

Cynthia D. Gentillon

2010-09-01

Projects for the Very High Temperature Reactor Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the very high temperature reactor. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high-temperature and high-fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The Very High Temperature Reactor Technology Development Office has established the NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory to ensure that very high temperature reactor data are (1) qualified for use, (2) stored in amore » readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and for data analysis to identify useful relationships among the measured quantities.« 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

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2010-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2011-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2012-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

Reference Reactor Module for the Affordable Fission Surface Power System

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

2008-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO2-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important ``affordability'' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

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

Hale, Richard Edward; Cetiner, Sacit M.; Fugate, David L.

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environmentmore » and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.« less

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

NASA Astrophysics Data System (ADS)

Berg, Thomas A.; Disney, Richard K.

2004-02-01

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs.

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

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

Berg, Thomas A.; Disney, Richard K.

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs.

Development of RF plasma simulations of in-reactor tests of small models of the nuclear light bulb fuel region

NASA Technical Reports Server (NTRS)

Roman, W. C.; Jaminet, J. F.

1972-01-01

Experiments were conducted to develop test configurations and technology necessary to simulate the thermal environment and fuel region expected to exist in in-reactor tests of small models of nuclear light bulb configurations. Particular emphasis was directed at rf plasma tests of approximately full-scale models of an in-reactor cell suitable for tests in Los Alamos Scientific Laboratory's Nuclear Furnace. The in-reactor tests will involve vortex-stabilized fissioning uranium plasmas of approximately 200-kW power, 500-atm pressure and equivalent black-body radiating temperatures between 3220 and 3510 K.

Advanced reactors and novel reactions for the conversion of triglyceride based oils into high quality renewable transportation fuels

NASA Astrophysics Data System (ADS)

Linnen, Michael James

Sustainable energy continues to grow more important to all societies, leading to the research and development of a variety of alternative and renewable energy technologies. Of these, renewable liquid transportation fuels may be the most visible to consumers, and this visibility is further magnified by the long-term trend of increasingly expensive petroleum fuels that the public consumes. While first-generation biofuels such as biodiesel and fuel ethanol have been integrated into the existing fuel infrastructures of several countries, the chemical differences between them and their petroleum counterparts reduce their effectiveness. This gives rise to the development and commercialization of second generation biofuels, many of which are intended to have equivalent properties to those of their petroleum counterparts. In this dissertation, the primary reactions for a second-generation biofuel process, known herein as the University of North Dakota noncatalytic cracking process (NCP), have been studied at the fundamental level and improved. The NCP is capable of producing renewable fuels and chemicals that are virtually the same as their petroleum counterparts in performance and quality (i.e., petroleum-equivalent). In addition, a novel analytical method, FIMSDIST was developed which, within certain limitations, can increase the elution capabilities of GC analysis and decrease sample processing times compared to other high resolution methods. These advances are particularly useful for studies of highly heterogeneous fuel and/or organic chemical intermediates, such as those studied for the NCP. However the data from FIMSDIST must be supplemented with data from other methods such as for certain carboxylic acid, to provide accurate, comprehensive results, From a series of TAG cracking experiments that were performed, it was found that coke formation during cracking is most likely the result of excessive temperature and/or residence time in a cracking reactor. Based on this, a tubular cracking reactor was developed that could operate continuously without coke formation. The design also was proven to be scalable. Yields from the reactor were determined under a variety of conditions in order to predict the outputs from the NCP and to establish relationships/correlations between operating parameters and the product distribution. These studies led to the conclusion that the most severe operating conditions which do not induce coking are optimal over the experimental domain. In order to develop economical deoxygenation catalysts for use within the NCP, a series of experiments were performed using nickel catalysts, demonstrating that nickel catalysts could outperform their predecessor, a high cost palladium-based catalyst. A nickel catalyst was then tested in a packed bed reactor in order to determine suitable operating conditions for its commercial utilization in packed bed reactors.

Development of a trickle bed reactor of electro-Fenton process for wastewater treatment.

PubMed

Lei, Yangming; Liu, Hong; Shen, Zhemin; Wang, Wenhua

2013-10-15

To avoid electrolyte leakage and gas bubbles in the electro-Fenton (E-Fenton) reactors using a gas diffusion cathode, we developed a trickle bed cathode by coating a layer composed of carbon black and polytetrafluoroethylene (C-PTFE) onto graphite chips instead of carbon cloth. The trickle bed cathode was optimized by single-factor and orthogonal experiments, in which carbon black, PTFE, and a surfactant were considered as the determinant of the performance of graphite chips. In the reactor assembled by the trickle bed cathode, H2O2 was generated with a current of 0.3A and a current efficiency of 60%. This performance was attributed to the fine distribution of electrolyte and air, as well as the effective oxygen transfer from the gas phase to the electrolyte-cathode interface. In terms of H2O2 generation and current efficiency, the developed trickle bed reactor had a performance comparable to that of the conventional E-Fenton reactor using a gas diffusion cathode. Further, 123 mg L(-1) of reactive brilliant red X-3B in aqueous solution was decomposed in the optimized trickle bed reactor as E-Fenton reactor. The decolorization ratio reached 97% within 20 min, and the mineralization reached 87% within 3h. Copyright © 2013 Elsevier B.V. All rights reserved.

Osiris and SOMBRERO inertial confinement fusion power plant designs. Volume 2, Designs, assessments, and comparisons, Final report

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

Meier, W.R.; Bieri, R.L.; Monsler, M.J.

1992-03-01

The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of ourmore » effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs.« less

Modelling biological Cr(VI) reduction in aquifer microcosm column systems.

PubMed

Molokwane, Pulane E; Chirwa, Evans M N

2013-01-01

Several chrome processing facilities in South Africa release hexavalent chromium (Cr(VI)) into groundwater resources. Pump-and-treat remediation processes have been implemented at some of the sites but have not been successful in reducing contamination levels. The current study is aimed at developing an environmentally friendly, cost-effective and self-sustained biological method to curb the spread of chromium at the contaminated sites. An indigenous Cr(VI)-reducing mixed culture of bacteria was demonstrated to reduce high levels of Cr(VI) in laboratory samples. The effect of Cr(VI) on the removal rate was evaluated at concentrations up to 400 mg/L. Following the detailed evaluation of fundamental processes for biological Cr(VI) reduction, a predictive model for Cr(VI) breakthrough through aquifer microcosm reactors was developed. The reaction rate in batch followed non-competitive rate kinetics with a Cr(VI) inhibition threshold concentration of approximately 99 mg/L. This study evaluates the application of the kinetic parameters determined in the batch reactors to the continuous flow process. The model developed from advection-reaction rate kinetics in a porous media fitted best the effluent Cr(VI) concentration. The model was also used to elucidate the logistic nature of biomass growth in the reactor systems.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

NASA Astrophysics Data System (ADS)

Scarlat, Raluca O.; Peterson, Per F.

2014-01-01

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

Non-destructive research methods applied on materials for the new generation of nuclear reactors

NASA Astrophysics Data System (ADS)

Bartošová, I.; Slugeň, V.; Veterníková, J.; Sojak, S.; Petriska, M.; Bouhaddane, A.

2014-06-01

The paper is aimed on non-destructive experimental techniques applied on materials for the new generation of nuclear reactors (GEN IV). With the development of these reactors, also materials have to be developed in order to guarantee high standard properties needed for construction. These properties are high temperature resistance, radiation resistance and resistance to other negative effects. Nevertheless the changes in their mechanical properties should be only minimal. Materials, that fulfil these requirements, are analysed in this work. The ferritic-martensitic (FM) steels and ODS steels are studied in details. Microstructural defects, which can occur in structural materials and can be also accumulated during irradiation due to neutron flux or alpha, beta and gamma radiation, were analysed using different spectroscopic methods as positron annihilation spectroscopy and Barkhausen noise, which were applied for measurements of three different FM steels (T91, P91 and E97) as well as one ODS steel (ODS Eurofer).

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

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

Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A

2012-09-01

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin nextmore » year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.« less

Modeling and analysis of tritium dynamics in a DT fusion fuel cycle

NASA Astrophysics Data System (ADS)

Kuan, William

1998-11-01

A number of crucial design issues have a profound effect on the dynamics of the tritium fuel cycle in a DT fusion reactor, where the development of appropriate solutions to these issues is of particular importance to the introduction of fusion as a commercial system. Such tritium-related issues can be classified according to their operational, safety, and economic impact to the operation of the reactor during its lifetime. Given such key design issues inherent in next generation fusion devices using the DT fuel cycle development of appropriate models can then lead to optimized designs of the fusion fuel cycle for different types of DT fusion reactors. In this work, two different types of modeling approaches are developed and their application to solving key tritium issues presented. For the first approach, time-dependent inventories, concentrations, and flow rates characterizing the main subsystems of the fuel cycle are simulated with a new dynamic modular model of a fusion reactor's fuel cycle, named X-TRUFFLES (X-Windows TRitiUm Fusion Fuel cycLE dynamic Simulation). The complex dynamic behavior of the recycled fuel within each of the modeled subsystems is investigated using this new integrated model for different reactor scenarios and design approaches. Results for a proposed fuel cycle design taking into account current technologies are presented, including sensitivity studies. Ways to minimize the tritium inventory are also assessed by examining various design options that could be used to minimize local and global tritium inventories. The second modeling approach involves an analytical model to be used for the calculation of the required tritium breeding ratio, i.e., a primary design issue which relates directly to the feasibility and economics of DT fusion systems. A time-integrated global tritium balance scheme is developed and appropriate analytical expressions are derived for tritium self-sufficiency relevant parameters. The easy exploration of the large parameter space of the fusion fuel cycle can thus be conducted as opposed to previous modeling approaches. Future guidance for R&D (research and development) in fusion nuclear technology is discussed in view of possible routes to take in reducing the tritium breeding requirements of DT fusion reactors.

On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

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

Chauvin, J.P.; Blaise, P.; Lyoussi, A.

2015-07-01

The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physicsmore » calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)« less

A study of the Coriolis effect on the fluid flow profile in a centrifugal bioreactor.

PubMed

Detzel, Christopher J; Thorson, Michael R; Van Wie, Bernard J; Ivory, Cornelius F

2009-01-01

Increasing demand for tissues, proteins, and antibodies derived from cell culture is necessitating the development and implementation of high cell density bioreactors. A system for studying high density culture is the centrifugal bioreactor (CCBR), which retains cells by increasing settling velocities through system rotation, thereby eliminating diffusional limitations associated with mechanical cell retention devices. This article focuses on the fluid mechanics of the CCBR system by considering Coriolis effects. Such considerations for centrifugal bioprocessing have heretofore been ignored; therefore, a simpler analysis of an empty chamber will be performed. Comparisons are made between numerical simulations and bromophenol blue dye injection experiments. For the non-rotating bioreactor with an inlet velocity of 4.3 cm/s, both the numerical and experimental results show the formation of a teardrop shaped plume of dye following streamlines through the reactor. However, as the reactor is rotated, the simulation predicts the development of vortices and a flow profile dominated by Coriolis forces resulting in the majority of flow up the leading wall of the reactor as dye initially enters the chamber, results are confirmed by experimental observations. As the reactor continues to fill with dye, the simulation predicts dye movement up both walls while experimental observations show the reactor fills with dye from the exit to the inlet. Differences between the simulation and experimental observations can be explained by excessive diffusion required for simulation convergence, and a slight density difference between dyed and un-dyed solutions. Implications of the results on practical bioreactor use are also discussed. (c) 2009 American Institute of Chemical Engineers Biotechnol. Prog., 2009.

A Study of the Coriolis Effect on the Fluid Flow Profile in a Centrifugal Bioreactor

PubMed Central

Detzel, Christopher J.; Thorson, Michael R.; Van Wie, Bernard J.; Ivory, Cornelius F.

2011-01-01

Increasing demand for tissues, proteins, and antibodies derived from cell culture is necessitating the development and implementation of high cell density bioreactors. A system for studying high density culture is the centrifugal bioreactor (CCBR) which retains cells by increasing settling velocities through system rotation, thereby eliminating diffusional limitations associated with mechanical cell retention devices. This paper focuses on the fluid mechanics of the CCBR system by considering Coriolis effects. Such considerations for centrifugal bioprocessing have heretofore been ignored; therefore a simpler analysis of an empty chamber will be performed. Comparisons are made between numerical simulations and bromophenol blue dye injection experiments. For the non-rotating bioreactor with an inlet velocity of 4.3 cm/s, both the numerical and experimental results show the formation of a teardrop shaped plume of dye following streamlines through the reactor. However, as the reactor is rotated the simulation predicts the development of vortices and a flow profile dominated by Coriolis forces resulting in the majority of flow up the leading wall of the reactor as dye initially enters the chamber, results confirmed by experimental observations. As the reactor continues to fill with dye, the simulation predicts dye movement up both walls while experimental observations show the reactor fills with dye from the exit to the inlet. Differences between the simulation and experimental observations can be explained by excessive diffusion required for simulation convergence, and a slight density difference between dyed and un-dyed solutions. Implications of the results on practical bioreactor use are also discussed. PMID:19455639

Significance of breeding in fast nuclear reactors

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

Raza, S.M.; Abidi, S.B.M.

1983-12-01

Only breeder reactors--nuclear power plants that produce more fuel than they consume--are capable in principle of extracting the maximum amount of fission energy contained in uranium ore, thus offering a practical long-term solution to uranium supply problems. Uranium would then constitute a virtually inexhaustible fuel reserve for the world's future energy needs. The ultimate argument for breeding is to conserve the energy resources available to mankind. A long-term role for nuclear power with fast reactors is proven to be economically viable, environmentally acceptable and capable of wide scale exploitation in many countries. In this paper, various suggestions pertaining to themore » fuel fabrication route, fuel cycle economics, studies of the physics of fast nuclear reactors and of engineering design simplifications are presented. Fast reactors contain no moderator and inherently require enriched fuel. In general, the main aim is to suggest an improvement in the understanding of the safety and control characteristics of fast breeder power reactors. Development work is also being devoted to new carbide and nitride fuels, which are likely to exhibit breeding characteristics superior to those of the oxides of plutonium and uranium.« less

Phase 1 Space Fission Propulsion Energy Source Design

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems with a specific mass at or below 50 kg/kWjet could enhance or enable numerous robotic outer solar system missions of interest. At the required specific mass, it is possible to develop safe, affordable systems that meet mission requirements. To help select the system design to pursue, eight evaluation criteria were identified: system integration, safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of four potential concepts was performed: a Testable, Passive, Redundant Reactor (TPRR), a Testable Multi-Cell In-Core Thermionic Reactor (TMCT), a Direct Gas Cooled Reactor (DGCR), and a Pumped Liquid Metal Reactor.(PLMR). Development of any of the four systems appears feasible. However, for power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the TPRR has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the TPRR approach. Successful development and utilization of a "Phase I" fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system.

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

Tan, Lizhen; Yang, Ying; Tyburska-Puschel, Beata

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools)more » is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.« less

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

Asquith, J. G.; Mason, D. G.; Stamp, S.

1972-01-01

The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

A Review of Study on Thermal Energy Transport System by Synthesis and Decomposition Reactions of Methanol

NASA Astrophysics Data System (ADS)

Liu, Qiusheng; Yabe, Akira; Kajiyama, Shiro; Fukuda, Katsuya

The study on thermal energy transport system by synthesis and decomposition reactions of methanol was reviewed. To promote energy conservation and global environment protection, a two-step liquid-phase methanol synthesis process, which starts with carbonylation of methanol to methyl formate, then followed by the hydrogenolysis of the formate, was studied to recover wasted or unused discharged heat from industrial sources for the thermal energy demands of residential and commercial areas by chemical reactions. The research and development of the system were focused on the following three points. (1) Development of low-temperature decomposition and synthetic catalysts, (2) Development of liquid phase reactor (heat exchanger accompanying chemical reaction), (3) Simulation of the energy transport efficiency of entire system which contains heat recovery and supply sections. As the result of the development of catalyst, promising catalysts which agree with the development purposes for the methyl formate decomposition reaction and the synthetic reaction are being developed though some studies remain for the methanol decomposition and synthetic reactions. In the fundamental development of liquid phase reactor, the solubilities of CO and H2 gases in methanol and methyl formate were measured by the method of total pressure decrease due to absorption under pressures up to 1500kPa and temperatures up to 140°C. The diffusivity of CO gas in methanol was determined by measuring the diameter and solution time of single CO bubbles in methanol. The chemical reaction rate of methanol synthesis by hydrogenolysis of methyl formate was measured using a plate-type of Raney copper catalyst in a reactor with rectangular channel and in an autoclave reactor. The reaction characteristics were investigated by carrying out the experiments at various temperatures, flow rates and at various catalyst development conditions. We focused on the effect of Raney copper catalyst thickness on the liquid-phase chemical reaction by varying the development time of the catalyst. Investigation results of the catalyst such as surface area, pore radius, lattice size, and photographs of scanning electron microscope (SEM) were also given. In the simulation of energy transport efficiency of this system, by simulating the energy transfer system using two-step liquid phase methanol decomposition and synthetic reactions, and comparing with the technology so far, it can be expected that an innovative energy transfer system is possible to realize.

Experience of on-site disposal of production uranium-graphite nuclear reactor.

PubMed

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

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.

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.

Modeling and Design Optimization of Multifunctional Membrane Reactors for Direct Methane Aromatization

PubMed Central

Fouty, Nicholas J.; Carrasco, Juan C.; Lima, Fernando V.

2017-01-01

Due to the recent increase of natural gas production in the U.S., utilizing natural gas for higher-value chemicals has become imperative. Direct methane aromatization (DMA) is a promising process used to convert methane to benzene, but it is limited by low conversion of methane and rapid catalyst deactivation by coking. Past work has shown that membrane separation of the hydrogen produced in the DMA reactions can dramatically increase the methane conversion by shifting the equilibrium toward the products, but it also increases coke production. Oxygen introduction into the system has been shown to inhibit this coke production while not inhibiting the benzene production. This paper introduces a novel mathematical model and design to employ both methods in a multifunctional membrane reactor to push the DMA process into further viability. Multifunctional membrane reactors, in this case, are reactors where two different separations occur using two differently selective membranes, on which no systems studies have been found. The proposed multifunctional membrane design incorporates a hydrogen-selective membrane on the outer wall of the reaction zone, and an inner tube filled with airflow surrounded by an oxygen-selective membrane in the middle of the reactor. The design is shown to increase conversion via hydrogen removal by around 100%, and decrease coke production via oxygen addition by 10% when compared to a tubular reactor without any membranes. Optimization studies are performed to determine the best reactor design based on methane conversion, along with coke and benzene production. The obtained optimal design considers a small reactor (length = 25 cm, diameter of reaction tube = 0.7 cm) to subvert coke production and consumption of the product benzene as well as a high permeance (0.01 mol/s·m2·atm1/4) through the hydrogen-permeable membrane. This modeling and design approach sets the stage for guiding further development of multifunctional membrane reactor models and designs for natural gas utilization and other chemical reaction systems. PMID:28850068

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

PubMed

Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-08-06

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

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

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

PubMed Central

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-01-01

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

Improving the cyanide toxicity tolerance of anaerobic reactor: Microbial interactions and toxin reduction.

PubMed

Gupta, Pragya; Ahammad, S Z; Sreekrishnan, T R

2016-09-05

Anaerobic biological treatment of high organics containing wastewater is amongst the preferred treatment options but poor tolerance to toxins makes its use prohibitive. In this study, efforts have been made to understand the key parameters for developing anaerobic reactor, resilient to cyanide toxicity. A laboratory scale anaerobic batch reactor was set up to treat cyanide containing wastewater. The reactor was inoculated with anaerobic sludge obtained from a wastewater treatment plant and fresh cow dung in the ratio of 3:1. The focus was on acclimatization and development of cyanide-degrading biomass and to understand the toxic effects of cyanide on the dynamic equilibrium between various microbial groups. The sludge exposed to cyanide was found to have higher bacterial diversity than the control. It was observed that certain hydrogenotrophic methanogens and bacterial groups were able to grow and produce methane in the presence of cyanide. Also, it was found that hydrogen utilizing methanogens were more cyanide tolerant than acetate utilizing methanogens. So, effluents from various industries like electroplating, coke oven plant, petroleum refining, explosive manufacturing, and pesticides industries which are having high concentrations of cyanide can be treated by favoring the growth of the tolerant microbes in the reactors. It will provide much better treatment efficiency by overcoming the inhibitory effects of cyanide to certain extent. Copyright © 2016 Elsevier B.V. All rights reserved.

Comparison of different solar reactors for household disinfection of drinking water in developing countries: evaluation of their efficacy in relation to the waterborne enteropathogen Cryptosporidium parvum.

PubMed

Gómez-Couso, H; Fontán-Sainz, M; Navntoft, C; Fernández-Ibáñez, P; Ares-Mazás, E

2012-11-01

Solar water disinfection (SODIS) is a type of treatment that can significantly improve the microbiological quality of drinking water at household level and therefore prevent waterborne diseases in developing countries. Cryptosporidium parvum is an obligate protozoan parasite responsible for the diarrhoeal disease cryptosporidiosis in humans and animals. Recently, this parasite has been selected by the WHO as a reference pathogen for protozoan parasites in the evaluation of household water treatment options. In this study, the field efficacy of different static solar reactors [1.5 l transparent plastic polyethylene terephthalate (PET) bottles as well as 2.5 l borosilicate glass and 25 l methacrylate reactors fitted with compound parabolic concentrators (CPC)] for solar disinfection of turbid waters experimentally contaminated with C. parvum oocysts was compared. Potential oocyst viability was determined by inclusion/exclusion of the fluorogenic vital dye propidium iodide. The results demonstrate that static solar reactors fitted with CPCs are an excellent alternative to the conventional SODIS method with PET bottles. These reactors improved the efficacy of the SODIS method by enabling larger volumes of water to be treated and, in some cases, the C. parvum oocysts were rendered totally unviable, minimising the negative effects of turbidity. Copyright © 2012 Royal Society of Tropical Medicine and Hygiene. Published by Elsevier Ltd. All rights reserved.

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

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

Wang, Jy-An John

Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regardingmore » Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.« less

Quantity and management of spent fuel from prototype and research reactors in Germany

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

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on themore » information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)« less

Biofilm development during the start-up of a sulfate-reducing down-flow fluidized bed reactor at different COD/SO4(2-) ratios and HRT.

PubMed

Piña-Salazar, E Z; Cervantes, F J; Meraz, M; Celis, L B

2011-01-01

In sulfate-reducing reactors, it has been reported that the sulfate removal efficiency increases when the COD/SO4(2-) ratio is increased. The start-up of a down-flow fluidized bed reactor constitutes an important step to establish a microbial community in the biofilm able to survive under the operational bioreactor conditions in order to achieve effective removal of both sulfate and organic matter. In this work the influence of COD/SO4(2-) ratio and HRT in the development of a biofilm during reactor start-up (35 days) was studied. The reactor was inoculated with 1.6 g VSS/L of granular sludge, ground low density polyethylene was used as support material; the feed consisted of mineral medium at pH 5.5 containing 1 g COD/L (acetate:lactate, 70:30) and sodium sulfate. Four experiments were conducted at HRT of 1 or 2 days and COD/SO4(2-) ratio of 0.67 or 2.5. The results obtained indicated that a COD/SO4(2-) ratio of 2.5 and HRT 2 days allowed high sulfate and COD removal (66.1 and 69.8%, respectively), whereas maximum amount of attached biomass (1.9 g SVI/L support) and highest sulfate reducing biofilm activity (10.1 g COD-H2S/g VSS-d) was achieved at HRT of 1 day and at COD/sulfate ratios of 0.67 and 2.5, respectively, which suggests that suspended biomass also played a key role in the performance of the reactors.

MODELING THE ELECTROLYTIC DECHLORINATION OF TRICHLOROETHYLENE IN A GRANULAR GRAPHITE-PACKED REACTOR

EPA Science Inventory

A comprehensive reactor model was developed for the electrolytic dechlorination of trichloroethylene (TCE) at a granular-graphite cathode. The reactor model describes the dynamic processes of TCE dechlorination and adsorption, and the formation and dechlorination of all the major...

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.

Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

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

Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.

2014-01-01

The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic usedmore » fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.« less

Multi-Megawatt Space Nuclear Power Generation

DTIC Science & Technology

1993-06-28

electric generation, both for open- and closed-cycle opera- tion. These reactors use the particulate fuel of the type developed for HTGR reactors. What...commercial HTGR power reactors, the particles are held in place and directly cooled. Figure 2.7 shows the two types of fuel particles developed for...of MW(e), for pulsed energy devices. The FBR would use HTGR -type particle fuel , contained in a annular bed be- tween two porous frits. Helium would

The role of nuclear reactors in space exploration and development

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

Lipinski, R.J.

2000-07-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. Onemore » reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built and flew space reactors; it is time to do so again.« 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

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

DOE PAGES

Melin, Alexander M.; Kisner, Roger A.

2018-04-03

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operatingmore » in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.« less

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

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

Melin, Alexander M.; Kisner, Roger A.

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operatingmore » in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.« less

The Sustainable Nuclear Future: Fission and Fusion E.M. Campbell Logos Technologies

NASA Astrophysics Data System (ADS)

Campbell, E. Michael

2010-02-01

Global industrialization, the concern over rising CO2 levels in the atmosphere and other negative environmental effects due to the burning of hydrocarbon fuels and the need to insulate the cost of energy from fuel price volatility have led to a renewed interest in nuclear power. Many of the plants under construction are similar to the existing light water reactors but incorporate modern engineering and enhanced safety features. These reactors, while mature, safe and reliable sources of electrical power have limited efficiency in converting fission power to useful work, require significant amounts of water, and must deal with the issues of nuclear waste (spent fuel), safety, and weapons proliferation. If nuclear power is to sustain its present share of the world's growing energy needs let alone displace carbon based fuels, more than 1000 reactors will be needed by mid century. For this to occur new reactors that are more efficient, versatile in their energy markets, require minimal or no water, produce less waste and more robust waste forms, are inherently safe and minimize proliferation concerns will be necessary. Graphite moderated, ceramic coated fuel, and He cooled designs are reactors that can satisfy these requirements. Along with other generation IV fast reactors that can further reduce the amounts of spent fuel and extend fuel resources, such a nuclear expansion is possible. Furthermore, facilities either in early operations or under construction should demonstrate the next step in fusion energy development in which energy gain is produced. This demonstration will catalyze fusion energy development and lead to the ultimate development of the next generation of nuclear reactors. In this presentation the role of advanced fission reactors and future fusion reactors in the expansion of nuclear power will be discussed including synergies with the existing worldwide nuclear fleet. )

Heat Pipe Solar Receiver for Oxygen Production of Lunar Regolith

NASA Astrophysics Data System (ADS)

Hartenstine, John R.; Anderson, William G.; Walker, Kara L.; Ellis, Michael C.

2009-03-01

A heat pipe solar receiver operating in the 1050° C range is proposed for use in the hydrogen reduction process for the extraction of oxygen from the lunar soil. The heat pipe solar receiver is designed to accept, isothermalize and transfer solar thermal energy to reactors for oxygen production. This increases the available area for heat transfer, and increases throughput and efficiency. The heat pipe uses sodium as the working fluid, and Haynes 230 as the heat pipe envelope material. Initial design requirements have been established for the heat pipe solar receiver design based on information from the NASA In-Situ Resource Utilization (ISRU) program. Multiple heat pipe solar receiver designs were evaluated based on thermal performance, temperature uniformity, and integration with the solar concentrator and the regolith reactor(s). Two designs were selected based on these criteria: an annular heat pipe contained within the regolith reactor and an annular heat pipe with a remote location for the reactor. Additional design concepts have been developed that would use a single concentrator with a single solar receiver to supply and regulate power to multiple reactors. These designs use variable conductance or pressure controlled heat pipes for passive power distribution management between reactors. Following the design study, a demonstration heat pipe solar receiver was fabricated and tested. Test results demonstrated near uniform temperature on the outer surface of the pipe, which will ultimately be in contact with the regolith reactor.

BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

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

Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel

2015-09-01

BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on themore » formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.« less

Alternate Tritium Production Methods Using A Liquid Lithium Target

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

Wilson, J.

For over 60 years, the Savannah River Site’s primary mission has been the production of tritium. From the beginning, the Savannah River National Laboratory (SRNL) has provided the technical foundation to ensure the successful execution of this critical defense mission. SRNL has developed most of the processes used in the tritium mission and provides the research and development necessary to supply this critical component. This project was executed by first developing reactor models that could be used as a neutron source. In parallel to this development calculations were carried out testing the feasibility of accelerator technologies that could also bemore » used for tritium production. Targets were designed with internal moderating material and optimized target was calculated to be capable of 3000 grams using a 1400 MWt sodium fast reactor, 850 grams using a 400 MWt sodium fast reactor, and 100 grams using a 62 MWt reactor, annually.« less

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

Kristine Barrett; Shannon Bragg-Sitton

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system thatmore » would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.« less

Cure kinetics, morphologies, and mechanical properties of thermoplastic/MWCNT modified multifunctional glassy epoxies prepared via continuous reaction methods

NASA Astrophysics Data System (ADS)

Cheng, Xiaole

The primary goal of this dissertation is to develop a novel continuous reactor method to prepare partially cured epoxy prepolymers for aerospace prepreg applications with the aim of replacing traditional batch reactors. Compared to batch reactors, the continuous reactor is capable of solubilizing and dispersing a broad range of additives including thermoplastic tougheners, stabilizers, nanoparticles and curatives and advancing epoxy molecular weights and viscosities while reducing energy consumption. In order to prove this concept, polyethersulfone (PES) modified 4, 4'-diaminodiphenylsulfone (44DDS)/tetraglycidyl-4, 4'-diaminodiphenylmethane (TGDDM) epoxy prepolymers were firstly prepared using both continuous reactor and batch reactor methods. Kinetic studies confirmed the chain extension reaction in the continuous reactor is similar to the batch reactor, and the molecular weights and viscosities of prepolymers were readily controlled through reaction kinetics. Atomic force microscopy (AFM) confirmed similar cured network morphologies for formulations prepared from batch and continuous reactors. Additionally tensile strength, tensile modulus and fracture toughness analyses concluded mechanical properties of cured epoxy matrices produced from both reactors were equivalent. Effects of multifunctional epoxy compositions on thermoplastics phase-separated morphologies were systematically studied using a combination of AFM with nanomechanical mapping, spectroscopic and calorimetric techniques to provide new insights to tailor cured reaction induced phase separation (CRIPS) in multifunctional epoxy blend networks. Furthermore, how resultant crosslinked glassy polymer network and phase-separated morphologies correlated with mechanical properties are discussed in detail. Multiwall carbon nanotube (MWCNT)/TGDDM epoxy prepolymers were further prepared by combining the successful strategies for advancing epoxy chemistries and dispersing nanotubes using the continuous reactor. Optical microscopy (OM) and scanning electron microscopy (SEM) were used to characterize the MWCNT dispersion states and stabilization in epoxy prepolymer matrix after continuous process and during curing cycles. Additionally, electrical conductivities and mechanical properties of final cured MWCNT/TGDDM composites were measured and discussed in view of their corresponding MWCNT dispersion states. Ternary blends of MWCNT reinforced thermoplastic/epoxy prepolymers were prepared by the continuous reactor. Influence of MWCNT on the CRIPS mechanism and the cured morphologies were systematically investigated using SEM and rheological analysis. Incorporation of MWCNT in thermoplastic/epoxy matrices can lead to a morphological transformation from phase inverted, to co-continuous, and to droplet dispersed morphology. In additional, dynamic mechanical analysis revealed the heterogeneity of MWCNT dispersion in thermoplastic/thermosets systems.

Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

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

Spencer, Benjamin; Backman, Marie; Williams, Paul

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decisionmore » making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.« less

Pilot scale system for the production of palm-based Monoester-OH

NASA Astrophysics Data System (ADS)

Ngah, Muhammad Syukri; Badri, Khairiah Haji

2016-11-01

A mechanically agitate reactor vessel in a moderate scale size of 500 L has been developed. This vessel was constructed to produce palm-based polyurethane polyol with a capacity of maximum 400 L. This is to accomodate the demand required for marketing trial run as part of the commercialization intention. The chemistry background of the process design was thoroughly studied. The esterification and condensation in batch process was maintained from the laboratory scale. Only RBD palm kernel oil was used in this study. This paper will describe the engineering design for the reactor vessel development beginning at the stoichiometric equations for the production process to the detail engineering including the equipment selection and fabrication in order to meet the design and objective specifications.

Evaluation of infrared thermography as a diagnostic tool in CVD applications

NASA Astrophysics Data System (ADS)

Johnson, E. J.; Hyer, P. V.; Culotta, P. W.; Clark, I. O.

1998-05-01

This research is focused on the feasibility of using infrared temperature measurements on the exterior of a chemical vapor deposition (CVD) reactor to ascertain both real-time information on the operating characteristics of a CVD system and provide data which could be post-processed to provide quantitative information for research and development on CVD processes. Infrared thermography techniques were used to measure temperatures on a horizontal CVD reactor of rectangular cross section which were correlated with the internal gas flow field, as measured with the laser velocimetry (LV) techniques. For the reactor tested, thermal profiles were well correlated with the gas flow field inside the reactor. Correlations are presented for nitrogen and hydrogen carrier gas flows. The infrared data were available to the operators in real time with sufficient sensitivity to the internal flow field so that small variations such as misalignment of the reactor inlet could be observed. The same data were post-processed to yield temperature measurements at known locations on the reactor surface. For the experiments described herein, temperatures associated with approximately 3.3 mm 2 areas on the reactor surface were obtained with a precision of ±2°C. These temperature measurements were well suited for monitoring a CVD production reactor, development of improved thermal boundary conditions for use in CFD models of reactors, and for verification of expected thermal conditions.

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

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

E.T. Robinson; John Sirman; Prasad Apte

2005-05-01

This final report summarizes work accomplished in the Program from January 1, 2001 through December 31, 2004. Most of the key technical objectives for this program were achieved. A breakthrough material system has lead to the development of an OTM (oxygen transport membrane) compact planar reactor design capable of producing either syngas or hydrogen. The planar reactor shows significant advantages in thermal efficiency and a step change reduction in costs compared to either autothermal reforming or steam methane reforming with CO{sub 2} recovery. Syngas derived ultra-clean transportation fuels were tested in the Nuvera fuel cell modular pressurized reactor and inmore » International Truck and Engine single cylinder test engines. The studies compared emission and engine performance of conventional base fuels to various formulations of ultra-clean gasoline or diesel fuels. A proprietary BP oxygenate showed significant advantage in both applications for reducing emissions with minimal impact on performance. In addition, a study to evaluate new fuel formulations for an HCCI engine was completed.« less

Biodrying process: A sustainable technology for treatment of municipal solid waste with high moisture content.

PubMed

Tom, Asha P; Pawels, Renu; Haridas, Ajit

2016-03-01

Municipal solid waste with high moisture content is the major hindrance in the field of waste to energy conversion technologies and here comes the importance of biodrying process. Biodrying is a convective evaporation process, which utilizes the biological heat developed from the aerobic reactions of organic components. The numerous end use possibilities of the output are making the biodrying process versatile, which is possible by achieving the required moisture reduction, volume reduction and bulk density enhancement through the effective utilization of biological heat. In the present case study the detailed research and development of an innovative biodrying reactor has been carried out for the treatment of mixed municipal solid waste with high moisture content. A pilot scale biodrying reactor of capacity 565 cm(3) was designed and set up in the laboratory. The reactor dimensions consisted of an acrylic chamber of 60 cm diameter and 200 cm height, and it was enveloped by an insulation chamber. The insulation chamber was provided to minimise the heat losses through the side walls of the reactor. It simulates the actual condition in scaling up of the reactor, since in bigger scale reactors the heat losses through side walls will be negligible while comparing the volume to surface area ratio. The mixed municipal solid waste with initial moisture content of 61.25% was synthetically prepared in the laboratory and the reactor was fed with 109 kg of this substrate. Aerobic conditions were ensured inside the reactor chamber by providing the air at a constant rate of 40 litre per minute, and the direction of air flow was from the specially designed bottom air chamber to the reactor matrix top. The self heating inside reactor matrix was assumed in the range of 50-60°C during the design stage. Innovative biodrying reactor was found to be efficiently working with the temperature inside the reactor matrix rising to a peak value of 59°C by the fourth day of experiment (the peak observed at a height of 60 cm from the air supply). The process analyses results were promising with a reduction of 56.5% of volume, and an increase of 52% of bulk density of the substrate at the end of 33 days of biodrying. Also the weight of mixed MSW substrate has been reduced by 33.94% in 20 days of reaction and the average moisture reduction of the matrix was 20.81% (reduced from the initial value of 61.25% to final value of 48.5%). The moisture reduction would have been higher, if the condensation of evaporated water at the reactor matrix has been avoided. The non-homogeneous moisture reduction along the height of the reactor is evident and this needs further innovation. The leachate production has been completely eliminated in the innovative biodrying reactor and that is a major achievement in the field of municipal solid waste management technology. Copyright © 2016 Elsevier Ltd. All rights reserved.

The present situations and perspectives on utilization of research reactors in Thailand

NASA Astrophysics Data System (ADS)

Chongkum, Somporn

2002-01-01

The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×10 13 n/cm 2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future.

Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

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

Beatty, Randy L; Harrison, Thomas J

IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical ofmore » commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.« less

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

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

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

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

Bucknor, M.; Farmer, M.; Grabaskas, D.

The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination ofmore » gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.« less

Nitrate removal from groundwater by cooperating heterotrophic with autotrophic denitrification in a biofilm-electrode reactor.

PubMed

Zhao, Yingxin; Feng, Chuanping; Wang, Qinghong; Yang, Yingnan; Zhang, Zhenya; Sugiura, Norio

2011-09-15

An intensified biofilm-electrode reactor (IBER) combining heterotrophic and autotrophic denitrification was developed for treatment of nitrate contaminated groundwater. The reactor was evaluated with synthetic groundwater (NO(3)(-)-N50 mg L(-1)) under different hydraulic retention times (HRTs), carbon to nitrogen ratios (C/N) and electric currents (I). The experimental results demonstrate that high nitrate and nitrite removal efficiency (100%) were achieved at C/N = 1, HRT = 8h, and I = 10 mA. C/N ratios were reduced from 1 to 0.5 and the applied electric current was changed from 10 to 100 mA, showing that the optimum running condition was C/N = 0.75 and I = 40 mA, under which over 97% of NO(3)(-)-N was removed and organic carbon (methanol) was completely consumed in treated water. Simultaneously, the denitrification mechanism in this system was analyzed through pH variation in effluent. The CO(2) produced from the anode acted as a good pH buffer, automatically controlling pH in the reaction zone. The intensified biofilm-electrode reactor developed in the study was effective for the treatment of groundwater polluted by nitrate. Copyright © 2011 Elsevier B.V. All rights reserved.

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

Development of a New Design Procedure for Overland Flow System.

DTIC Science & Technology

1982-06-18

reactor kinetics, a concept familiar to most environmental engi- neers. In the case of overland flow, the reactor is the soil surface where various physical...site during the entire study. Perforated plastic pipe was used to distri- bute wastewater along the top of each section, and a bed of crushed stone...particulate BOD. The soluble BOD is oxidized by microorganisms which are probably similar to the attached biomass found in trickling filters. However, some

INCAS: an analytical model to describe displacement cascades

NASA Astrophysics Data System (ADS)

Jumel, Stéphanie; Claude Van-Duysen, Jean

2004-07-01

REVE (REactor for Virtual Experiments) is an international project aimed at developing tools to simulate neutron irradiation effects in Light Water Reactor materials (Fe, Ni or Zr-based alloys). One of the important steps of the project is to characterise the displacement cascades induced by neutrons. Accordingly, the Department of Material Studies of Electricité de France developed an analytical model based on the binary collision approximation. This model, called INCAS (INtegration of CAScades), was devised to be applied on pure elements; however, it can also be used on diluted alloys (reactor pressure vessel steels, etc.) or alloys composed of atoms with close atomic numbers (stainless steels, etc.). INCAS describes displacement cascades by taking into account the nuclear collisions and electronic interactions undergone by the moving atoms. In particular, it enables to determine the mean number of sub-cascades induced by a PKA (depending on its energy) as well as the mean energy dissipated in each of them. The experimental validation of INCAS requires a large effort and could not be carried out in the framework of the study. However, it was verified that INCAS results are in conformity with those obtained from other approaches. As a first application, INCAS was applied to determine the sub-cascade spectrum induced in iron by the neutron spectrum corresponding to the central channel of the High Flux Irradiation Reactor of Oak Ridge National Laboratory.

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

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

Worrall, Andrew; Todosow, Michael

2016-01-01

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include:more » increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance metrics for a small modular reactor are compared to a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. Metrics performance for a small modular reactor are degraded for mass of spent nuclear fuel and high level waste disposed, mass of depleted uranium disposed, land use per energy generated, and carbon emission per energy generated« less

Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

NASA Technical Reports Server (NTRS)

Gould, R. E.; Petticrew, R. W.

1973-01-01

This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

Assessing the influence of reactor system design criteria on the performance of model colon fermentation units.

PubMed

Moorthy, Arun S; Eberl, Hermann J

2014-04-01

Fermentation reactor systems are a key platform in studying intestinal microflora, specifically with respect to questions surrounding the effects of diet. In this study, we develop computational representations of colon fermentation reactor systems as a way to assess the influence of three design elements (number of reactors, emptying mechanism, and inclusion of microbial immobilization) on three performance measures (total biomass density, biomass composition, and fibre digestion efficiency) using a fractional-factorial experimental design. It was determined that the choice of emptying mechanism showed no effect on any of the performance measures. Additionally, it was determined that none of the design criteria had any measurable effect on reactor performance with respect to biomass composition. It is recommended that model fermentation systems used in the experimenting of dietary effects on intestinal biomass composition be streamlined to only include necessary system design complexities, as the measured performance is not benefited by the addition of microbial immobilization mechanisms or semi-continuous emptying scheme. Additionally, the added complexities significantly increase computational time during simulation experiments. It was also noted that the same factorial experiment could be directly adapted using in vitro colon fermentation systems. Copyright © 2013 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

Formation of aerobic granular sludge during the treatment of petrochemical wastewater.

PubMed

Caluwé, Michel; Dobbeleers, Thomas; D'aes, Jolien; Miele, Solange; Akkermans, Veerle; Daens, Dominique; Geuens, Luc; Kiekens, Filip; Blust, Ronny; Dries, Jan

2017-08-01

In this study, petrochemical wastewater from the port of Antwerp was used for the development of aerobic granular sludge. Two different reactor setups were used, (1) a completely aerated sequencing batch reactor (SBR ae ) with a feast/famine regime and (2) a sequencing batch reactor operated with an anaerobic feast/aerobic famine strategy (SBR an ). The seed sludge showed poor settling characteristics with a sludge volume index (SVI) of 285mL.gMLSS -1 and a median particle size by volume of 86.0µm±1.9µm. In both reactors, granulation was reached after 30days with a SVI of 71mL.gMLSS -1 and median granule size of 264.7µm in SBR an and a SVI of 56mL.gMLSS -1 and median granule size of 307.4µm in SBR ae . The chemical oxygen demand (COD) and dissolved organic carbon (DOC) removal was similar in both reactors and above 95%. The anaerobic DOC uptake increased from 0.13% to 43.2% in 60days in SBR an . Copyright © 2017 Elsevier Ltd. All rights reserved.

Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

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

Hess, J. D.; Gauld, I. C.; Gulliford, J.

2017-01-01

Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Projectmore » (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.« less

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

Core Design Characteristics of the Fluoride Salt-Cooled High Temperature Demonstration Reactor

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

Brown, Nicholas R; Qualls, A L; Betzler, Benjamin R

2016-01-01

Fluoride salt-cooled high temperature reactors (FHRs) are a promising reactor technology option with significant knowledge gaps to implementation. One potential approach to address those technology gaps is via a small-scale demonstration reactor with the goal of increasing the technology readiness level (TRL) of the overall system for the longer term. The objective of this paper is to outline a notional concept for such a system, and to address how the proposed concept would advance the TRL of FHR concepts. Development of the proposed FHR Demonstration Reactor (DR) will enable commercial FHR deployment through disruptive and rapid technology development and demonstration.more » The FHR DR will close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. Important capabilities that will be demonstrated by building and operating the FHR DR include core design methodologies; fabrication and operation of high temperature reactors; salt procurement, handling, maintenance, and ultimate disposal; salt chemistry control to maximize vessel life; tritium management; heat exchanger performance; pump performance; and reactivity control. The FHR DR is considered part of a broader set of FHR technology development and demonstration efforts, some of which are already underway. Nonreactor test efforts (e.g., heated salt loops or loops using simulant fluids) can demonstrate many technologies necessary for commercial deployment of FHRs. The FHR DR, however, fulfills a crucial role in FHR technology development by advancing the technical maturity and readiness level of the system as a whole.« less

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.

Strategy proposed by Electricite de France in the development of automatic tools

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

Castaing, C.; Cazin, B.

1995-03-01

The strategy proposed by EDF in the development of a means to limit personal and collective dosimetry is recent. It follows in the steps of a policy that consisted of developing remote operation means for those activities of inspection and maintenance on the reactor, pools bottom, steam generators (SGs), also reactor building valves; target activities because of their high dosimetric cost. One of the main duties of the UTO (Technical Support Department), within the EDF, is the maintenance of Pressurized Water Reactors in French Nuclear Power Plant Operations (consisting of 54 units) and the development and monitoring of specialized tools.more » To achieve this, the UTO has started a national think-tank on the implementation of the ALARA process in its field of activity and created an ALARA Committee responsible for running and monitoring it, as well as a policy for developing tools. This point will be illustrated in the second on reactor vessel heads.« less

Small reactor power system for space application

NASA Technical Reports Server (NTRS)

Shirbacheh, M.

1987-01-01

A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

The U.S. RERTR program status and progress.

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

Travelli, A.

1998-01-21

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement ofmore » equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU Fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be satisfied by any existing fuel. These new advanced fuels will enable conversion of the reactors which cannot be converted today, ensure better efficiency and performance for all research reactors, and allow the design of more powerful new advanced LEU reactors. As in the past, the success of the RERTR program will depend on free exchange of ideas and information, and on the international friendship and cooperation that have been a trademark of the RERTR program since its inception.« less

Can high fields save the tokamak? The challenge of steady-state operation for low cost compact reactors

NASA Astrophysics Data System (ADS)

Freidberg, Jeffrey; Dogra, Akshunna; Redman, William; Cerfon, Antoine

2016-10-01

The development of high field, high temperature superconductors is thought to be a game changer for the development of fusion power based on the tokamak concept. We test the validity of this assertion for pilot plant scale reactors (Q 10) for two different but related missions: pulsed operation and steady-state operation. Specifically, we derive a set of analytic criteria that determines the basic design parameters of a given fusion reactor mission. As expected there are far more constraints than degrees of freedom in any given design application. However, by defining the mission of the reactor under consideration, we have been able to determine the subset of constraints that drive the design, and calculate the values for the key parameters characterizing the tokamak. Our conclusions are as follows: 1) for pulsed reactors, high field leads to more compact designs and thus cheaper reactors - high B is the way to go; 2) steady-state reactors with H-mode like transport are large, even with high fields. The steady-state constraint is hard to satisfy in compact designs - high B helps but is not enough; 3) I-mode like transport, when combined with high fields, yields relatively compact steady-state reactors - why is there not more research on this favorable transport regime?

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

A study on the use of the BioBall® as a biofilm carrier in a sequencing batch reactor.

PubMed

Masłoń, Adam; Tomaszek, Janusz A

2015-11-01

Described in this study are experiments conducted to evaluate the removal of organics and nutrients from synthetic wastewater by a moving bed sequencing batch biofilm reactor using BioBall® carriers as biofilm media. The work involving a 15L-laboratory scale MBSBBR (moving bed sequencing batch biofilm reactor) model showed that the wastewater treatment system was based on biochemical processes taking place with activated sludge and biofilm microorganisms developing on the surface of the BioBall® carriers. Classical nitrification and denitrification and the typical enhanced biological phosphorus removal process were achieved in the reactor analyzed, which operated with a volumetric organic loading of 0.84-0.978gCODL(-1)d(-1). The average removal efficiencies for COD, total nitrogen and total phosphorus were found to be 97.7±0.5%, 87.8±2.6% and 94.3±1.3%, respectively. Nitrification efficiency reached levels in the range 96.5-99.7%. Copyright © 2015 Elsevier Ltd. All rights reserved.

Preliminary Studies on Oleochemical Wastewater Treatment using Submerged Bed Biofilm Reactor (SBBR)

NASA Astrophysics Data System (ADS)

Ismail, Z.; Mahmood, N. A. N.; Ghafar, U. S. A.; Umor, N. A.; Muhammad, S. A. F.

2017-06-01

Wastewater discharge from the industry into water sources is one of the main reason for water pollution. The oleochemicals industry effluent produces high content of chemical oxygen demand (COD) with value between 6000-20,000 ppm. Effective treatment is required before wastewater effluent is discharged to environment. The aim of the study is to develop submerged bed biofilm reactor (SBBR) with packing materials in the cosmoball® carrier. Water quality such as chemical oxygen demands (COD), turbidity and pH were analysed. The result shows that the initial COD of 6000 ppm was reduced below 200 ppm. The optimum conditions for SBBR were obtained when green sponges used as packing material in cosmoball® effluent flowrate set at 100 mL/min; 1:1 ratio of cosmoball® volume to reactor volume and 1:1 ratio of active sludge (mixed culture) volume to reactor volume. Turbidity and pH were recorded with 9.0 NTU and 7.0 respectively, which indicated that SBBR is feasible as an alternative for conventional biological treatment in oleochemical industry.

Tailoring the response of Autonomous Reactivity Control (ARC) systems

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

Qvist, Staffan A.; Hellesen, Carl; Gradecka, Malwina

The Autonomous Reactivity Control (ARC) system was developed to ensure inherent safety of Generation IV reactors while having a minimal impact on reactor performance and economic viability. In this study we present the transient response of fast reactor cores to postulated accident scenarios with and without ARC systems installed. Using a combination of analytical methods and numerical simulation, the principles of ARC system design that assure stability and avoids oscillatory behavior have been identified. A comprehensive transient analysis study for ARC-equipped cores, including a series of Unprotected Loss of Flow (ULOF) and Unprotected Loss of Heat Sink (ULOHS) simulations, weremore » performed for Argonne National Laboratory (ANL) Advanced Burner Reactor (ABR) designs. With carefully designed ARC-systems installed in the fuel assemblies, the cores exhibit a smooth non-oscillatory transition to stabilization at acceptable temperatures following all postulated transients. To avoid oscillations in power and temperature, the reactivity introduced per degree of temperature change in the ARC system needs to be kept below a certain threshold the value of which is system dependent, the temperature span of actuation needs to be as large as possible.« less

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.

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

Baines, B.D.

The development of the two types of Jason reactor is reported (10-kw Standard Jason, 100-kw Jason). Essential data are given on their construction and operation. The projects which were, or could be, carried out with these reactors are briefiy mentioned, with special emphasis on the adaptability of the reactor to various uses. (autb)

Visualizing and quantifying dose distribution in a UV reactor using three-dimensional laser-induced fluorescence.

PubMed

Gandhi, Varun N; Roberts, Philip J W; Kim, Jae-Hong

2012-12-18

Evaluating the performance of typical water treatment UV reactors is challenging due to the complexity in assessing spatial and temporal variation of UV fluence, resulting from highly unsteady, turbulent nature of flow and variation in UV intensity. In this study, three-dimensional laser-induced fluorescence (3DLIF) was applied to visualize and quantitatively analyze a lab-scale UV reactor consisting of one lamp sleeve placed perpendicular to flow. Mapping the spatial and temporal fluence delivery and MS2 inactivation revealed the highest local fluence in the wake zone due to longer residence time and higher UV exposure, while the lowest local fluence occurred in a region near the walls due to short-circuiting flow and lower UV fluence rate. Comparing the tracer based decomposition between hydrodynamics and IT revealed similar coherent structures showing the dependency of fluence delivery on the reactor flow. The location of tracer injection, varying the height and upstream distance from the lamp center, was found to significantly affect the UV fluence received by the tracer. A Lagrangian-based analysis was also employed to predict the fluence along specific paths of travel, which agreed with the experiments. The 3DLIF technique developed in this study provides new insight on dose delivery that fluctuates both spatially and temporally and is expected to aid design and optimization of UV reactors as well as validate computational fluid dynamics models that are widely used to simulate UV reactor performances.

System Concepts for Affordable Fission Surface Power

NASA Technical Reports Server (NTRS)

Mason, Lee; Poston, David; Qualls, Louis

2008-01-01

This paper presents an overview of an affordable Fission Surface Power (FSP) system that could be used for NASA applications on the Moon and Mars. The proposed FSP system uses a low temperature, uranium dioxide-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The concept was determined by a 12 month NASA/DOE study that examined design options and development strategies based on affordability and risk. The system is considered a low development risk based on the use of terrestrial-derived reactor technology, high efficiency power conversion, and conventional materials. The low-risk approach was selected over other options that could offer higher performance and/or lower mass.

Method for depleting BWRs using optimal control rod patterns

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

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1991-01-01

Control rod (CR) programming is an essential core management activity for boiling water reactors (BWRs). After establishing a core reload design for a BWR, CR programming is performed to develop a sequence of exposure-dependent CR patterns that assure the safe and effective depletion of the core through a reactor cycle. A time-variant target power distribution approach has been assumed in this study. The authors have developed OCTOPUS to implement a new two-step method for designing semioptimal CR programs for BWRs. The optimization procedure of OCTOPUS is based on the method of approximation programming and uses the SIMULATE-E code for nucleonicsmore » calculations.« less

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

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee

The purpose of this collaborative Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, plant engineering, material compatibility studies, and coolant activation. The publications derived from workmore » on this project (since project inception) are listed in Appendix A.« less

Status of the US RERTR Program

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

Travelli, A.

1995-02-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1994 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1993 in collaboration with its many international partners. The RERTR Program has moved aggressively to support President Clinton`s nonproliferation policy and his goal {open_quotes}to minimize the use of highly-enriched uranium in civil nuclear programs{close_quotes}. An Environmental Assessment which addresses the urgent-relief acceptance of 409 spent fuel elements was completed, and the first shipment of spent fuel elements is scheduledmore » for this month. An Environmental Impact Statement addressing the acceptance of spent research reactor fuel containing enriched uranium of U.S. origin is scheduled for completion by the end of June 1995. The U.S. administration has decided to resume development of high-density LEU research reactor fuels. DOE funding and guidance are expected to begin soon. A preliminary plan for the resumption of fuel development has been prepared and is ready for implementation. The scope and main technical activities of a plan to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels was agreed upon by the RERTR Program and four Russian institutes lead by RDIPE. Both Secretary O`Leary and Minister Michailov have expressed strong support for this initiative. Joint studies have made significant progress, especially in assessing the technical and economic feasibility of using reduced enrichment fuels in the SAFARI-I reactor in South Africa and in the Advanced Neutron Source reactor under design at ORNL. Significant progress was achieved on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU to the achievement of the common goal.« less

Strengthening IAEA Safeguards for Research Reactors

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

Reid, Bruce D.; Anzelon, George A.; Budlong-Sylvester, Kory

During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half amore » dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To broaden the IAEA safeguards toolbox, the study recommends that the Agency consider closing potential gaps in safeguards coverage by, among other things: 1) adapting its safeguards measures based on a case-by-case assessment; 2) using more frequent and expanded/enhanced mailbox declarations (ideally with remote transmission of the data to IAEA Headquarters in Vienna) coupled with short-notice or unannounced inspections; 3) putting more emphasis on the collection and analysis of environmental samples at hot cells and waste storage tanks; 4) taking Safeguards by Design into account for the construction of new research reactors and best practices for existing research reactors; 5) utilizing fully all legal authorities to enhance inspection access (including a strengthened and continuing DIV process); and 6) utilizing new approaches to improve auditing activities, verify reactor operating data history, and track/monitor the movement and storage of spent fuel.« less

Development of probabilistic design method for annular fuels

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

Ozawa, Takayuki

2007-07-01

The increase of linear power and burn-up during the reactor operation is considered as one measure to ensure the utility of fast reactors in the future; for this the application of annular oxide fuels is under consideration. The annular fuel design code CEPTAR was developed in the Japan Atomic Energy Agency (JAEA) and verified by using many irradiation experiences with oxide fuels. In addition, the probabilistic fuel design code BORNFREE was also developed to provide a safe and reasonable fuel design and to evaluate the design margins quantitatively. This study aimed at the development of a probabilistic design method formore » annular oxide fuels; this was implemented in the developed BORNFREE-CEPTAR code, and the code was used to make a probabilistic evaluation with regard to the permissive linear power. (author)« less

A Small Fission Power System with Stirling Power Conversion for NASA Science Missions

NASA Technical Reports Server (NTRS)

Mason, Lee; Carmichael, Chad

2011-01-01

In early 2010, a joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) study team developed a concept for a 1 kWe Fission Power System with a 15-year design life that could be available for a 2020 launch to support future NASA science missions. The baseline concept included a solid block uranium-molybdenum reactor core with embedded heat pipes and distributed thermoelectric converters directly coupled to aluminum radiator fins. A short follow-on study was conducted at NASA Glenn Research Center (GRC) to evaluate an alternative power conversion approach. The GRC study considered the use of free-piston Stirling power conversion as a substitution to the thermoelectric converters. The resulting concept enables a power increase to 3 kWe with the same reactor design and scalability to 10 kW without changing the reactor technology. This paper presents the configuration layout, system performance, mass summary, and heat transfer analysis resulting from the study.

DEMONSTRATION BULLETIN: FLAME REACTOR - HORSEHEAD RESOURCE DEVELOPMENT COMPANY, INC.

EPA Science Inventory

The Horsehead Resource Development Company, Inc. (HRD) Flame Reactor is a patented and proven high temperature thermal process designed to safely treat industrial residues and wastes containing metals. During processing, the waste material is introduced into the hottest portio...

Year One Summary of X-energy Pebble Fuel Development at ORNL

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

Helmreich, Grant W.; Hunn, John D.; McMurray, Jake W.

2017-06-01

The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

Thermal Stratification Analysis for Sodium Fast Reactors

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

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

Imaging Fukushima Daiichi reactors with muons

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

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.

2013-05-15

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi tomore » make this determination in the near future.« less

Imaging Fukushima Daiichi reactors with muons

NASA Astrophysics Data System (ADS)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Lukić, Zarija; Masuda, Koji; Milner, Edward C.; Morris, Christopher L.; Perry, John O.

2013-05-01

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

Assessing pretreatment reactor scaling through empirical analysis

DOE PAGES

Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; ...

2016-10-10

Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was within the near-optimal space for total sugar yield for the LHR. This indicates that the ASE is a good tool for cost effectively finding near-optimal conditions for operating pilot-scale systems, which may be used as starting points for further optimization. Additionally, using a severity-factor approach to optimization was found to be inadequate compared to a multivariate optimization method. As a result, the ASE and the LHR were able to enable significantly higher total sugar yields after enzymatic hydrolysis relative to the ZCR, despite having similar optimal conditions and total xylose yields. This underscores the importance of incorporating mechanical disruption into pretreatment reactor designs to achieve high enzymatic digestibilities.« less

Assessing pretreatment reactor scaling through empirical analysis

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

Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik

Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was within the near-optimal space for total sugar yield for the LHR. This indicates that the ASE is a good tool for cost effectively finding near-optimal conditions for operating pilot-scale systems, which may be used as starting points for further optimization. Additionally, using a severity-factor approach to optimization was found to be inadequate compared to a multivariate optimization method. As a result, the ASE and the LHR were able to enable significantly higher total sugar yields after enzymatic hydrolysis relative to the ZCR, despite having similar optimal conditions and total xylose yields. This underscores the importance of incorporating mechanical disruption into pretreatment reactor designs to achieve high enzymatic digestibilities.« less

Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

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

Tarchalski, M.; Pytel, K.; Wroblewska, M.

2015-07-01

Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to themore » qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from measurements is of the order of 2.5 W/g at half of the possible MARIA power - 15 MW. The approach and the detailed program for experimental verification of calculations will be presented. The following points will be discussed: - Development of a gamma heating model of MARIA reactor with TRIPOLI 4 (coupled neutron-photon mode) and APOLLO2 model taking into account the key parameters like: configuration of the core, experimental loading, control rod location, reactor power, fuel depletion); - Design of specific measurement tools for MARIA experiments including for instance a new single-cell calorimeter called KAROLINA calorimeter; - MARIA experimental program description and a preliminary analysis of results; - Comparison of calculations for JHR and MARIA cores with experimental verification analysis, calculation behavior and n-γ 'environments'. (authors)« less

An advanced carbon reactor subsystem for carbon dioxide reduction

NASA Technical Reports Server (NTRS)

Noyes, Gary P.; Cusick, Robert J.

1986-01-01

An evaluation is presented of the development status of an advanced carbon-reactor subsystem (ACRS) for the production of water and dense, solid carbon from CO2 and hydrogen, as required in physiochemical air revitalization systems for long-duration manned space missions. The ACRS consists of a Sabatier Methanation Reactor (SMR) that reduces CO2 with hydrogen to form methane and water, a gas-liquid separator to remove product water from the methane, and a Carbon Formation Reactor (CFR) to pyrolize methane to carbon and hydrogen; the carbon is recycled to the SMR, while the produce carbon is periodically removed from the CFR. A preprototype ACRS under development for the NASA Space Station is described.

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

NASA Astrophysics Data System (ADS)

Gelles, D. S.

1990-05-01

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

Development of a Novel Catalytic Membrane Reactor for Heterogeneous Catalysis in Supercritical CO2

PubMed Central

Islam, Nazrul M.; Chatterjee, Maya; Ikushima, Yutaka; Yokoyama, Toshiro; Kawanami, Hajime

2010-01-01

A novel type of high-pressure membrane reactor has been developed for hydrogenation in supercritical carbon dioxide (scCO2). The main objectives of the design of the reactor are the separate feeding of hydrogen and substrate in scCO2 for safe reactions in a continuous flow process, and to reduce the reaction time. By using this new reactor, hydrogenation of cinnamaldehyde into hydrocinnamaldehyde has been successfully carried out with 100% selectivity at 50 °C in 10 MPa (H2: 1 MPa, CO2: 9 MPa) with a flow rate of substrate ranging from 0.05 to 1.0 mL/min. PMID:20162008

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

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

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Advances of zeolite based membrane for hydrogen production via water gas shift reaction

NASA Astrophysics Data System (ADS)

Makertihartha, I. G. B. N.; Zunita, M.; Rizki, Z.; Dharmawijaya, P. T.

2017-07-01

Hydrogen is considered as a promising energy vector which can be obtained from various renewable sources. However, an efficient hydrogen production technology is still challenging. One technology to produce hydrogen with very high capacity with low cost is through water gas shift (WGS) reaction. Water gas shift reaction is an equilibrium reaction that produces hydrogen from syngas mixture by the introduction of steam. Conventional WGS reaction employs two or more reactors in series with inter-cooling to maximize conversion for a given volume of catalyst. Membrane reactor as new technology can cope several drawbacks of conventional reactor by removing reaction product and the reaction will favour towards product formation. Zeolite has properties namely high temperature, chemical resistant, and low price makes it suitable for membrane reactor applications. Moreover, it has been employed for years as hydrogen selective layer. This review paper is focusing on the development of membrane reactor for efficient water gas shift reaction to produce high purity hydrogen and carbon dioxide. Development of membrane reactor is discussed further related to its modification towards efficient reaction and separation from WGS reaction mixture. Moreover, zeolite framework suitable for WGS membrane reactor will be discussed more deeply.

A low-cost municipal sewage treatment system with a combination of UASB and the "fourth-generation" downflow hanging sponge reactors.

PubMed

Tandukar, M; Uemura, S; Machdar, I; Ohashi, A; Harada, H

2005-01-01

This paper presents an evaluation of the process performance of a pilot-scale "fourth generation" downflow hanging sponge (DHS) post-treatment system combined with a UASB pretreatment unit treating municipal wastewater. After the successful operation of the second- and third-generation DHS reactors, the fourth-generation DHS reactor was developed to overcome a few shortcomings of its predecessors. This reactor was designed to further enhance the treatment efficiency and simplify the construction process in real scale, especially for the application in developing countries. Configuration of the reactor was modified to enhance the dissolution of air into the wastewater and to avert the possible clogging of the reactor especially during sudden washout from the UASB reactor. The whole system was operated at a total hydraulic retention time (HRT) of 8 h (UASB: 6 h and DHS: 2 h) for a period of over 600 days. The combined system was able to remove 96% of unfiltered BOD with only 9 mg/L remaining in the final effluent. Likewise, F. coli were removed by 3.45 log with the final count of 10(3) to 10(4) MPN/100 ml. Nutrient removal by the system was also satisfactory.

Molecular beam mass spectrometer equipped with a catalytic wall reactor for in situ studies in high temperature catalysis research

NASA Astrophysics Data System (ADS)

Horn, R.; Ihmann, K.; Ihmann, J.; Jentoft, F. C.; Geske, M.; Taha, A.; Pelzer, K.; Schlögl, R.

2006-05-01

A newly developed apparatus combining a molecular beam mass spectrometer and a catalytic wall reactor is described. The setup has been developed for in situ studies of high temperature catalytic reactions (>1000°C), which involve besides surface reactions also gas phase reactions in their mechanism. The goal is to identify gas phase radicals by threshold ionization. A tubular reactor, made from the catalytic material, is positioned in a vacuum chamber. Expansion of the gas through a 100μm sampling orifice in the reactor wall into differentially pumped nozzle, skimmer, and collimator chambers leads to the formation of a molecular beam. A quadrupole mass spectrometer with electron impact ion source designed for molecular beam inlet and threshold ionization measurements is used as the analyzer. The sampling time from nozzle to detector is estimated to be less than 10ms. A detection time resolution of up to 20ms can be reached. The temperature of the reactor is measured by pyrometry. Besides a detailed description of the setup components and the physical background of the method, this article presents measurements showing the performance of the apparatus. After deriving the shape and width of the energy spread of the ionizing electrons from measurements on N2 and He we estimated the detection limit in threshold ionization measurements using binary mixtures of CO in N2 to be in the range of several hundreds of ppm. Mass spectra and threshold ionization measurements recorded during catalytic partial oxidation of methane at 1250°C on a Pt catalyst are presented. The detection of CH3• radicals is successfully demonstrated.

Molecular beam mass spectrometer equipped with a catalytic wall reactor for in situ studies in high temperature catalysis research

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

Horn, R.; Ihmann, K.; Ihmann, J.

2006-05-15

A newly developed apparatus combining a molecular beam mass spectrometer and a catalytic wall reactor is described. The setup has been developed for in situ studies of high temperature catalytic reactions (>1000 deg. C), which involve besides surface reactions also gas phase reactions in their mechanism. The goal is to identify gas phase radicals by threshold ionization. A tubular reactor, made from the catalytic material, is positioned in a vacuum chamber. Expansion of the gas through a 100 {mu}m sampling orifice in the reactor wall into differentially pumped nozzle, skimmer, and collimator chambers leads to the formation of a molecularmore » beam. A quadrupole mass spectrometer with electron impact ion source designed for molecular beam inlet and threshold ionization measurements is used as the analyzer. The sampling time from nozzle to detector is estimated to be less than 10 ms. A detection time resolution of up to 20 ms can be reached. The temperature of the reactor is measured by pyrometry. Besides a detailed description of the setup components and the physical background of the method, this article presents measurements showing the performance of the apparatus. After deriving the shape and width of the energy spread of the ionizing electrons from measurements on N{sub 2} and He we estimated the detection limit in threshold ionization measurements using binary mixtures of CO in N{sub 2} to be in the range of several hundreds of ppm. Mass spectra and threshold ionization measurements recorded during catalytic partial oxidation of methane at 1250 deg. C on a Pt catalyst are presented. The detection of CH{sub 3}{center_dot} radicals is successfully demonstrated.« less

76 FR 78173 - Options for Developing the Regulatory Basis for Streamlining Non-Power Reactor License Renewal...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-16

... Emergency Preparedness AGENCY: Nuclear Regulatory Commission. ACTION: Notice of public meeting. SUMMARY: The... non-power reactor license renewal and non-power reactor emergency preparedness. This meeting is a... potential enhancements to emergency preparedness requirements. This meeting is open to the public. DATES...

Impact of VOC Composition and Reactor Conditions on the Aging of Biomass Cookstove Emission in an Oxidation Flow Reactor

EPA Science Inventory

Oxidation flow reactor (OFR) experiments in our lab have explored secondary organic aerosol (SOA) production during photochemical aging of emissions from cookstoves used by billions in developing countries. Previous experiments, conducted with red oak fuel under conditions of hig...

Impact of VOC Composition and Reactor Conditions on the Aging of Biomass Cookstove Emissions in an Oxidation Flow Reactor

EPA Science Inventory

Oxidation flow reactor (OFR) experiments in our lab have explored secondary organic aerosol (SOA) production during photochemical aging of emissions from cookstoves used by billions in developing countries. Previous experiments, conducted with red oak fuel under conditions of hig...

10 CFR 725.15 - Requirements for approval of applications.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Secret Restricted Data in C-91, Nuclear Reactors for Rocket Propulsion, will be approved only if the... capable of making a contribution to research and development in the field of nuclear reactors for rocket... the field of nuclear reactors for rocket propulsion preparatory to the submission of a research and...

10 CFR 725.15 - Requirements for approval of applications.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Secret Restricted Data in C-91, Nuclear Reactors for Rocket Propulsion, will be approved only if the... capable of making a contribution to research and development in the field of nuclear reactors for rocket... the field of nuclear reactors for rocket propulsion preparatory to the submission of a research and...

10 CFR 725.15 - Requirements for approval of applications.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Secret Restricted Data in C-91, Nuclear Reactors for Rocket Propulsion, will be approved only if the... capable of making a contribution to research and development in the field of nuclear reactors for rocket... the field of nuclear reactors for rocket propulsion preparatory to the submission of a research and...

Space nuclear reactors — A post-operational disposal strategy

NASA Astrophysics Data System (ADS)

Angelo, Joseph A.; Buden, David

If 100-kWe and multimegawatt-electric class space nuclear reactors are to play a significant role in humanity's push into cislunar and heliocentric space in the next millennium, the obvious advantages of space nuclear power plants should not be denied to space mission planners due to a failure to develop internationally-acceptable post-operational disposal strategies for spent reactor cores. This is true whether the space reactor has shut down at the end of its normal mission lifetime or in response to an onboard system failure/emergency which causes a premature mission termination. Up until now the great majority of aerospace nuclear safety efforts have concentrated on prelaunch, launch and reactor startup activities. In fact, with the exception of the development of the "nuclear safe orbit" (NSO) concept, little technical attention has yet been given to the post-operational disposal of future space reactors. This paper describes the technical alternatives available for the safe, acceptable disposal of space reactors that could be used in a wide variety of space applications in the 21st Century. Post-operational core radioactivity levels for typical advanced design (hundred kWe-class) space reactors are presented as a function of decay time and contrasted to the spent core radionuclide inventory of the SNAP-10A system, the only nuclear reactor operated in space by the United States. The role of a permanent space station, smart robotic systems, and an operating lunar base in support of spent core disposal strategies is also presented, including use of a selected portion of the lunar surface as an internationally-designated spent reactor core repository.

RELAP-7 Development Updates

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

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

Effect of mechanical disruption on the effectiveness of three reactors used for dilute acid pretreatment of corn stover Part 2: morphological and structural substrate analysis

PubMed Central

2014-01-01

Background Lignocellulosic biomass is a renewable, naturally mass-produced form of stored solar energy. Thermochemical pretreatment processes have been developed to address the challenge of biomass recalcitrance, however the optimization, cost reduction, and scalability of these processes remain as obstacles to the adoption of biofuel production processes at the industrial scale. In this study, we demonstrate that the type of reactor in which pretreatment is carried out can profoundly alter the micro- and nanostructure of the pretreated materials and dramatically affect the subsequent efficiency, and thus cost, of enzymatic conversion of cellulose. Results Multi-scale microscopy and quantitative image analysis was used to investigate the impact of different biomass pretreatment reactor configurations on plant cell wall structure. We identify correlations between enzymatic digestibility and geometric descriptors derived from the image data. Corn stover feedstock was pretreated under the same nominal conditions for dilute acid pretreatment (2.0 wt% H2SO4, 160°C, 5 min) using three representative types of reactors: ZipperClave® (ZC), steam gun (SG), and horizontal screw (HS) reactors. After 96 h of enzymatic digestion, biomass treated in the SG and HS reactors achieved much higher cellulose conversions, 88% and 95%, respectively, compared to the conversion obtained using the ZC reactor (68%). Imaging at the micro- and nanoscales revealed that the superior performance of the SG and HS reactors could be explained by reduced particle size, cellular dislocation, increased surface roughness, delamination, and nanofibrillation generated within the biomass particles during pretreatment. Conclusions Increased cellular dislocation, surface roughness, delamination, and nanofibrillation revealed by direct observation of the micro- and nanoscale change in accessibility explains the superior performance of reactors that augment pretreatment with physical energy. PMID:24690534

SUPERHEATING IN A BOILING WATER REACTOR

DOEpatents

Treshow, M.

1960-05-31

A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

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.

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

Evaluation of parallel milliliter-scale stirred-tank bioreactors for the study of biphasic whole-cell biocatalysis with ionic liquids.

PubMed

Dennewald, Danielle; Hortsch, Ralf; Weuster-Botz, Dirk

2012-01-01

As clear structure-activity relationships are still rare for ionic liquids, preliminary experiments are necessary for the process development of biphasic whole-cell processes involving these solvents. To reduce the time investment and the material costs, the process development of such biphasic reaction systems would profit from a small-scale high-throughput platform. Exemplarily, the reduction of 2-octanone to (R)-2-octanol by a recombinant Escherichia coli in a biphasic ionic liquid/water system was studied in a miniaturized stirred-tank bioreactor system allowing the parallel operation of up to 48 reactors at the mL-scale. The results were compared to those obtained in a 20-fold larger stirred-tank reactor. The maximum local energy dissipation was evaluated at the larger scale and compared to the data available for the small-scale reactors, to verify if similar mass transfer could be obtained at both scales. Thereafter, the reaction kinetics and final conversions reached in different reactions setups were analysed. The results were in good agreement between both scales for varying ionic liquids and for ionic liquid volume fractions up to 40%. The parallel bioreactor system can thus be used for the process development of the majority of biphasic reaction systems involving ionic liquids, reducing the time and resource investment during the process development of this type of applications. Copyright © 2011. Published by Elsevier B.V.

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

PubMed

Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

2014-01-01

The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

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.

Arc dynamics of a pulsed DC nitrogen rotating gliding arc discharge

NASA Astrophysics Data System (ADS)

Zhu, Fengsen; Zhang, Hao; Li, Xiaodong; Wu, Angjian; Yan, Jianhua; Ni, Mingjiang; Tu, Xin

2018-03-01

In this study, a novel pulsed direct current (DC) rotating gliding arc (RGA) plasma reactor co-driven by an external magnetic field and a tangential gas flow has been developed. The dynamic characteristics of the rotating gliding arc have been investigated by means of numerical simulation and experiment. The simulation results show that a highly turbulent vortex flow can be generated at the bottom of the RGA reactor to accelerate the arc rotation after arc ignition, whereas the magnitude of gas velocity declined significantly along the axial direction of the RGA reactor. The calculated arc rotation frequency (14.4 Hz) is reasonably close to the experimental result (18.5 Hz) at a gas flow rate of 10 l min-1. In the presence of an external magnet, the arc rotation frequency is around five times higher than that of the RGA reactor without using a magnet, which suggests that the external magnetic field plays a dominant role in the maintenance of the arc rotation in the upper zone of the RGA reactor. In addition, when the magnet is placed outside the reactor reversely to form a reverse external magnetic field, the arc can be stabilized at a fixed position in the inner wall of the outer electrode at a critical gas flow rate of 16 l min-1.

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

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

Paolo Balestra; Carlo Parisi; Andrea Alfonsi

2016-02-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution).more » Comparison between both solutions is briefly illustrated in this summary.« less

Impact of Gas Heating in Inductively Coupled Plasmas

NASA Technical Reports Server (NTRS)

Hash, D. B.; Bose, D.; Rao, M. V. V. S.; Cruden, B. A.; Meyyappan, M.; Sharma, S. P.; Biegel, Bryan (Technical Monitor)

2001-01-01

Recently it has been recognized that the neutral gas in inductively coupled plasma reactors heats up significantly during processing. The resulting gas density variations across the reactor affect reaction rates, radical densities, plasma characteristics, and uniformity within the reactor. A self-consistent model that couples the plasma generation and transport to the gas flow and heating has been developed and used to study CF4 discharges. A Langmuir probe has been used to measure radial profiles of electron density and temperature. The model predictions agree well with the experimental results. As a result of these comparisons along with the poorer performance of the model without the gas-plasma coupling, the importance of gas heating in plasma processing has been verified.

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

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.

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

Carla J. Miller

This report provides a summary of the literature review that was performed and based on previous work performed at the Idaho National Laboratory studying the Three Mile Island 2 (TMI-2) nuclear reactor accident, specifically the melted fuel debris. The purpose of the literature review was to document prior published work that supports the feasibility of the analytical techniques that were developed to provide quantitative results of the make-up of the fuel and reactor component debris located inside and outside the containment. The quantitative analysis provides a technique to perform nuclear fuel accountancy measurements

Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

NASA Technical Reports Server (NTRS)

Tew, R. C.; Jefferies, K. S.

1974-01-01

A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

Methods and strategies for future reactor safety goals

NASA Astrophysics Data System (ADS)

Arndt, Steven Andrew

There have been significant discussions over the past few years by the United States Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and others as to the adequacy of the NRC safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. In its review of the new plant licensing policy the ACRS has looked at the safety goals, as has the NRC. A number of issues have been raised including what the Commission had in mind when it drafted the safety goals and QHOs, how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. The NRC safety goals and the QHOs refer to a "nuclear power plant," but do not discuss whether a "plant" refers to only a single unit or all of the units on a site. There has been much discussion on this issue recently due to the development of modular reactors. Additionally, the risk of multiple reactor accidents on the same site has been largely ignored in the probabilistic risk assessments (PRAs) done to date, and in most risk-informed analyses and discussions. This dissertation examines potential approaches to updating the safety goals that include the establishment of new quantitative safety goal associated with the comparative risk of generating electricity by viable competing technologies and modifications of the goals to account for multi-plant reactor sites, and issues associated with the use of safety goals in both initial licensing and operational decision making. This research develops a new quantitative health objective that uses a comparable benefit risk metric based on the life-cycle risk of the construction, operation and decommissioning of a comparable non-nuclear electric generation facility, as well as the risks associated with mining and transportation. This dissertation also evaluates the effects of using various methods for aggregating site risk as a safety metric, as opposed to using single plant safety goals. Additionally, a number of important assumptions inherent in the current safety goals, including the effect of other potential negative societal effects such as the generation of greenhouse gases (e.g., carbon dioxide) have on the risk of electric power production and their effects on the setting of safety goals, is explored. Finally, the role risk perception should play in establishing safety goals has been explored. To complete this evaluation, a new method to analytically compare alternative technologies of generating electricity was developed, including development of a new way to evaluate risk perception, and a new method was developed for evaluating the risk at multiple units on a single site. To test these modifications to the safety goals a number of possible reactor designs and configurations were evaluated using these new proposed safety goals to determine the goals' usefulness and utility. The results of the analysis showed that the modifications provide measures that more closely evaluate the potential risk to the public from the operation of nuclear power plants than the current safety goals, while still providing a straight-forward process for assessment of reactor design and operation.

Development of an advanced antineutrino detector for reactor monitoring

DOE PAGES

Classen, T.; Bernstein, A.; Bowden, N. S.; ...

2014-11-05

We present the development of a compact antineutrino detector for the purpose of nuclear reactor monitoring, improving upon a previously successful design. Our paper will describe the design improvements of the detector which increases the antineutrino detection efficiency threefold over the previous effort. There are two main design improvements over previous generations of detectors for nuclear reactor monitoring: dual-ended optical readout and single volume detection mass. The dual-ended optical readout eliminates the need for fiducialization and increases the uniformity of the detector's optical response. The containment of the detection mass in a single active volume provides more target mass permore » detector footprint, a key design criteria for operating within a nuclear power plant. This technology could allow for real-time monitoring of the evolution of a nuclear reactor core, independent of reactor operator declarations of fuel inventories, and may be of interest to the safeguards community.« less

Model predictive control of a solar-thermal reactor

NASA Astrophysics Data System (ADS)

Saade Saade, Maria Elizabeth

Solar-thermal reactors represent a promising alternative to fossil fuels because they can harvest solar energy and transform it into storable and transportable fuels. The operation of solar-thermal reactors is restricted by the available sunlight and its inherently transient behavior, which affects the performance of the reactors and limits their efficiency. Before solar-thermal reactors can become commercially viable, they need to be able to maintain a continuous high-performance operation, even in the presence of passing clouds. A well-designed control system can preserve product quality and maintain stable product compositions, resulting in a more efficient and cost-effective operation, which can ultimately lead to scale-up and commercialization of solar thermochemical technologies. In this work, we propose a model predictive control (MPC) system for a solar-thermal reactor for the steam-gasification of biomass. The proposed controller aims at rejecting the disturbances in solar irradiation caused by the presence of clouds. A first-principles dynamic model of the process was developed. The model was used to study the dynamic responses of the process variables and to identify a linear time-invariant model used in the MPC algorithm. To provide an estimation of the disturbances for the control algorithm, a one-minute-ahead direct normal irradiance (DNI) predictor was developed. The proposed predictor utilizes information obtained through the analysis of sky images, in combination with current atmospheric measurements, to produce the DNI forecast. In the end, a robust controller was designed capable of rejecting disturbances within the operating region. Extensive simulation experiments showed that the controller outperforms a finely-tuned multi-loop feedback control strategy. The results obtained suggest that our controller is suitable for practical implementation.

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

NASA Astrophysics Data System (ADS)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

The IRIS Spool-Type Reactor Coolant Pump

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

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

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

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

Bond, Leonard J.; Bowler, John R.

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-servicemore » inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO 3-xPbTiO 3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.« less

Filamentous bacteria existence in aerobic granular reactors.

PubMed

Figueroa, M; Val del Río, A; Campos, J L; Méndez, R; Mosquera-Corral, A

2015-05-01

Filamentous bacteria are associated to biomass settling problems in wastewater treatment plants. In systems based on aerobic granular biomass they have been proposed to contribute to the initial biomass aggregation process. However, their development on mature aerobic granular systems has not been sufficiently studied. In the present research work, filamentous bacteria were studied for the first time after long-term operation (up to 300 days) of aerobic granular systems. Chloroflexi and Sphaerotilus natans have been observed in a reactor fed with synthetic wastewater. These filamentous bacteria could only come from the inoculated sludge. Thiothrix and Chloroflexi bacteria were observed in aerobic granular biomass treating wastewater from a fish canning industry. Meganema perideroedes was detected in a reactor treating wastewater from a plant processing marine products. As a conclusion, the source of filamentous bacteria in these mature aerobic granular systems fed with industrial effluents was the incoming wastewater.

Development of small-size tubular-flow continuous reactors for the analysis of operational stability of enzymes in low-water systems.

PubMed

Pirozzi, D; Halling, P J

2001-01-20

A very small-scale continuous flow reactor has been designed for use with enzymes in organic media, particularly for operational stability studies. It is constructed from fairly inexpensive components, and typically uses 5 mg of catalyst and flow rates of 1 to 5 mL/h, so only small quantities of feedstock need to be handled. The design allows control of the thermodynamic water activity of the feed, and works with temperatures up to at least 80 degrees C. The reactor has been operated with both nonpolar (octane) and polar (4-methyl-pentan-2-one) solvents, and with the more viscous solvent-free reactant mixture. It has been applied to studies of the operational stability of lipases from Chromobacterium viscosum (lyophilized powder or polypropylene-adsorbed) and Rhizomucor miehei (Lipozyme) in different experimental conditions. Transesterification of geraniol and ethylcaproate has been adopted as a model transformation.

Environmental radiation protection studies related to nuclear industries, using AMS

NASA Astrophysics Data System (ADS)

Hellborg, Ragnar; Erlandsson, Bengt; Faarinen, Mikko; Hâkansson, Helena; Hâkansson, Kjell; Kiisk, Madis; Magnusson, Carl-Erik; Persson, Per; Skog, Göran; Stenström, Kristina; Mattsson, Sören; Thornberg, Charlotte

2001-07-01

14C is produced in nuclear reactors during normal operation and part of it is continuously released into the environment. Because of the biological importance of carbon and the long physical half-life of 14C it is of interest to study these releases. The 14C activity concentrations in the air and vegetation around some Swedish as well as foreign nuclear facilities have been measured by accelerator mass spectrometry (AMS). 59Ni is produced by neutron activation in the stainless steel close to the core of a nuclear reactor. The 59Ni levels have been measured in order to be able to classify the different parts of the reactor with respect to their content of long-lived radionuclides before final storage. The technique used to measure 59Ni at a small accelerator such as the Lund facility has been developed over the past few years and material from the Swedish nuclear industry has been analyzed.

Progress towards developing neutron tolerant magnetostrictive and piezoelectric transducers

NASA Astrophysics Data System (ADS)

Reinhardt, Brian; Tittmann, Bernhard; Rempe, Joy; Daw, Joshua; Kohse, Gordon; Carpenter, David; Ames, Michael; Ostrovsky, Yakov; Ramuhalli, Pradeep; Montgomery, Robert; Chien, Hualte; Wernsman, Bernard

2015-03-01

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is not feasible to obtain high resolution/microscale data with the current state of instrumentation technology. However, ultrasound-based sensors offer the ability to obtain such data if it is demonstrated that these sensors and their associated transducers are resistant to high neutron flux, high gamma radiation, and high temperature. To address this need, the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) is funding an irradiation, led by PSU, at the Massachusetts Institute of Technology Research Reactor to test the survivability of ultrasound transducers. As part of this effort, PSU and collaborators have designed, fabricated, and provided piezoelectric and magnetostrictive transducers that are optimized to perform in harsh, high flux, environments. Four piezoelectric transducers were fabricated with either aluminum nitride, zinc oxide, or bismuth titanate as the active element that were coupled to either Kovar or aluminum waveguides and two magnetostrictive transducers were fabricated with Remendur or Galfenol as the active elements. Pulse-echo ultrasonic measurements of these transducers are made in-situ. This paper will present an overview of the test design including selection criteria for candidate materials and optimization of test assembly parameters, data obtained from both out-of-pile and in-pile testing at elevated temperatures, and an assessment based on initial data of the expected performance of ultrasonic devices in irradiation conditions.

The Angra Neutrino Project: precise measurement of {theta}{sub 13} and safeguards applications of neutrino detectors

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

Casimiro, E.; Anjos, J. C.

2009-04-20

We present an introduction to the Angra Neutrino Project. The goal of the project is to explore the use of neutrino detectors to monitor the reactor activity. The Angra Project, willl employ as neutrino sources the reactors of the nuclear power complex in Brazil, located in Angra dos Reis, some 150 Km south from the city of Rio de Janeiro. The Angra collaboration will develop and operate a low-mass neutrino detector to monitor the nuclear reactor activity, in particular to measure the reactor thermal power and the reactor fuel isotopic composition.

The Angra Neutrino Project: precise measurement of θ13 and safeguards applications of neutrino detectors

NASA Astrophysics Data System (ADS)

Casimiro, E.; Anjos, J. C.

2009-04-01

We present an introduction to the Angra Neutrino Project. The goal of the project is to explore the use of neutrino detectors to monitor the reactor activity. The Angra Project, willl employ as neutrino sources the reactors of the nuclear power complex in Brazil, located in Angra dos Reis, some 150 Km south from the city of Rio de Janeiro. The Angra collaboration will develop and operate a low-mass neutrino detector to monitor the nuclear reactor activity, in particular to measure the reactor thermal power and the reactor fuel isotopic composition.

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

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

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Computational Modeling in Plasma Processing for 300 mm Wafers

NASA Technical Reports Server (NTRS)

Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

1997-01-01

Migration toward 300 mm wafer size has been initiated recently due to process economics and to meet future demands for integrated circuits. A major issue facing the semiconductor community at this juncture is development of suitable processing equipment, for example, plasma processing reactors that can accomodate 300 mm wafers. In this Invited Talk, scaling of reactors will be discussed with the aid of computational fluid dynamics results. We have undertaken reactor simulations using CFD with reactor geometry, pressure, and precursor flow rates as parameters in a systematic investigation. These simulations provide guidelines for scaling up in reactor design.

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

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

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

Not Available

The Department of Energy (DOE) has contracted with Asea Brown Boveri-Combustion Engineering (ABB-CE) to provide information on the capability of ABB-CE`s System 80 + Advanced Light Water Reactor (ALWR) to transform, through reactor burnup, 100 metric tonnes (MT) of weapons grade plutonium (Pu) into a form which is not readily useable in weapons. This information is being developed as part of DOE`s Plutonium Disposition Study, initiated by DOE in response to Congressional action. This document, Volume 1, presents a technical description of the various elements of the System 80 + Standard Plant Design upon which the Plutonium Disposition Study wasmore » based. The System 80 + Standard Design is fully developed and directly suited to meeting the mission objectives for plutonium disposal. The bass U0{sub 2} plant design is discussed here.« less

Supercritical Water Experimental Setup for µSR

NASA Astrophysics Data System (ADS)

Liu, Guangdong; Chen, Yanggang; Morrison, Alexander H.; Koda, Akihiro; Percival, Paul W.; Ghandi, Khashayar

The Canadian design for Generation IV nuclear reactors uses supercritical water (SCW, water above its critical point of 374 °C, 221 bar (1 bar = 100 kPa)) as the coolant. Supercritical water-cooled reactors (SCWRs) are designed towards sustainability, economic benefits, improved safety, and longer lifespan. Despite the potential advantages of SCWRs, we know very little about the kinetics of radiolysis products that are formed in them because of the limitations of experimental instruments under the extreme conditions of SCW. The radiolysis products can accumulate over time and create a very corrosive environment. Our group has developed and tested an apparatus suitable for muon spin rotation (µSR) studies of water and aqueous solutions up to 550 °C and 250 bar, close to the conditions at the reactor outlet of the proposed Canadian SCWR design (625 °C and 250 bar). The reaction kinetics information obtained from our setup, together with computer simulations, will aid us in developing chemical control strategies to minimize corrosion in SCWRs.

Development of a small specimen test machine to evaluate irradiation embrittlement of fusion reactor materials

NASA Astrophysics Data System (ADS)

Ishii, T.; Ohmi, M.; Saito, J.; Hoshiya, T.; Ooka, N.; Jitsukawa, S.; Eto, M.

2000-12-01

Small specimen test techniques (SSTT) are essential to use an accelerator-driven deuterium-lithium stripping reaction neutron source for the study of fusion reactor materials because of the limitation of the available irradiation volume. A remote-controlled small punch (SP) test machine was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR) in the Japan Atomic Energy Research Institute (JAERI). This report describes the SP test method and machine for use in a hot cell, and test results on irradiated ferritic steels. The specimen was either a coupon 10×10×0.25 mm 3 or a TEM disk 3 mm in diameter by 0.25 mm in thickness. Tests can be performed at temperatures ranging from 93 to 1123 K in a vacuum or in an inert gas environment. The ductile to brittle transition temperature of the irradiated ferritic steel as determined by the SP test is also evaluated.

Experimental validation of an 8 element EMAT phased array probe for longitudinal wave generation

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

Le Bourdais, Florian, E-mail: florian.lebourdais@cea.fr; Marchand, Benoit, E-mail: florian.lebourdais@cea.fr

2015-03-31

Sodium cooled Fast Reactors (SFR) use liquid sodium as a coolant. Liquid sodium being opaque, optical techniques cannot be applied to reactor vessel inspection. This makes it necessary to develop alternative ways of assessing the state of the structures immersed in the medium. Ultrasonic pressure waves are well suited for inspection tasks in this environment, especially using pulsed electromagnetic acoustic transducers (EMAT) that generate the ultrasound directly in the liquid sodium. The work carried out at CEA LIST is aimed at developing phased array EMAT probes conditioned for reactor use. The present work focuses on the experimental validation of amore » newly manufactured 8 element probe which was designed for beam forming imaging in a liquid sodium environment. A parametric study is carried out to determine the optimal setup of the magnetic assembly used in this probe. First laboratory tests on an aluminium block show that the probe has the required beam steering capabilities.« less

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

78 FR 76600 - Proposed Subsequent Arrangement

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-18

... development of Traveling Wave Reactor (TWR) design information and related technology between the United...; REALIZING that the successful development of traveling wave reactors for the production of power for..., component or equipment) that has not yet entered into the public domain and that is especially designed...

APPLICATION ANALYSIS REPORT: HORSEHEAD RESOURCE DEVELOPMENT COMPANY INC., FLAME REACTOR TECHNOLOGY

EPA Science Inventory

A SITE demonstration of the Horsehead Resource Development (HRD) company, Inc. Flame Reactor Technology was conducted in March 1991 at the HRD facility in Monaca, Pennsylvania. For this demonstration, secondary lead smelter soda slag was treated to produce a potentially recyclabl...

An Approach for Assessing Development and Deployment Risks in the DOE Fuel Cycle Options Evaluation and Screening Study

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

Gehin, Jess C; Oakley, Brian; Worrall, Andrew

2015-01-01

Abstract One of the key objectives of the U.S. Department of Energy (DOE) Nuclear Energy R&D Roadmap is the development of sustainable nuclear fuel cycles that can improve natural resource utilization and provide solutions to the management of nuclear wastes. Recently, an evaluation and screening (E&S) of fuel cycle systems has been conducted to identify those options that provide the best opportunities for obtaining such improvements and also to identify the required research and development activities that can support the development of advanced fuel cycle options. In order to evaluate and screen the E&S study included nine criteria including Developmentmore » and Deployment Risk (D&DR). More specifically, this criterion was represented by the following metrics: Development time, development cost, deployment cost from prototypic validation to first-of-a-kind commercial, compatibility with the existing infrastructure, existence of regulations for the fuel cycle and familiarity with licensing, and existence of market incentives and/or barriers to commercial implementation of fuel cycle processes. Given the comprehensive nature of the study, a systematic approach was needed to determine metric data for the D&DR criterion, and is presented here. As would be expected, the Evaluation Group representing the once-through use of uranium in thermal reactors is always the highest ranked fuel cycle Evaluation Group for this D&DR criterion. Evaluation Groups that consist of once-through fuel cycles that use existing reactor types are consistently ranked very high. The highest ranked limited and continuous recycle fuel cycle Evaluation Groups are those that recycle Pu in thermal reactors. The lowest ranked fuel cycles are predominately continuous recycle single stage and multi-stage fuel cycles that involve TRU and/or U-233 recycle.« less

Characteristics of sludge developed under different loading conditions during UASB reactor start-up and granulation.

PubMed

Ghangrekar, M M; Asolekar, S R; Joshi, S G

2005-03-01

Sludge characteristics available inside the reactor are of vital importance to maximize advantages of UASB reactor. The organic loading rate and sludge loading rate applied during start-up are among the important parameters to govern the sludge characteristics. Effects of these loading rates on the characteristics of the sludge developed are evaluated in six laboratory scale UASB reactors. The sludge characteristics considered are VSS/SS ratio of the sludge, sludge volume index, specific gravity, settling velocity and metal contents of the sludge developed under different loading rates. The experimental results indicate that, for developing good characteristics sludge, during primary start-up from flocculent inoculum sludge, organic loading rate and sludge loading rate should be in the range of 2.0-4.5 kg COD/m3 d and 0.1-0.25 kg COD/kg VSS d, respectively (chemical oxygen demand, COD). Proper sludge granulation and higher COD removal efficiency will be achieved by these loading rates.

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

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

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01

This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operatingmore » experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.« less

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

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

The Effect of Stochastic Perturbation of Fuel Distribution on the Criticality of a One Speed Reactor and the Development of Multi-Material Multinomial Line Statistics

NASA Technical Reports Server (NTRS)

Jahshan, S. N.; Singleterry, R. C.

2001-01-01

The effect of random fuel redistribution on the eigenvalue of a one-speed reactor is investigated. An ensemble of such reactors that are identical to a homogeneous reference critical reactor except for the fissile isotope density distribution is constructed such that it meets a set of well-posed redistribution requirements. The average eigenvalue, , is evaluated when the total fissile loading per ensemble element, or realization, is conserved. The perturbation is proven to increase the reactor criticality on average when it is uniformly distributed. The various causes of the change in reactivity, and their relative effects are identified and ranked. From this, a path towards identifying the causes. and relative effects of reactivity fluctuations for the energy dependent problem is pointed to. The perturbation method of using multinomial distributions for representing the perturbed reactor is developed. This method has some advantages that can be of use in other stochastic problems. Finally, some of the features of this perturbation problem are related to other techniques that have been used for addressing similar problems.

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

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

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

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« less

Time history of diesel particle deposition in cylindrical dielectric barrier discharge reactors

NASA Astrophysics Data System (ADS)

Talebizadeh, P.; Rahimzadeh, H.; Ahmadi, G.; Brown, R.; Inthavong, K.

2016-12-01

Non-thermal plasma (NTP) treatment reactors have recently been developed for elimination of diesel particulate matter for reducing both the mass and number concentration of particles. The role of the plasma itself is obscured by the phenomenon of particle deposition on the reactor surface. Therefore, in this study, the Lagrangian particle transport model is used to simulate the dispersion and deposition of nano-particles in the range of 5 to 500 nm in a NTP reactor in the absence of an electric field. A conventional cylindrical dielectric barrier discharge reactor is selected for the analysis. Brownian diffusion, gravity and Saffman lift forces were included in the simulations, and the deposition efficiencies of different sized diesel particles were studied. The results show that for the studied particle diameters, the effect of Saffman lift is negligible and gravity only affects the motion of particles with a diameter of 500 nm or larger. Time histories of particle transport and deposition were evaluated for one-time injection and a continuous (multiple-time) injection. The results show that the number of deposited particles for one-time injection is identical to the number of deposited particles for multiple-time injections when adjusted with the shift in time. Furthermore, the maximum number of escaped particles occurs at 0.045 s after the injection for all particle diameters. The presented results show that some particle reduction previously ascribed to plasma treatment has ignored contributions from the surface deposition.

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

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

Biaxial Creep Specimen Fabrication

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

JL Bump; RF Luther

This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Navalmore » Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.« less

Dry fermentation of manure with straw in continuous plug flow reactor: Reactor development and process stability at different loading rates.

PubMed

Patinvoh, Regina J; Kalantar Mehrjerdi, Adib; Sárvári Horváth, Ilona; Taherzadeh, Mohammad J

2017-01-01

In this work, a plug flow reactor was developed for continuous dry digestion processes and its efficiency was investigated using untreated manure bedded with straw at 22% total solids content. This newly developed reactor worked successfully for 230days at increasing organic loading rates of 2.8, 4.2 and 6gVS/L/d and retention times of 60, 40 and 28days, respectively. Organic loading rates up to 4.2gVS/L/d gave a better process stability, with methane yields up to 0.163LCH 4 /gVS added /d which is 56% of the theoretical yield. Further increase of organic loading rate to 6gVS/L/d caused process instability with lower volatile solid removal efficiency and cellulose degradation. Copyright © 2016 Elsevier Ltd. All rights reserved.

Nano-metal oxides: Exposure and engineering control assessment.

PubMed

Garcia, Alberto; Eastlake, Adrienne; Topmiller, Jennifer L; Sparks, Christopher; Martinez, Kenneth; Geraci, Charles L

2017-09-01

In January 2007, the National Institute for Occupational Safety and Health (NIOSH) conducted a field study to evaluate process specific emissions during the production of ENMs. This study was performed using the nanoparticle emission assessment technique (NEAT). During this study, it was determined that ENMs were released during production and cleaning of the process reactor. Airborne concentrations of silver, nickel, and iron were found both in the employee's personal breathing zone and area samples during reactor cleaning. At the completion of this initial survey, it was suggested that a flanged attachment be added to the local exhaust ventilation system.  NIOSH re-evaluated the facility in December 2011 to assess worker exposures following an increase in production rates. This study included a fully comprehensive emissions, exposure, and engineering control evaluation of the entire process. This study made use of the nanoparticle exposure assessment technique (NEAT 2.0). Data obtained from filter-based samples and direct reading instruments indicate that reactor cleanout increased the overall particle concentration in the immediate area. However, it does not appear that these concentrations affect areas outside of the production floor. As the distance between the reactor and the sample location increased, the observed particle number concentration decreased, creating a concentration gradient with respect to the reactor. The results of this study confirm that the flanged attachment on the local exhaust ventilation system served to decrease exposure potential.  Given the available toxicological data of the metals evaluated, caution is warranted. One should always keep in mind that occupational exposure levels were not developed specifically for nanoscale particles. With data suggesting that certain nanoparticles may be more toxic than the larger counterparts of the same material; employers should attempt to control emissions of these particles at the source, to limit the potential for exposure.

Status of liquid metal fast breeder reactor fuel development in Japan

NASA Astrophysics Data System (ADS)

Katsuragawa, M.; Kashihara, H.; Akebi, M.

1993-09-01

The mixed-oxide fuel technology for a liquid metal fast breeder reactor (LMFBR) in Japan is progressing toward commercial deployment of LMFBR. Based on accumulated experience in Joyo and Monju fuel development, efforts for large scale LMFBR fuel development are devoted to improved irradiation performance, reliability and economy. This paper summarizes accomplishments, current activities and future plans for LMFBR fuel development in Japan.

New Technological Platform for the National Nuclear Energy Strategy Development

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Rachkov, V. I.

2017-12-01

The paper considers the need to update the development strategy of Russia's nuclear power industry and various approaches to the large-scale nuclear power development. Problems of making decisions on fast neutron reactors and closed nuclear fuel cycle (NFC) arrangement are discussed. The current state of the development of fast neutron reactors and closed NFC technologies in Russia is considered and major problems are highlighted.

IECEC '83; Proceedings of the Eighteenth Intersociety Energy Conversion Engineering Conference, Orlando, FL, August 21-26, 1983. Volume 1 - Thermal energy systems

NASA Astrophysics Data System (ADS)

Among the topics discussed are the nuclear fuel cycle, advanced nuclear reactor designs, developments in central status power reactors, space nuclear reactors, magnetohydrodynamic devices, thermionic devices, thermoelectric devices, geothermal systems, solar thermal energy conversion systems, ocean thermal energy conversion (OTEC) developments, and advanced energy conversion concepts. Among the specific questions covered under these topic headings are a design concept for an advanced light water breeder reactor, energy conversion in MW-sized space power systems, directionally solidified cermet electrodes for thermionic energy converters, boron-based high temperature thermoelectric materials, geothermal energy commercialization, solar Stirling cycle power conversion, and OTEC production of methanol. For individual items see A84-30027 to A84-30055

Fusion reactor blanket/shield design study

NASA Astrophysics Data System (ADS)

Smith, D. L.; Clemmer, R. G.; Harkness, S. D.; Jung, J.; Krazinski, J. L.; Mattas, R. F.; Stevens, H. C.; Youngdahl, C. K.; Trachsel, C.; Bowers, D.

1979-07-01

A joint study of Tokamak reactor first wall/blanket/shield technology was conducted to identify key technological limitations for various tritium breeding blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium breeding blanket concepts were evaluated according to the proposed coolant. The effort concentrated on evaluation of lithium and water cooled blanket designs and helium and molten salt cooled designs. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a Tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented.

Critical analysis of submerged membrane sequencing batch reactor operating conditions.

PubMed

McAdam, Ewan; Judd, Simon J; Gildemeister, René; Drews, Anja; Kraume, Matthias

2005-10-01

To evaluate the Submerged Membrane Sequencing Batch Reactor process, several short-term studies were conducted to define critical flux, membrane aeration and intermittent filtration operation. Critical flux trials indicated that as mixed liquor suspended solids increased in concentration so would the propensity for membrane fouling. Consequently in order to characterise the impact of biomass concentration increase (that develops during permeate withdrawal) upon submerged microfiltration operation, two longer term studies were conducted, one with a falling hydraulic head and another with a continuous hydraulic head (as in membrane bio-reactors). Trans membrane pressure data was used to predict the maximum possible operating periods at 10 and 62 days for the falling hydraulic head and continuous hydraulic head respectively. Further analysis revealed that falling hydraulic head operation would require 21% more aeration to maintain a consistent crossflow velocity than continuous operation and would rely on pumping for full permeate withdrawal 80% earlier. This study concluded that further optimisation would be required to make this technology technically and economically viable.

Energy spectrum of 208Pb(n,x) reactions

NASA Astrophysics Data System (ADS)

Tel, E.; Kavun, Y.; Özdoǧan, H.; Kaplan, A.

2018-02-01

Fission and fusion reactor technologies have been investigated since 1950's on the world. For reactor technology, fission and fusion reaction investigations are play important role for improve new generation technologies. Especially, neutron reaction studies have an important place in the development of nuclear materials. So neutron effects on materials should study as theoretically and experimentally for improve reactor design. For this reason, Nuclear reaction codes are very useful tools when experimental data are unavailable. For such circumstances scientists created many nuclear reaction codes such as ALICE/ASH, CEM95, PCROSS, TALYS, GEANT, FLUKA. In this study we used ALICE/ASH, PCROSS and CEM95 codes for energy spectrum calculation of outgoing particles from Pb bombardment by neutron. While Weisskopf-Ewing model has been used for the equilibrium process in the calculations, full exciton, hybrid and geometry dependent hybrid nuclear reaction models have been used for the pre-equilibrium process. The calculated results have been discussed and compared with the experimental data taken from EXFOR.

Lessons Learned about Liquid Metal Reactors from FFTF Experience

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

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

Demonstration of Robustness and Integrated Operation of a Series-Bosch System

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew; Barnett, Bill; Stanley, Christine M.; Junaedi, Christian; Vilekar, Saurabh A.; Kent, Ryan

2016-01-01

Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a RWGS reactor containing Incofoam(TradeMark) catalyst and designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith(TradeMark) technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The Microlith(TradeMark) RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with the Incofoam(TradeMark) RWGS reactor. Separately, in 2015, a fully integrated demonstration of an S-Bosch system was conducted. In an effort to mitigate risk, a second integrated test was conducted to evaluate the effect of membrane failure on a closed-loop Bosch system. Here, we report and discuss the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level. 1

The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

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

Hollaway, W.R.

1991-08-01

If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issuemore » through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).« less

Development of a Gas Filled Magnet spectrometer coupled with the Lohengrin spectrometer for fission study

NASA Astrophysics Data System (ADS)

Kessedjian, G.; Chebboubi, A.; Faust, H.; Köster, U.; Materna, T.; Sage, C.; Serot, O.

2013-03-01

The accurate knowledge of the fission of actinides is necessary for studies of innovative nuclear reactor concepts. The fission yields have a direct influence on the evaluation of the fuel inventory or the reactor residual power after shutdown. A collaboration between the ILL, LPSC and CEA has developed a measurement program on fission fragment distributions at ILL in order to measure the isotopic and isomeric yields. The method is illustrated using the 233U(n,f)98Y reaction. However, the extracted beam from the Lohengrin spectrometer is not isobaric ions which limits the low yield measurements. Presently, the coupling of the Lohengrin spectrometer with a Gas Filled Magnet (GFM) is studied at the ILL in order to define and validate the enhanced purification of the extracted beam. This work will present the results of the spectrometer characterisation, along with a comparison with a dedicated Monte Carlo simulation especially developed for this purpose.

Evaluation of the performance of high temperature conversion reactors for compound-specific oxygen stable isotope analysis.

PubMed

Hitzfeld, Kristina L; Gehre, Matthias; Richnow, Hans-Hermann

2017-05-01

In this study conversion conditions for oxygen gas chromatography high temperature conversion (HTC) isotope ratio mass spectrometry (IRMS) are characterised using qualitative mass spectrometry (IonTrap). It is shown that physical and chemical properties of a given reactor design impact HTC and thus the ability to accurately measure oxygen isotope ratios. Commercially available and custom-built tube-in-tube reactors were used to elucidate (i) by-product formation (carbon dioxide, water, small organic molecules), (ii) 2nd sources of oxygen (leakage, metal oxides, ceramic material), and (iii) required reactor conditions (conditioning, reduction, stability). The suitability of the available HTC approach for compound-specific isotope analysis of oxygen in volatile organic molecules like methyl tert-butyl ether is assessed. Main problems impeding accurate analysis are non-quantitative HTC and significant carbon dioxide by-product formation. An evaluation strategy combining mass spectrometric analysis of HTC products and IRMS 18 O/ 16 O monitoring for future method development is proposed.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

DOE PAGES

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

2014-06-30

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

Plasma characterization studies for materials processing

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

Pfender, E.; Heberlein, J.

New applications for plasma processing of materials require a more detailed understanding of the fundamental processes occurring in the processing reactors. We have developed reactors offering specific advantages for materials processing, and we are using modeling and diagnostic techniques for the characterization of these reactors. The emphasis is in part set by the interest shown by industry pursuing specific plasma processing applications. In this paper we report on the modeling of radio frequency plasma reactors for use in materials synthesis, and on the characterization of the high rate diamond deposition process using liquid precursors. In the radio frequency plasma torchmore » model, the influence of specific design changes such as the location of the excitation coil on the enthalpy flow distribution is investigated for oxygen and air as plasma gases. The diamond deposition with liquid precursors has identified the efficient mass transport in form of liquid droplets into the boundary layer as responsible for high growth, and the chemical properties of the liquid for the film morphology.« less

Investigating the kinetics of the enzymatic depolymerization of polygalacturonic acid in continuous UF-membrane reactors.

PubMed

Gallifuoco, Alberto; Cantarella, Maria; Marucci, Mariagrazia

2007-01-01

A stirred tank membrane reactor is used to study the kinetics of polygalacturonic acid (PGA) enzymatic hydrolysis. The reactor operates in semicontinuous configuration: the native biopolymer is loaded at the initial time and the system is continuously fed with the buffer. The effect of retention time (from 101 to 142 min) and membrane molecular weight cutoff (from 1 to 30 kDa) on the rate of permeable oligomers production is investigated. Reaction products are clustered in two different classes, those sized below the membrane cutoff and those above. The reducing power measured in the permeate is used as an estimate of total product concentration. The characteristic breakdown times range from 40 to 100 min. The overall kinetics obeys a first-order law with a characteristic time estimated to 24 min. New mathematical data handling are developed and illustrated using the experimental data obtained. Finally, the body of the experimental results suggests useful indications (reactor productivity, breakdown induction period) for implementing the bioprocess at the industrial scale.

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

NASA Astrophysics Data System (ADS)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

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.

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

Underwood, R.P.

As part of the DOE-sponsored contract Synthesis of Dimethyl Ether and Alternative Fuels in the Liquid Phase from Coal-Derived Syngas'' experimental evaluations of the one-step synthesis of alternative fuels were carried out. The objective of this work was to develop novel processes for converting coal-derived syngas to fuels or fuel additives. Building on a technology base acquired during the development of the Liquid Phase Methanol (LPMEOH) process, this work focused on the development of slurry reactor based processes. The experimental investigations, which involved bench-scale reactor studies, focused primarily on three areas: (1) One-step, slurry-phase syngas conversion to hydrocarbons or methanol/hydrocarbonmore » mixtures using a mixture of methanol synthesis catalyst and methanol conversion catalyst in the same slurry reactor. (2) Slurry-phase conversion of syngas to mixed alcohols using various catalysts. (3) One-step, slurry-phase syngas conversion to mixed ethers using a mixture of mixed alcohols synthesis catalyst and dehydration catalyst in the same slurry reactor. The experimental results indicate that, of the three types of processes investigated, slurry phase conversion of syngas to mixed alcohols shows the most promise for further process development. Evaluations of various mixed alcohols catalysts show that a cesium-promoted Cu/ZnO/Al[sub 2]O[sub 3] methanol synthesis catalyst, developed in Air Products' laboratories, has the highest performance in terms of rate and selectivity for C[sub 2+]-alcohols. In fact, once-through conversion at industrially practical reaction conditions yielded a mixed alcohols product potentially suitable for direct gasoline blending. Moreover, an additional attractive aspect of this catalyst is its high selectivity for branched alcohols, potential precursors to iso-olefins for use in etherification.« less

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

Underwood, R.P.

As part of the DOE-sponsored contract ``Synthesis of Dimethyl Ether and Alternative Fuels in the Liquid Phase from Coal-Derived Syngas`` experimental evaluations of the one-step synthesis of alternative fuels were carried out. The objective of this work was to develop novel processes for converting coal-derived syngas to fuels or fuel additives. Building on a technology base acquired during the development of the Liquid Phase Methanol (LPMEOH) process, this work focused on the development of slurry reactor based processes. The experimental investigations, which involved bench-scale reactor studies, focused primarily on three areas: (1) One-step, slurry-phase syngas conversion to hydrocarbons or methanol/hydrocarbonmore » mixtures using a mixture of methanol synthesis catalyst and methanol conversion catalyst in the same slurry reactor. (2) Slurry-phase conversion of syngas to mixed alcohols using various catalysts. (3) One-step, slurry-phase syngas conversion to mixed ethers using a mixture of mixed alcohols synthesis catalyst and dehydration catalyst in the same slurry reactor. The experimental results indicate that, of the three types of processes investigated, slurry phase conversion of syngas to mixed alcohols shows the most promise for further process development. Evaluations of various mixed alcohols catalysts show that a cesium-promoted Cu/ZnO/Al{sub 2}O{sub 3} methanol synthesis catalyst, developed in Air Products` laboratories, has the highest performance in terms of rate and selectivity for C{sub 2+}-alcohols. In fact, once-through conversion at industrially practical reaction conditions yielded a mixed alcohols product potentially suitable for direct gasoline blending. Moreover, an additional attractive aspect of this catalyst is its high selectivity for branched alcohols, potential precursors to iso-olefins for use in etherification.« less

Technology development for cobalt F-T catalysts. Quarterly technical progress report number 10, January 1--March 31, 1995

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

Singleton, A.H.

1995-06-28

The goal of this project is the development of a commercially-viable, cobalt-based Fischer-Tropsch (F-T) catalyst for use in a slurry bubble column reactor. The major objectives of this work are (1) to develop a cobalt-based F-T catalyst with low (< 5%) methane selectivity, (2) to develop a cobalt-based F-T catalyst with water-gas shift activity, and (3) to combine both these improvements into one catalyst. The project consists of five major tasks: catalyst development; catalyst testing; catalyst reproducibility tests; catalyst aging tests; and preliminary design and cost estimate for a demonstrate scale catalyst production facility. Technical accomplishments during this reporting periodmore » include the following. It appears that the higher activity obtained for the catalysts prepared using an organic solution and reduced directly without prior calcination was the result of higher dispersions obtained under such pretreatment. A Ru-promoted Co catalyst on alumina with 30% Co loading exhibited a 4-fold increase in dispersion and a 2-fold increase in activity in the fixed-bed reactor from that obtained with the non-promoted catalyst. Several reactor runs have again focused on pushing conversion to higher levels. The maximum conversion obtained has been 49.7% with 26g catalyst. Further investigations of the effect of reaction temperature on the performance of Co catalysts during F-T synthesis were started using a low activity catalyst and one of the most active catalysts. The three 1 kg catalyst batches prepared by Calsicat for the reproducibility and aging studies were tested in both the fixed-bed and slurry bubble column reactors under the standard reaction conditions. The effects of adding various promoters to some cobalt catalysts have also been addressed. Results are presented and discussed.« less

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

NASA Astrophysics Data System (ADS)

Hançerliogulları, Aybaba; Cini, Mesut

2013-10-01

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

Development of TDLAS sensor for diagnostics of CO, H2O and soot concentrations in reactor core of pilot-scale gasifier

NASA Astrophysics Data System (ADS)

Sepman, A.; Ögren, Y.; Gullberg, M.; Wiinikka, H.

2016-02-01

This paper reports on the development of the tunable diode laser absorption spectroscopy sensor near 4350 cm-1 (2298 nm) for measurements of CO and H2O mole fractions and soot volume fraction under gasification conditions. Due to careful selection of the molecular transitions [CO ( υ″ = 0 → υ' = 2) R34-R36 and H2O at 4349.337 cm-1], a very weak (negligible) sensitivity of the measured species mole fractions to the temperature distribution inside the high-temperature zone (1000 K < T < 1900 K) of the gasification process is achieved. The selected transitions are covered by the tuning range of single diode laser. The CO and H2O concentrations measured in flat flames generally agree better than 10 % with the results of 1-D flame simulations. Calibration-free absorption measurements of studied species in the reactor core of atmospheric pilot-scale entrained-flow gasifier operated at 0.1 MW power are reported. Soot concentration is determined from the measured broadband transmittance. The estimated uncertainties in the reactor core CO and H2O measurements are 15 and 20 %, respectively. The reactor core average path CO mole fractions are in quantitative agreement with the µGC CO concentrations sampled at the gasifier output.

Technology development for iron Fischer-Tropsch catalysts

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

O`Brien, R.J.; Raje, A.; Keogh, R.A.

1995-12-31

The objective of this research project is to develop the technology for the production of physically robust iron-based Fischer-Tropsch catalysts that have suitable activity, selectivity and stability to be used in the slurry phase synthesis reactor development. The catalysts that are developed shall be suitable for testing in the Advanced Fuels Development Facility at LaPorte, Texas, to produce either low-or high-alpha product distributions. Previous work by the offeror has produced a catalyst formulation that is 1.5 times as active as the {open_quotes}standard-catalyst{close_quotes} developed by German workers for slurry phase synthesis. In parallel, work will be conducted to design a high-alphamore » iron catalyst this is suitable for slurry phase synthesis. Studies will be conducted to define the chemical phases present at various stages of the pretreatment and synthesis stages and to define the course of these changes. The oxidation/reduction cycles that are anticipated to occur in large, commercial reactors will be studied at the laboratory scale. Catalyst performance will be determined for catalysts synthesized in this program for activity, selectivity and aging characteristics.« less

Development of a real-time simulation tool towards self-consistent scenario of plasma start-up and sustainment on helical fusion reactor FFHR-d1

NASA Astrophysics Data System (ADS)

Goto, T.; Miyazawa, J.; Sakamoto, R.; Suzuki, Y.; Suzuki, C.; Seki, R.; Satake, S.; Huang, B.; Nunami, M.; Yokoyama, M.; Sagara, A.; the FFHR Design Group

2017-06-01

This study closely investigates the plasma operation scenario for the LHD-type helical reactor FFHR-d1 in view of MHD equilibrium/stability, neoclassical transport, alpha energy loss and impurity effect. In 1D calculation code that reproduces the typical pellet discharges in LHD experiments, we identify a self-consistent solution of the plasma operation scenario which achieves steady-state sustainment of the burning plasma with a fusion gain of Q ~ 10 was found within the operation regime that has been already confirmed in LHD experiment. The developed calculation tool enables systematic analysis of the operation regime in real time.

Process of forming catalytic surfaces for wet oxidation reactions

NASA Technical Reports Server (NTRS)

Jagow, R. B. (Inventor)

1977-01-01

A wet oxidation process was developed for oxidizing waste materials, comprising dissolved ruthenium salt in a reactant feed stream containing the waste materials. The feed stream is introduced into a reactor, and the reactor contents are then raised to an elevated temperature to effect deposition of a catalytic surface of ruthenium black on the interior walls of the reactor. The feed stream is then maintained in the reactor for a period of time sufficient to effect at least partial oxidation of the waste materials.

NEUTRONIC REACTOR CONTROL

DOEpatents

Metcalf, H.E.

1958-10-14

Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

2010-10-01

The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

Effect of initiator concentration to low-density polyethylene production in a tubular reactor

NASA Astrophysics Data System (ADS)

Azmi, A.; Aziz, N.

2016-11-01

Low-density polyethylene (LDPE) is one of the most widely used polymers in the world, which is produced in high-capacity tubular and autoclave reactors. As the LDPE industry turn into more competitive and its market profit margins become tighter, manufacturers have to develop solutions to debottleneck the reactor output while abiding to the stringent product specification. A single polyolefin plant producing ten to forty grades of LDPE with various melt flow index (MFI), therefore understanding the reaction mechanism, the operating conditions as well as the dynamic behavior of tubular reactor is essential before any improvement can take place. In the present work, a steady state mathematical model representing a tubular reactor for the production of LDPE is simulated using MATLAB R2015a®. The model developed is a function of feed inlet, reactor jacket, single initiator injector and outlet stream. Analysis on the effect of initiator concentration (CI) shows sudden declining trend of initiator's concentration which indicates that all of the initiators are exhausted after polymerization reaction and no further reaction occur from this point onwards. Furthermore, the results demonstrate that the concentration of initiator gives significant impact on reactor temperature's profile and monomer conversion rate, since higher initiator concentration promotes greater polymerization rate, and therefore leads to higher monomer conversion throughput.

The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

NASA Astrophysics Data System (ADS)

Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

2013-10-01

A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

Modeling moving systems with RELAP5-3D

DOE PAGES

Mesina, G. L.; Aumiller, David L.; Buschman, Francis X.; ...

2015-12-04

RELAP5-3D is typically used to model stationary, land-based reactors. However, it can also model reactors in other inertial and accelerating frames of reference. By changing the magnitude of the gravitational vector through user input, RELAP5-3D can model reactors on a space station or the moon. The field equations have also been modified to model reactors in a non-inertial frame, such as occur in land-based reactors during earthquakes or onboard spacecraft. Transient body forces affect fluid flow in thermal-fluid machinery aboard accelerating crafts during rotational and translational accelerations. It is useful to express the equations of fluid motion in the acceleratingmore » frame of reference attached to the moving craft. However, careful treatment of the rotational and translational kinematics is required to accurately capture the physics of the fluid motion. Correlations for flow at angles between horizontal and vertical are generated via interpolation where no experimental studies or data exist. The equations for three-dimensional fluid motion in a non-inertial frame of reference are developed. As a result, two different systems for describing rotational motion are presented, user input is discussed, and an example is given.« less

A novel approach for toluene gas treatment using a downflow hanging sponge reactor.

PubMed

Yamaguchi, Tsuyoshi; Nakamura, Syoichiro; Hatamoto, Masashi; Tamura, Eisuke; Tanikawa, Daisuke; Kawakami, Shuji; Nakamura, Akinobu; Kato, Kaoru; Nagano, Akihiro; Yamaguchi, Takashi

2018-05-01

A novel gas-scrubbing bioreactor based on a downflow hanging sponge (DHS) reactor was developed as a new volatile organic compound (VOC) treatment system. In this study, the effects of varying the space velocity and gas/liquid ratio were investigated to assess the effectiveness of using toluene gas as a model VOC. Under optimal conditions, the toluene removal rate was greater than 80%, and the maximum elimination capacity was observed at approximately 13 g-C m -3  h -1 . The DHS reactor demonstrated slight pressure loss (20 Pa) and a high concentration of suspended solids (up to 30,000 mg/L-sponge). Cloning analysis of the 16S rRNA and functional genes of toluene degradation pathways (tmoA, todC, tbmD, xylA, and bssA) revealed that the clones belonging to the toluene-degrading bacterium Pseudomonas putida constituted the predominant species detected at the bottom of the DHS reactor. The toluene-degrading bacteria Pseudoxanthomonas spadix and Pseudomonas sp. were also detected by tmoA- and todC-targeted cloning analyses, respectively. These results demonstrate the potential for the industrial application of this novel DHS reactor for toluene gas treatment.

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

Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id

A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation atmore » inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.« less

Microreactor Development for Martian In-Situ Propellant Production

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

Holladay, Jamie D.; Brooks, Kriston P.; Wegeng, Robert S.

2007-01-30

The second part of the Martian In-situ Propellant Production (MIPPS) system reviews the development of the Sabatier Reactor (SR). The microchannel SR had integrated cooling channels as well as reaction channels. It was <100cc in volume. The reactor utilized a proprietary catalyst. When operated at 400oC 70-80% CO2 conversion was achieved which enabled ~0.0125 kg CH4/hr production, or 1/8th the target mission. The modular design of the microchannel reactors would enable simple scale up to full scale production for the proposed mission.

Standardized reactors for the study of medical biofilms: a review of the principles and latest modifications.

PubMed

Gomes, Inês B; Meireles, Ana; Gonçalves, Ana L; Goeres, Darla M; Sjollema, Jelmer; Simões, Lúcia C; Simões, Manuel

2018-08-01

Biofilms can cause severe problems to human health due to the high tolerance to antimicrobials; consequently, biofilm science and technology constitutes an important research field. Growing a relevant biofilm in the laboratory provides insights into the basic understanding of the biofilm life cycle including responses to antibiotic therapies. Therefore, the selection of an appropriate biofilm reactor is a critical decision, necessary to obtain reproducible and reliable in vitro results. A reactor should be chosen based upon the study goals and a balance between the pros and cons associated with its use and operational conditions that are as similar as possible to the clinical setting. However, standardization in biofilm studies is rare. This review will focus on the four reactors (Calgary biofilm device, Center for Disease Control biofilm reactor, drip flow biofilm reactor, and rotating disk reactor) approved by a standard setting organization (ASTM International) for biofilm experiments and how researchers have modified these standardized reactors and associated protocols to improve the study and understanding of medical biofilms.