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

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.

Continuous AE crack monitoring of a dissimilar metal weldment at Limerick Unit 1

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

Hutton, P.H.; Friesel, M.A.; Dawson, J.F.

1993-12-01

Acoustic emission (AE) technology for continuous surveillance of a reactor component(s) to detect crack initiation and/or crack growth has been developed at Pacific Northwest Laboratory (PNL). The technology was validated off-reactor in several major tests, but it had not been validated by monitoring crack growth on an operating reactor system. A flaw indication was identified during normal inservice inspection of piping at Philadelphia Electric Company (PECO) Limerick Unit 1 reactor during the 1989 refueling outage. Evaluation of the flaw indication showed that it could remain in place during the subsequent fuel cycle without compromising safety. The existence of this flawmore » indication offered a long sought opportunity to validate AE surveillance to detect and evaluate crack growth during reactor operation. AE instrumentation was installed by PNL and PECO to monitor the flaw indication during two complete fuel cycles. This report discusses the results obtained from the AE monitoring over the period May 1989 to March 1992 (two fuel cycles).« less

Transient Testing of Nuclear Fuels and Materials in the United States

NASA Astrophysics Data System (ADS)

Wachs, Daniel M.

2012-12-01

The United States has established that transient irradiation testing is needed to support advanced light water reactors fuel development. The U.S. Department of Energy (DOE) has initiated an effort to reestablish this capability. Restart of the Transient Testing Reactor (TREAT) facility located at the Idaho National Laboratory (INL) is being considered for this purpose. This effort would also include the development of specialized test vehicles to support stagnant capsule and flowing loop tests as well as the enhancement of postirradiation examination capabilities and remote device assembly capabilities at the Hot Fuel Examination Facility. It is anticipated that the capability will be available to support testing by 2018, as required to meet the DOE goals for the development of accident-tolerant LWR fuel designs.

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

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

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

A Summary of Closed Brayton Cycle Development Activities at NASA

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2009-01-01

NASA has been involved in the development of Closed Brayton Cycle (CBC) power conversion technology since the 1960's. CBC systems can be coupled to reactor, isotope, or solar heat sources and offer the potential for high efficiency, long life, and scalability to high power. In the 1960's and 1970's, NASA and industry developed the 10 kW Brayton Rotating Unit (BRU) and the 2 kW mini-BRU demonstrating technical feasibility and performance, In the 1980's, a 25 kW CBC Solar Dynamic (SD) power system option was developed for Space Station Freedom and the technology was demonstrated in the 1990's as part of the 2 kW SO Ground Test Demonstration (GTD). Since the early 2000's, NASA has been pursuing CBC technology for space reactor applications. Before it was cancelled, the Jupiter Icy Moons Orbiter (HMO) mission was considering a 100 kWclass CBC system coupled to a gas-cooled fission reactor. Currently, CBC technology is being explored for Fission Surface Power (FSP) systems to provide base power on the moon and Mars. These recent activities have resulted in several CBC-related technology development projects including a 50 kW Alternator Test Unit, a 20 kW Dual Brayton Test Loop, a 2 kW Direct Drive Gas Brayton Test Loop, and a 12 kW FSP Power Conversion Unit design.

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

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

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

Not Available

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

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

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

76 FR 18586 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of The ACRS Subcommittee on United...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-04

... as technical reports related to the Gas Turbine Generator design. The Subcommittee will hear... Subcommittee on United States-Advanced Pressurized Water Reactor (US-APWR); Notice of Meeting The ACRS Subcommittee on United States-Advanced Pressurized Water Reactor (US-APWR) will hold a meeting on April 22...

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

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

Mynatt, F.R.

1987-03-18

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

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

Nuclear power plant 5,000 to 10,000 kilowatts

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

Not Available

The purpose of this proposal is to present a suggested program for the development of an Aqueous Homogeneous Reactor Power Plant for the production of power in the 5000 to 10,000 kilowatt range under the terms of the Atomic Energy Commission's invitation of September 21, 1955. It envisions a research and development program prior to finalizing fabricating commitments of full scale components for the purpose of proving mechanical and hydraulic operating and chemical processing feasibility with the expectation that such preliminary effort will assure the contruction of the reactor at the lowest cost and successful operation at the earliest date.more » It proposes the construction of a reactor for an eventual net electrical output of ten megawatts but initially in conjunction with a five megawatt turbo-generating unit. This unit would be constructed at the site of the existing Hersey diesel generating plant of the Wolverine Electric Cooperative approximately ten miles north of Big Rapids, Michigan.« less

The thermal circuit of a nuclear power station's unit built around a supercritical-pressure water-cooled reactor

NASA Astrophysics Data System (ADS)

Silin, V. A.; Zorin, V. M.; Tagirov, A. M.; Tregubova, O. I.; Belov, I. V.; Povarov, P. V.

2010-12-01

Main results obtained from calculations of the steam generator and thermal circuit of the steam turbine unit for a nuclear power unit with supercritical-pressure water coolant and integral layout are presented. The obtained characteristics point to the advisability of carrying out further developments of this promising nuclear power technology.

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

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

Rempe, Joy L.; Knudson, Darrell L.

2015-02-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken bymore » these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

Coupled field effects in BWR stability simulations using SIMULATE-3K

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

Borkowski, J.; Smith, K.; Hagrman, D.

1996-12-31

The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17.

Analysis of dosimetry from the H.B. Robinson unit 2 pressure vessel benchmark using RAPTOR-M3G and ALPAN

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

Fischer, G.A.

2011-07-01

Document available in abstract form only, full text of document follows: The dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark is analyzed with a suite of Westinghouse-developed codes and data libraries. The radiation transport from the reactor core to the surveillance capsule and ex-vessel locations is performed by RAPTOR-M3G, a parallel deterministic radiation transport code that calculates high-resolution neutron flux information in three dimensions. The cross-section library used in this analysis is the ALPAN library, an Evaluated Nuclear Data File (ENDF)/B-VII.0-based library designed for reactor dosimetry and fluence analysis applications. Dosimetry is evaluated with the industry-standard SNLRMLmore » reactor dosimetry cross-section data library. (authors)« less

Nuclear Power Plants. Revised.

ERIC Educational Resources Information Center

Lyerly, Ray L.; Mitchell, Walter, III

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

Modifications to the NRAD Reactor, 1977 to present

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

Weeks, A.A.; Pruett, D.P.; Heidel, C.C.

1986-01-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems.« less

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.

A diesel fuel processor for fuel-cell-based auxiliary power unit applications

NASA Astrophysics Data System (ADS)

Samsun, Remzi Can; Krekel, Daniel; Pasel, Joachim; Prawitz, Matthias; Peters, Ralf; Stolten, Detlef

2017-07-01

Producing a hydrogen-rich gas from diesel fuel enables the efficient generation of electricity in a fuel-cell-based auxiliary power unit. In recent years, significant progress has been achieved in diesel reforming. One issue encountered is the stable operation of water-gas shift reactors with real reformates. A new fuel processor is developed using a commercial shift catalyst. The system is operated using optimized start-up and shut-down strategies. Experiments with diesel and kerosene fuels show slight performance drops in the shift reactor during continuous operation for 100 h. CO concentrations much lower than the target value are achieved during system operation in auxiliary power unit mode at partial loads of up to 60%. The regeneration leads to full recovery of the shift activity. Finally, a new operation strategy is developed whereby the gas hourly space velocity of the shift stages is re-designed. This strategy is validated using different diesel and kerosene fuels, showing a maximum CO concentration of 1.5% at the fuel processor outlet under extreme conditions, which can be tolerated by a high-temperature PEFC. The proposed operation strategy solves the issue of strong performance drop in the shift reactor and makes this technology available for reducing emissions in the transportation sector.

Materials and fabrication technology of modules intended for irradiation tests of blanket tritium-breeding zones in Russian fusion reactor projects

NASA Astrophysics Data System (ADS)

Kapychev, V.; Davydov, D.; Gorokhov, V.; Ioltukhovskiy, A.; Kazennov, Yu; Tebus, V.; Frolov, V.; Shikov, A.; Shishkov, N.; Kovalenko, V.; Shishkin, N.; Strebkov, Yu

2000-12-01

This paper surveys the modules and materials of blanket tritium-breeding zones developed in the Russian Federation for fusion reactors. Synthesis of lithium orthosilicate, metasilicate and aluminate, fabrication of ceramic pellets and pebbles and experimental reactor units are described. Results of tritium extraction kinetics under irradiation in a water-graphite reactor at a thermal neutron flux of 5×10 13 neutron/(s cm2) are considered. At the present time, development and fabrication of lithium orthosilicate-beryllium modules of the tritium-breeding zone (TBZ), have been carried out within the framework of the ITER and DEMO projects. Two modules containing orthosilicate pellets, porous beryllium and beryllium pebbles are suggested for irradiation tests in the temperature range of 350-700°C. Technical problems associated with manufacturing of the modules are discussed.

Evaluation of selected chemical processes for production of low-cost silicon, phase 3

NASA Technical Reports Server (NTRS)

Blocher, J. M., Jr.; Browning, M. F.; Seifert, D. A.

1981-01-01

A Process Development Unit (PDU), which consisted of the four major units of the process, was designed, installed, and experimentally operated. The PDU was sized to 50MT/Yr. The deposition took place in a fluidized bed reactor. As a consequences of the experiments, improvements in the design an operation of these units were undertaken and their experimental limitations were partially established. A parallel program of experimental work demonstrated that Zinc can be vaporized for introduction into the fluidized bed reactor, by direct induction-coupled r.f. energy. Residual zinc in the product can be removed by heat treatment below the melting point of silicon. Current efficiencies of 94 percent and above, and power efficiencies around 40 percent are achievable in the laboratory-scale electrolysis of ZnCl2.

Safety system augmentation at Russian nuclear power plants

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

Scerbo, J.A.; Satpute, S.N.; Donkin, J.Y.

1996-12-31

This paper describes the design and procurement of a Class IE DC power supply system to upgrade plant safety at the Kola Nuclear Power Plant (NPP). Kola NPP is located above the Arctic circle at Polyarnie Zorie, Murmansk, Russia. Kola NPP consists of four units. Units 1 and 2 have VVER-440/230 type reactors: Units 3 and 4 have VVER-440/213 type reactors. The VVER-440 reactor design is similar to the pressurized water reactor design used in the US. This project provided redundant, Class 1E DC station batteries and DC switchboards for Kola NPP, Units 1 and 2. The new DC powermore » supply system was designed and procured in compliance with current nuclear design practices and requirements. Technical issues that needed to be addressed included reconciling the requirements in both US and Russian codes and satisfying the requirements of the Russian nuclear regulatory authority. Close interface with ATOMENERGOPROEKT (AEP), the Russian design organization, KOLA NPP plant personnel, and GOSATOMNADZOR (GAN), the Russian version of US Nuclear Regulatory Commission, was necessary to develop a design that would assure compliance with current Russian design requirements. Hence, this project was expected to serve as an example for plant upgrades at other similar VVER-440 nuclear plants. In addition to technical issues, the project needed to address language barriers and the logistics of shipping equipment to a remote section of the Former Soviet Union (FSU). This project was executed by Burns and Roe under the sponsorship of the US DOE as part of the International Safety Program (INSP). The INSP is a comprehensive effort, in cooperation with partners in other countries, to improve nuclear safety worldwide. A major element within the INSP is the improvement of the safety of Soviet-designed nuclear reactors.« less

The United States Naval Nuclear Propulsion Program - Over 151 Million Miles Safely Steamed on Nuclear Power

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

None, None

NNSA’s third mission pillar is supporting the U.S. Navy’s ability to protect and defend American interests across the globe. The Naval Reactors Program remains at the forefront of technological developments in naval nuclear propulsion and ensures a commanding edge in warfighting capabilities by advancing new technologies and improvements in naval reactor performance and reliability. In 2015, the Naval Nuclear Propulsion Program pioneered advances in nuclear reactor and warship design – such as increasing reactor lifetimes, improving submarine operational effectiveness, and reducing propulsion plant crewing. The Naval Reactors Program continued its record of operational excellence by providing the technical expertise requiredmore » to resolve emergent issues in the Nation’s nuclear-powered fleet, enabling the Fleet to safely steam more than two million miles. Naval Reactors safely maintains, operates, and oversees the reactors on the Navy’s 82 nuclear-powered warships, constituting more than 45 percent of the Navy’s major combatants.« less

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

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

Gauntt, Randall O.; Mattie, Patrick D.

Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this studymore » was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.« less

Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

NASA Technical Reports Server (NTRS)

Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

2010-01-01

Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

Flat-plate collector research area: Silicon material task

NASA Technical Reports Server (NTRS)

Lutwack, R.

1982-01-01

Silane decomposition in a fluidized-bed reactor (FBR) process development unit (PDU) to make semiconductor-grade Si is reviewed. The PDU was modified by installation of a new heating system to provide the required temperature profile and better control, and testing was resumed. A process for making trichlorosilane by the hydrochlorination of metallurgical-grade Si and silicon tetrachloride is reported. Fabrication and installation of the test system employing a new 2-in.-dia reactor was completed. A process that converts trichlorosilane to dichlorosilane (DCS), which is reduced by hydrogen to make Si by a chemical vapor deposition step in a Siemens-type reactor is described. Testing of the DCS PDU integraled with Si deposition reactors continued. Experiments in a 2-in.-dia reactor to define the operating window and to investigate the Si deposition kinetics were completed.

Analysis of the Browns Ferry Unit 3 irradiation experiments. Final report

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

Simmons, G.L.

1984-11-01

The results of the analysis of two experiments performed at the Browns Ferry-3 reactor are presented. These calculations utilize state-of-the-art neutron transport techniques and a new neutron cross-section library that has been developed for LWR applications. The calculations agree well with the experimental data obtained in irradiations inside the reactor vessel. For the measurements performed in the reactor cavity, the calculations agree well at the reactor midplane. Accurate determination of the axial distribution of the neutron fluence in the reactor cavity depends on having a concise representation of the axial-void distribution in the core. Detailed data are presented describing themore » procedures used in the generation of the new cross-section library that has been named SAILOR. This library is available from the Radiation-Shielding Information Center.« 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

75 FR 13142 - Florida Power and Light Company; Turkey Point, Units 3 and 4; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-18

... Light Company; Turkey Point, Units 3 and 4; Exemption 1.0 Background Florida Power and Light Company... ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water... reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of...

Real-time pH monitoring of industrially relevant enzymatic reactions in a microfluidic side-entry reactor (μSER) shows potential for pH control.

PubMed

Gruber, Pia; Marques, Marco P C; Sulzer, Philipp; Wohlgemuth, Roland; Mayr, Torsten; Baganz, Frank; Szita, Nicolas

2017-06-01

Monitoring and control of pH is essential for the control of reaction conditions and reaction progress for any biocatalytic or biotechnological process. Microfluidic enzymatic reactors are increasingly proposed for process development, however typically lack instrumentation, such as pH monitoring. We present a microfluidic side-entry reactor (μSER) and demonstrate for the first time real-time pH monitoring of the progression of an enzymatic reaction in a microfluidic reactor as a first step towards achieving pH control. Two different types of optical pH sensors were integrated at several positions in the reactor channel which enabled pH monitoring between pH 3.5 and pH 8.5, thus a broader range than typically reported. The sensors withstood the thermal bonding temperatures typical of microfluidic device fabrication. Additionally, fluidic inputs along the reaction channel were implemented to adjust the pH of the reaction. Time-course profiles of pH were recorded for a transketolase and a penicillin G acylase catalyzed reaction. Without pH adjustment, the former showed a pH increase of 1 pH unit and the latter a pH decrease of about 2.5 pH units. With pH adjustment, the pH drop of the penicillin G acylase catalyzed reaction was significantly attenuated, the reaction condition kept at a pH suitable for the operation of the enzyme, and the product yield increased. This contribution represents a further step towards fully instrumented and controlled microfluidic reactors for biocatalytic process development. © 2017 The Authors. Biotechnology Journal published by WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Fuel processing in integrated micro-structured heat-exchanger reactors

NASA Astrophysics Data System (ADS)

Kolb, G.; Schürer, J.; Tiemann, D.; Wichert, M.; Zapf, R.; Hessel, V.; Löwe, H.

Micro-structured fuel processors are under development at IMM for different fuels such as methanol, ethanol, propane/butane (LPG), gasoline and diesel. The target application are mobile, portable and small scale stationary auxiliary power units (APU) based upon fuel cell technology. The key feature of the systems is an integrated plate heat-exchanger technology which allows for the thermal integration of several functions in a single device. Steam reforming may be coupled with catalytic combustion in separate flow paths of a heat-exchanger. Reactors and complete fuel processors are tested up to the size range of 5 kW power output of a corresponding fuel cell. On top of reactor and system prototyping and testing, catalyst coatings are under development at IMM for numerous reactions such as steam reforming of LPG, ethanol and methanol, catalytic combustion of LPG and methanol, and for CO clean-up reactions, namely water-gas shift, methanation and the preferential oxidation of carbon monoxide. These catalysts are investigated in specially developed testing reactors. In selected cases 1000 h stability testing is performed on catalyst coatings at weight hourly space velocities, which are sufficiently high to meet the demands of future fuel processing reactors.

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.

Development of advanced strain diagnostic techniques for reactor environments.

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

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.

2013-02-01

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding.more » During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.« less

75 FR 63867 - DTE Energy; Enrico Fermi Atomic Power Plant Unit 1, Exemption From Certain Security Requirements

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-18

... procedures, physical security plan, guard training and qualification plan, or cyber security plan for the... Power Plant Unit 1, Exemption From Certain Security Requirements 1.0 Background DTE Energy (DTE) is the... atmospheric pressure. In November 1972, the Power Reactor Development Company (PRDC), the licensee at that...

Method for Biochar Passivation Using Low Percent Oxygen

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

Smith, Kristin; Dupuis, Dan; Wilcox, Esther

2016-06-06

The thermochemical process development unit may be configured for pyrolysis or gasification. The pyrolysis unit operations include: feed transport system; entrained flow reactor; solids removal and collection; and liquid scrubbing, collection, and filtration. Char accumulates in the collection drums at a rate of ~1.5 kg/hr and must be passivated before it is stored or transported.

INL Director Discusses the Future for Nuclear Energy in the United States

ScienceCinema

Grossenbacher, John

2018-01-15

Idaho National Laboratory's Director John Grossenbacher explains that the United States should develop its energy policies based on an assessment of the current events at Japan's Fukushima nuclear reactors and the costs and benefits of providing electricity through various energy sources. For more information about INL's nuclear energy research, visit http://www.facebook.com/idahonationallaboratory.

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

NASA Astrophysics Data System (ADS)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control methods, fuel form and matrix, determination of shielding requirements, as well as power conversion and heat rejection techniques to generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces such as Mars, other moons, and asteroids.

Developing a concept for a national used fuel interim storage facility in the United States

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

Lewis, Donald Wayne

2013-07-01

In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a 'Monitored Retrievable Storage' facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to buildmore » a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant. The purpose of this paper is to present a concept developed to address the DOE's goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities. The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility. (authors)« less

Extruder system for high-throughput/steady-state hydrogen ice supply and application for pellet fueling of reactor-scale fusion experiments

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

Combs, S.K.; Foust, C.R.; Qualls, A.L.

Pellet injection systems for the next-generation fusion devices, such as the proposed International Thermonuclear Experimental Reactor (ITER), will require feed systems capable of providing a continuous supply of hydrogen ice at high throughputs. A straightforward concept in which multiple extruder units operate in tandem has been under development at the Oak Ridge National Laboratory. A prototype with three large-volume extruder units has been fabricated and tested in the laboratory. In experiments, it was found that each extruder could provide volumetric ice flow rates of up to {approximately}1.3 cm{sup 3}/s (for {approximately}10 s), which is sufficient for fueling fusion reactors atmore » the gigawatt power level. With the three extruders of the prototype operating in sequence, a steady rate of {approximately}0.33 cm{sup 3}/s was maintained for a duration of 1 h. Even steady-state rates approaching the full ITER design value ({approximately}1 cm{sup 3}/s) may be feasible with the prototype. However, additional extruder units (1{endash}3) would facilitate operations at the higher throughputs and reduce the duty cycle of each unit. The prototype can easily accommodate steady-state pellet fueling of present large tokamaks or other near-term plasma experiments.« less

NETL - Chemical Looping Reactor

ScienceCinema

None

2018-02-14

NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

Uranium, its impact on the national and global energy mix; and its history, distribution, production, nuclear fuel-cycle, future, and relation to the environment

USGS Publications Warehouse

Finch, Warren Irvin

1997-01-01

The many aspects of uranium, a heavy radioactive metal used to generate electricity throughout the world, are briefly described in relatively simple terms intended for the lay reader. An adequate glossary of unfamiliar terms is given. Uranium is a new source of electrical energy developed since 1950, and how we harness energy from it is explained. It competes with the organic coal, oil, and gas fuels as shown graphically. Uranium resources and production for the world are tabulated and discussed by country and for various energy regions in the United States. Locations of major uranium deposits and power reactors in the United States are mapped. The nuclear fuel-cycle of uranium for a typical light-water reactor is illustrated at the front end-beginning with its natural geologic occurrence in rocks through discovery, mining, and milling; separation of the scarce isotope U-235, its enrichment, and manufacture into fuel rods for power reactors to generate electricity-and at the back end-the reprocessing and handling of the spent fuel. Environmental concerns with the entire fuel cycle are addressed. The future of the use of uranium in new, simplified, 'passively safe' reactors for the utility industry is examined. The present resource assessment of uranium in the United States is out of date, and a new assessment could aid the domestic uranium industry.

First flush reactor for stormwater treatment for elevated linear transportation projects.

DOT National Transportation Integrated Search

2009-06-01

The United States EPA (Environmental Protection Agency) MS4 (Municipal Separate Storm Water Sewer System) Program regulations : require municipalities and government agencies including the Louisiana Department of Transportation and Development (LADOT...

Lunar Resource Utilization: Development of a Reactor for Volatile Extraction from Regolith

NASA Technical Reports Server (NTRS)

Kleinhenz, Julie E.; Sacksteder, Kurt R.; Nayagam, Vedha

2007-01-01

The extraction and processing of planetary resources into useful products, known as In- Situ Resource Utilization (ISRU), will have a profound impact on the future of planetary exploration. One such effort is the RESOLVE (Regolith and Environment Science, Oxygen and Lunar Volatiles Extraction) Project, which aims to extract and quantify these resources. As part of the first Engineering Breadboard Unit, the Regolith Volatiles Characterization (RVC) reactor was designed and built at the NASA Glenn Research Center. By heating and agitating the lunar regolith, loosely bound volatiles, such as hydrogen and water, are released and stored in the reactor for later analysis and collection. Intended for operation on a robotic rover, the reactor features a lightweight, compact design, easy loading and unloading of the regolith, and uniform heating of the regolith by means of vibrofluidization. The reactor performance was demonstrated using regolith simulant, JSC1, with favorable results.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

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

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. Themore » objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.« less

77 FR 41814 - Entergy Operations, Inc.; Grand Gulf Nuclear Station, Unit 1

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-16

... Unit 1 result primarily from periodic testing of diesel generators and fire water pump diesel engines... rural. GGNS Unit 1 is a General Electric Mark 3 boiling-water reactor. Identification of the Proposed... following: replacing the reactor feed pump turbine rotors; replacing the main generator current transformers...

An Analysis of Warfighter Sleep, Fatigue, and Performance on the USS Nimitz

DTIC Science & Technology

2014-09-01

35 1. Chernobyl Reactor 4 .............................................................. 36 2...deprivation and fatigue can be disastrous, as demonstrated by the accidents at Chernobyl Reactor 4, Three Mile Island Unit 2, Bhopal Union Carbide, and the...deprivation and fatigue can be disastrous, as demonstrated by the accidents at Chernobyl Reactor 4, Three Mile Island Unit 2, Bhopal Union Carbide, and

Design of virtual SCADA simulation system for pressurized water reactor

NASA Astrophysics Data System (ADS)

Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

2016-02-01

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

Radioactive Waste Management and Environmental Contamination Issues at the Chernobyl Site

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

Napier, Bruce A.; Schmieman, Eric A.; Voitsekhovitch, Oleg V.

2007-11-01

The destruction of the Unit 4 reactor at the Chernobyl Nuclear Power Plant resulted in the generation of radioactive contamination and radioactive waste at the site and in the surrounding area (referred to as the Exclusion Zone). In the course of remediation activities, large volumes of radioactive waste were generated and placed in temporary near surface waste-storage and disposal facilities. Trench and landfill type facilities were created from 1986 to 1987 in the Chernobyl Exclusion Zone at distances 0.5 to 15 km from the NPP site. This large number of facilities was established without proper design documentation, engineered barriers, ormore » hydrogeological investigations and they do not meet contemporary waste-safety requirements. Immediately following the accident, a Shelter was constructed over the destroyed reactor; in addition to uncertainties in stability at the time of its construction, structural elements of the Shelter have degraded as a result of corrosion. The main potential hazard of the Shelter is a possible collapse of its top structures and release of radioactive dust into the environment. A New Safe Confinement (NSC) with a 100-years service life is planned to be built as a cover over the existing Shelter as a longer-term solution. The construction of the NSC will enable the dismantlement of the current Shelter, removal of highly radioactive, fuel-containing materials from Unit 4, and eventual decommissioning of the damaged reactor. More radioactive waste will be generated during NSC construction, possible Shelter dismantling, removal of fuel containing materials, and decommissioning of Unit 4. The future development of the Exclusion Zone depends on the future strategy for converting Unit 4 into an ecologically safe system, i.e., the development of the NSC, the dismantlement of the current Shelter, removal of fuel containing material, and eventual decommissioning of the accident site. To date, a broadly accepted strategy for radioactive waste management at the reactor site and in the Exclusion Zone, and especially for high-level and long-lived waste, has not been developed.« 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.*

Preliminary Tritium Management Design Activities at ORNL

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

Harrison, Thomas J.; Felde, David K.; Logsdon, Randall J.

2016-09-01

Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritiummore » mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.« less

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.

Characterization of fast neutron spectrum in the TRIGA for hardness testing of electronic components

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

Nelson, George W.

1986-07-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering Laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems. (author)« less

A modular reactor to simulate biofilm development in orthopedic materials.

PubMed

Barros, Joana; Grenho, Liliana; Manuel, Cândida M; Ferreira, Carla; Melo, Luís F; Nunes, Olga C; Monteiro, Fernando J; Ferraz, Maria P

2013-09-01

Surfaces of medical implants are generally designed to encourage soft- and/or hard-tissue adherence, eventually leading to tissue- or osseo-integration. Unfortunately, this feature may also encourage bacterial adhesion and biofilm formation. To understand the mechanisms of bone tissue infection associated with contaminated biomaterials, a detailed understanding of bacterial adhesion and subsequent biofilm formation on biomaterial surfaces is needed. In this study, a continuous-flow modular reactor composed of several modular units placed in parallel was designed to evaluate the activity of circulating bacterial suspensions and thus their predilection for biofilm formation during 72 h of incubation. Hydroxyapatite discs were placed in each modular unit and then removed at fixed times to quantify biofilm accumulation. Biofilm formation on each replicate of material, unchanged in structure, morphology, or cell density, was reproducibly observed. The modular reactor therefore proved to be a useful tool for following mature biofilm formation on different surfaces and under conditions similar to those prevailing near human-bone implants.

Conceptual design and development of GEM based detecting system for tomographic tungsten focused transport monitoring

NASA Astrophysics Data System (ADS)

Chernyshova, M.; Czarski, T.; Malinowski, K.; Kowalska-Strzęciwilk, E.; Poźniak, K.; Kasprowicz, G.; Zabołotny, W.; Wojeński, A.; Kolasiński, P.; Mazon, D.; Malard, P.

2015-10-01

Implementing tungsten as a plasma facing material in ITER and future fusion reactors will require effective monitoring of not just its level in the plasma but also its distribution. That can be successfully achieved using detectors based on Gas Electron Multiplier (GEM) technology. This work presents the conceptual design of the detecting unit for poloidal tomography to be tested at the WEST project tokamak. The current stage of the development is discussed covering aspects which include detector's spatial dimensions, gas mixtures, window materials and arrangements inside and outside the tokamak ports, details of detector's structure itself and details of the detecting module electronics. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing the creation of sustainable nuclear fusion reactors a step closer. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

U.S.-Russian Cooperation in Science and Technology: A Case Study of the TOPAZ Space-Based Nuclear Reactor International Program

NASA Astrophysics Data System (ADS)

Dabrowski, Richard S.

2014-08-01

The TOPAZ International Program (TIP) was the final name given to a series of projects to purchase and test the TOPAZ-II, a space-based nuclear reactor of a type that had been further developed in the Soviet Union than in the United States. In the changing political situation associated with the break-up of the Soviet Union it became possible for the United States to not just purchase the system, but also to employ Russian scientists, engineers and testing facilities to verify its reliability. The lessons learned from the TIP illuminate some of the institutional and cultural challenges to U.S. - Russian cooperation in technology research which remain true today.

A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

DOE PAGES

Billings, Jay Jay; Deyton, Jordan H.; Forest Hull, S.; ...

2015-07-17

Building new fission reactors in the United States presents many technical and regulatory challenges. Chief among the technical challenges is the need to share and present results from new high- fidelity, high- performance simulations in an easily consumable way. In light of the modern multi-scale, multi-physics simulations can generate petabytes of data, this will require the development of new techniques and methods to reduce the data to familiar quantities of interest with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately availablemore » in the community and need to be developed. Our paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It enables easy qualitative and quantitative comparisons between simulation results with a graphical user interface and cross-platform, multi-language input- output libraries for use by developers to work with the data. One example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented along with a detailed discussion of the system s requirements and design.« less

Emergency Preparedness technology support to the Health and Safety Executive (HSE), Nuclear Installations Inspectorate (NII) of the United Kingdom. Appendix A

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

O`Kula, K.R.

1994-03-01

The Nuclear Installations Inspectorate (NII) of the United Kingdom (UK) suggested the use of an accident progression logic model method developed by Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) for K Reactor to predict the magnitude and timing of radioactivity releases (the source term) based on an advanced logic model methodology. Predicted releases are output from the personal computer-based model in a level-of-confidence format. Additional technical discussions eventually led to a request from the NII to develop a proposal for assembling a similar technology to predict source terms for the UK`s advanced gas-cooled reactor (AGR) type.more » To respond to this request, WSRC is submitting a proposal to provide contractual assistance as specified in the Scope of Work. The work will produce, document, and transfer technology associated with a Decision-Oriented Source Term Estimator for Emergency Preparedness (DOSE-EP) for the NII to apply to AGRs in the United Kingdom. This document, Appendix A is a part of this proposal.« less

The influence of perforation of foil reactors on greenhouse gas emission rates during aerobic biostabilization of the undersize fraction of municipal wastes.

PubMed

Stegenta, Sylwia; Dębowski, Marcin; Bukowski, Przemysław; Randerson, Peter F; Białowiec, Andrzej

2018-02-01

The opinion, that the use of foil reactors for the aerobic biostabilization of municipal wastes is not a valid method, due to vulnerability to perforation, and risk of uncontrolled release of exhaust gasses, was verified. This study aimed to determine the intensity of greenhouse gas (GHG) emissions to the atmosphere from the surface of foil reactors in relation to the extent of foil surface perforation. Three scenarios were tested: intact (airtight) foil reactor, perforated foil reactor, and torn foil reactor. Each experimental variant was triplicated, and the duration of each experiment cycle was 5 weeks. Temperature measurements demonstrated a significant decrease in temperature of the biostabilization in the torn reactor. The highest emissions of CO 2 , CO and SO 2 were observed at the beginning of the process, and mostly in the torn reactor. During the whole experiment, observed emissions of CO, H 2 S, NO, NO 2 , and SO 2 were at a very low level which in extreme cases did not exceed 0.25 mg t -1 .h -1 (emission of gasses mass unit per waste mass unit per unit time). The lowest average emissions of greenhouse gases were determined in the case of the intact reactor, which shows that maintaining the foil reactors in an airtight condition during the process is extremely important. Copyright © 2017 Elsevier Ltd. All rights reserved.

The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine.

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

Kot, C.

1999-05-10

Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Departmentmore » of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments.« less

Design of virtual SCADA simulation system for pressurized water reactor

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

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles ofmore » energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.« less

Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

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

OHara J. M.; Higgins, J.; DAgostino, A.

2012-01-17

The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a singlemore » operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.« less

76 FR 72007 - ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-21

...; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security Requirements 1.0 Background Zion Nuclear Power Station (ZNPS or Zion), Unit 1, is a Westinghouse 3250 MWt Pressurized Water Reactor... activities in nuclear power reactors against radiological sabotage,'' paragraph (b)(1) states, ``The licensee...

From the first nuclear power plant to fourth-generation nuclear power installations [on the 60th anniversary of the World's First nuclear power plant

NASA Astrophysics Data System (ADS)

Rachkov, V. I.; Kalyakin, S. G.; Kukharchuk, O. F.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

Successful commissioning in the 1954 of the World's First nuclear power plant constructed at the Institute for Physics and Power Engineering (IPPE) in Obninsk signaled a turn from military programs to peaceful utilization of atomic energy. Up to the decommissioning of this plant, the AM reactor served as one of the main reactor bases on which neutron-physical investigations and investigations in solid state physics were carried out, fuel rods and electricity generating channels were tested, and isotope products were bred. The plant served as a center for training Soviet and foreign specialists on nuclear power plants, the personnel of the Lenin nuclear-powered icebreaker, and others. The IPPE development history is linked with the names of I.V. Kurchatov, A.I. Leipunskii, D.I. Blokhintsev, A.P. Aleksandrov, and E.P. Slavskii. More than 120 projects of various nuclear power installations were developed under the scientific leadership of the IPPE for submarine, terrestrial, and space applications, including two water-cooled power units at the Beloyarsk NPP in Ural, the Bilibino nuclear cogeneration station in Chukotka, crawler-mounted transportable TES-3 power station, the BN-350 reactor in Kazakhstan, and the BN-600 power unit at the Beloyarsk NPP. Owing to efforts taken on implementing the program for developing fast-neutron reactors, Russia occupied leading positions around the world in this field. All this time, IPPE specialists worked on elaborating the principles of energy supertechnologies of the 21st century. New large experimental installations have been put in operation, including the nuclear-laser setup B, the EGP-15 accelerator, the large physical setup BFS, the high-pressure setup SVD-2; scientific, engineering, and technological schools have been established in the field of high- and intermediate-energy nuclear physics, electrostatic accelerators of multicharge ions, plasma processes in thermionic converters and nuclear-pumped lasers, physics of compact nuclear reactors and radiation protection, thermal physics, physical chemistry and technology of liquid metal coolants, and physics of radiation-induced defects, and radiation materials science. The activity of the institute is aimed at solving matters concerned with technological development of large-scale nuclear power engineering on the basis of a closed nuclear fuel cycle with the use of fast-neutron reactors (referred to henceforth as fast reactors), development of innovative nuclear and conventional technologies, and extension of their application fields.

Thermal characteristics analysis of microwaves reactor for pyrolysis of used cooking oil

NASA Astrophysics Data System (ADS)

Anis, Samsudin; Shahadati, Laily; Sumbodo, Wirawan; Wahyudi

2017-03-01

The research is objected to develop microwave reactor for pyrolysis of used cooking oil. The effect of microwave power as well as addition of char as absorber towards its thermal characteristic were investigated. Domestic microwave was modified and used to test the thermal characteristic of used cooking oil in the terms of temperature evolution, heating rate, and thermal efficiency. The samples were examined under various microwave power of 347W, 399W, 572W and 642W for 25 minutes of irradiation time. The char loading was tested in the level of 0, 50, and 100 g. Microwave reactor consists of microwave unit with a maximum power of 642W, a ceramic reactor, and a condenser equipped with temperature measurement system was successfully developed. It was found that microwave power and addition of absorber significantly influenced the thermal characteristic of microwave reactor. Under investigated condition, the optimum result was obtained at microwave power of 642W and 100 g of char. The condition was able to provide temperature of 480°C, heating rate of 18.2°C/min and thermal efficiency of 53% that is suitable to pyrolyze used cooking oil.

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

Biological Cr(VI) removal using bio-filters and constructed wetlands.

PubMed

Michailides, Michail K; Sultana, Mar-Yam; Tekerlekopoulou, Athanasia G; Akratos, Christos S; Vayenas, Dimitrios V

2013-01-01

The bioreduction of hexavalent chromium from aqueous solution was carried out using suspended growth and packed-bed reactors under a draw-fill operating mode, and horizontal subsurface constructed wetlands. Reactors were inoculated with industrial sludge from the Hellenic Aerospace Industry using sugar as substrate. In the suspended growth reactors, the maximum Cr(VI) reduction rate (about 2 mg/L h) was achieved for an initial concentration of 12.85 mg/L, while in the attached growth reactors, a similar reduction rate was achieved even with high initial concentrations (109 mg/L), thus confirming the advantage of these systems. Two horizontal subsurface constructed wetlands (CWs) pilot-scale units were also built and operated. The units contained fine gravel. One unit was planted with common reeds and one was kept unplanted. The mean influent concentrations of Cr(VI) were 5.61 and 5.47 mg/L for the planted and unplanted units, respectively. The performance of the planted CW units was very effective as mean Cr(VI) removal efficiency was 85% and efficiency maximum reached 100%. On the contrary, the unplanted CW achieved very low Cr(VI) removal with a mean value of 26%. Both attached growth reactors and CWs proved efficient and viable means for Cr(VI) reduction.

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

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

NASA Technical Reports Server (NTRS)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Developing and applying modern methods of leakage monitoring and state estimation of fuel at the Novovoronezh nuclear power plant

NASA Astrophysics Data System (ADS)

Povarov, V. P.; Tereshchenko, A. B.; Kravchenko, Yu. N.; Pozychanyuk, I. V.; Gorobtsov, L. I.; Golubev, E. I.; Bykov, V. I.; Likhanskii, V. V.; Evdokimov, I. A.; Zborovskii, V. G.; Sorokin, A. A.; Kanyukova, V. D.; Aliev, T. N.

2014-02-01

The results of developing and implementing the modernized fuel leakage monitoring methods at the shut-down and running reactor of the Novovoronezh nuclear power plant (NPP) are presented. An automated computerized expert system integrated with an in-core monitoring system (ICMS) and installed at the Novovoronezh NPP unit no. 5 is described. If leaky fuel elements appear in the core, the system allows one to perform on-line assessment of the parameters of leaky fuel assemblies (FAs). The computer expert system units designed for optimizing the operating regimes and enhancing the fuel usage efficiency at the Novovoronezh NPP unit no. 5 are now being developed.

Methyl chloride via oxyhydrochlorination of methane: A building block for chemicals and fuels from natural gas

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

Benson, R.L.; Brown, S.S.D.; Ferguson, S.P.

1995-12-31

The objectives of this program are to (a) develop a process for converting natural gas to methyl chloride via an oxyhydrochlorination route using highly selective, stable catalysts in a fixed-bed, (b) design a reactor capable of removing the large amount of heat generated in the process so as to control the reaction, (c) develop a recovery system capable of removing the methyl chloride from the product stream and (d) determine the economics and commercial viability of the process. The general approach has been as follows: (a) design and build a laboratory scale reactor, (b) define and synthesize suitable OHC catalystsmore » for evaluation, (c) select first generation OHC catalyst for Process Development Unit (PDU) trials, (d) design, construct and startup PDU, (e) evaluate packed bed reactor design, (f) optimize process, in particular, product recovery operations, (g) determine economics of process, (h) complete preliminary engineering design for Phase II and (i) make scale-up decision and formulate business plan for Phase II. Conclusions regarding process development and catalyst development are presented.« 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.

Semiconductor Chemical Reactor Engineering and Photovoltaic Unit Operations.

ERIC Educational Resources Information Center

Russell, T. W. F.

1985-01-01

Discusses the nature of semiconductor chemical reactor engineering, illustrating the application of this engineering with research in physical vapor deposition of cadmium sulfide at both the laboratory and unit operations scale and chemical vapor deposition of amorphous silicon at the laboratory scale. (JN)

Understanding Victims of Technological Disaster: Beliefs and Worries of Three Mile Island.

ERIC Educational Resources Information Center

Prince-Embury, Sandra; Rooney, James

The primary purpose of the present study was to examine how prevalent were concerns about restarting Three Mile Island nuclear reactor Unit I among people within a five-mile radius of the plant four years after the accident involving reactor Unit II. Also explored were concerns related to expectations about the restart of Unit I, perception of…

The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

NASA Astrophysics Data System (ADS)

Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

2015-05-01

The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

NASA Astrophysics Data System (ADS)

Shimomura, Y.

1994-05-01

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

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

Fusion energy

NASA Astrophysics Data System (ADS)

1990-09-01

The main purpose of the International Thermonuclear Experimental Reactor (ITER) is to develop an experimental fusion reactor through the united efforts of many technologically advanced countries. The ITER terms of reference, issued jointly by the European Community, Japan, the USSR, and the United States, call for an integrated international design activity and constitute the basis of current activities. Joint work on ITER is carried out under the auspices of the International Atomic Energy Agency (IAEA), according to the terms of quadripartite agreement reached between the European Community, Japan, the USSR, and the United States. The site for joint technical work sessions is at the Max Planck Institute of Plasma Physics. Garching, Federal Republic of Germany. The ITER activities have two phases: a definition phase performed in 1988 and the present design phase (1989 to 1990). During the definition phase, a set of ITER technical characteristics and supporting research and development (R and D) activities were developed and reported. The present conceptual design phase of ITER lasts until the end of 1990. The objectives of this phase are to develop the design of ITER, perform a safety and environmental analysis, develop site requirements, define future R and D needs, and estimate cost, manpower, and schedule for construction and operation. A final report will be submitted at the end of 1990. This paper summarizes progress in the ITER program during the 1989 design phase.

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.

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.

Implementation of ALARA at the design stage of Nuclear Power Plants

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

Brissaud, A.; Ridoux, P.

1995-03-01

In the 1970s, Electricite de France (EdF) had limited knowledge and experience of pressurized water reactors (PWRs). Electricity generation by nuclear units was oriented towards gas-graphite reactors, even though EdF had a share in the PWR unit of CHOOZ A-1 (250 MWe, later upgraded to 320 MWe). Some facts about the origin of doses in that king of reactor were known to the research and development (R&D) support staff of EdF, which mainly comprises the French Atomic Commission (CEA), but only a few of EdF`s engineers were aware of these facts. One has to bear in mind that CHOOZ A-1more » only went critical in April 1967 and was officially connected to the grid in May 1970 after some important problems had been solved. Meanwhile, the nuclear program was launched at full speed, beginning with the order for FESSENHEIM 1 in 1970, FESSENHEIM 2 and BUGEY 2 and 3 in 1971. TIHANGE 1, in which EdF had a share, went on-line in September 1975. Also, supposing that EdF had had such knowledge and experience, it is quite evident that it would have been very difficult to modify the lay-out inside the reactor building.« less

Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

NASA Astrophysics Data System (ADS)

Shchelik, S. V.; Pavlov, A. S.

2013-07-01

Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

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

2007-09-01

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

Mitigating Community Impacts of Energy Development: Some Examples for Coal and Nuclear Generating Plants in the United States.

ERIC Educational Resources Information Center

Peelle, Elizabeth

The Hartsville, Tennessee nuclear reactor site, the coal plant at Wheatland, Wyoming, and the nuclear plant at Skagit, Washington have mitigation plans developed in response to a federal, state, and local regulatory agency, respectively; the three mitigation plans aim at internalizing community-level social costs and benefits during the…

Development of process control capability through the Browns Ferry Integrated Computer System using Reactor Water Clanup System as an example. Final report

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

Smith, J.; Mowrey, J.

1995-12-01

This report describes the design, development and testing of process controls for selected system operations in the Browns Ferry Nuclear Plant (BFNP) Reactor Water Cleanup System (RWCU) using a Computer Simulation Platform which simulates the RWCU System and the BFNP Integrated Computer System (ICS). This system was designed to demonstrate the feasibility of the soft control (video touch screen) of nuclear plant systems through an operator console. The BFNP Integrated Computer System, which has recently. been installed at BFNP Unit 2, was simulated to allow for operator control functions of the modeled RWCU system. The BFNP Unit 2 RWCU systemmore » was simulated using the RELAP5 Thermal/Hydraulic Simulation Model, which provided the steady-state and transient RWCU process variables and simulated the response of the system to control system inputs. Descriptions of the hardware and software developed are also included in this report. The testing and acceptance program and results are also detailed in this report. A discussion of potential installation of an actual RWCU process control system in BFNP Unit 2 is included. Finally, this report contains a section on industry issues associated with installation of process control systems in nuclear power plants.« less

The chemical energy unit partial oxidation reactor operation simulation modeling

NASA Astrophysics Data System (ADS)

Mrakin, A. N.; Selivanov, A. A.; Batrakov, P. A.; Sotnikov, D. G.

2018-01-01

The chemical energy unit scheme for synthesis gas, electric and heat energy production which is possible to be used both for the chemical industry on-site facilities and under field conditions is represented in the paper. The partial oxidation reactor gasification process mathematical model is described and reaction products composition and temperature determining algorithm flow diagram is shown. The developed software product verification showed good convergence of the experimental values and calculations according to the other programmes: the temperature determining relative discrepancy amounted from 4 to 5 %, while the absolute composition discrepancy ranged from 1 to 3%. The synthesis gas composition was found out practically not to depend on the supplied into the partial oxidation reactor (POR) water vapour enthalpy and compressor air pressure increase ratio. Moreover, air consumption coefficient α increase from 0.7 to 0.9 was found out to decrease synthesis gas target components (carbon and hydrogen oxides) specific yield by nearly 2 times and synthesis gas target components required ratio was revealed to be seen in the water vapour specific consumption area (from 5 to 6 kg/kg of fuel).

A dense cell retention culture system using stirred ceramic membrane reactor.

PubMed

Suzuki, T; Sato, T; Kominami, M

1994-11-20

A novel reactor design incorporating porous ceramic tubes into a stirred jar fermentor was developed. The stirred ceramic membrane reactor has two ceramic tubular membrane units inside the vessel and maintains high filtration flux by alternating use for filtering and recovering from clogging. Each filter unit was linked for both extraction of culture broth and gas sparging. High permeability was maintained for long periods by applying the periodical control between filtering and air sparging during the stirred retention culture of Saccharomyces cerevisiae. The ceramic filter aeration system increased the k(L)a to about five times that of ordinary gas sparing. Using the automatic feeding and filtering system, cell mass concentration reached 207 g/L in a short time, while it was 64 g/L in a fed-batch culture. More than 99% of the growing cells were retained in the fermentor by the filtering culture. Both yield and productivity of cells were also increased by controlling the feeding of fresh medium and filtering the supernatant of the dense cells culture. (c) 1994 John Wiley & Sons, Inc.

Low cost solar aray project: Experimental process system development unit for producing semiconductor-grade silicon using the silane-to-silicon process

NASA Technical Reports Server (NTRS)

1981-01-01

This phase consists of the engineering design, fabrication, assembly, operation, economic analysis, and process support R&D for an Experimental Process System Development Unit (EPSDU). The mechanical bid package was issued and the bid responses are under evaluation. Similarly, the electrical bid package was issued, however, responses are not yet due. The majority of all equipment is on order or has been received at the EPSDU site. The pyrolysis/consolidation process design package was issued. Preparation of process and instrumentation diagram for the free-space reactor was started. In the area of melting/consolidation, Kayex successfully melted chunk silicon and have produced silicon shot. The free-space reactor powder was successfully transported pneumatically from a storage bin to the auger feeder twenty-five feet up and was melted. The fluid-bed PDU has successfully operated at silane feed concentrations up to 21%. The writing of the operating manual has started. Overall, the design phase is nearing completion.

Establishing a laboratory model of dental unit waterlines bacterial biofilms using a CDC biofilm reactor.

PubMed

Yoon, Hye Young; Lee, Si Young

2017-11-01

In this study, a laboratory model to reproduce dental unit waterline (DUWL) biofilms was developed using a CDC biofilm reactor (CBR). Bacteria obtained from DUWLs were filtered and cultured in Reasoner's 2A (R2A) for 10 days, and were subsequently stored at -70°C. This stock was cultivated on R2A in batch mode. After culturing for five days, the bacteria were inoculated into the CBR. Biofilms were grown on polyurethane tubing for four days. Biofilm accumulation and thickness was 1.3 × 10 5  CFU cm -2 and 10-14 μm respectively, after four days. Bacteria in the biofilms included cocci and rods of short and medium lengths. In addition, 38 bacterial genera were detected in biofilms. In this study, the suitability and reproducibility of the CBR model for DUWL biofilm formation were demonstrated. The model provides a foundation for the development of bacterial control methods for DUWLs.

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-11-03

separation list: ! 8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction...facilities like reprocessing and enrichment plants and breeder reactors could be viewed as providing a significant nonproliferation benefit because the... breeder reactors would support the 2002 U.S. National Strategy to Combat Weapons of Mass Destruction, in which the United States pledged to “continue to

ETR ELECTRICAL BUILDING, TRA648. EMERGENCY STANDBY GENERATOR AND DIESEL UNIT. ...

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

ETR ELECTRICAL BUILDING, TRA-648. EMERGENCY STANDBY GENERATOR AND DIESEL UNIT. METAL ROOF AND PUMICE BLOCK WALLS. CAMERA FACING SOUTHWEST. INL NEGATIVE NO. 56-3708. R.G. Larsen, Photographer, 11/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Low-Dissolved-Oxygen Nitrifying Systems Exploit Ammonia-Oxidizing Bacteria with Unusually High Yields▿

PubMed Central

Bellucci, Micol; Ofiţeru, Irina D.; Graham, David W.; Head, Ian M.; Curtis, Thomas P.

2011-01-01

In wastewater treatment plants, nitrifying systems are usually operated with elevated levels of aeration to avoid nitrification failures. This approach contributes significantly to operational costs and the carbon footprint of nitrifying wastewater treatment processes. In this study, we tested the effect of aeration rate on nitrification by correlating ammonia oxidation rates with the structure of the ammonia-oxidizing bacterial (AOB) community and AOB abundance in four parallel continuous-flow reactors operated for 43 days. Two of the reactors were supplied with a constant airflow rate of 0.1 liter/min, while in the other two units the airflow rate was fixed at 4 liters/min. Complete nitrification was achieved in all configurations, though the dissolved oxygen (DO) concentration was only 0.5 ± 0.3 mg/liter in the low-aeration units. The data suggest that efficient performance in the low-DO units resulted from elevated AOB levels in the reactors and/or putative development of a mixotrophic AOB community. Denaturing gel electrophoresis and cloning of AOB 16S rRNA gene fragments followed by sequencing revealed that the AOB community in the low-DO systems was a subset of the community in the high-DO systems. However, in both configurations the dominant species belonged to the Nitrosomonas oligotropha lineage. Overall, the results demonstrated that complete nitrification can be achieved at low aeration in lab-scale reactors. If these findings could be extended to full-scale plants, it would be possible to minimize the operational costs and greenhouse gas emissions without risk of nitrification failure. PMID:21926211

Advanced I&C for Fault-Tolerant Supervisory Control of Small Modular Reactors

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

Cole, Daniel G.

In this research, we have developed a supervisory control approach to enable automated control of SMRs. By design the supervisory control system has an hierarchical, interconnected, adaptive control architecture. A considerable advantage to this architecture is that it allows subsystems to communicate at different/finer granularity, facilitates monitoring of process at the modular and plant levels, and enables supervisory control. We have investigated the deployment of automation, monitoring, and data collection technologies to enable operation of multiple SMRs. Each unit's controller collects and transfers information from local loops and optimize that unit’s parameters. Information is passed from the each SMR unitmore » controller to the supervisory controller, which supervises the actions of SMR units and manage plant processes. The information processed at the supervisory level will provide operators the necessary information needed for reactor, unit, and plant operation. In conjunction with the supervisory effort, we have investigated techniques for fault-tolerant networks, over which information is transmitted between local loops and the supervisory controller to maintain a safe level of operational normalcy in the presence of anomalies. The fault-tolerance of the supervisory control architecture, the network that supports it, and the impact of fault-tolerance on multi-unit SMR plant control has been a second focus of this research. To this end, we have investigated the deployment of advanced automation, monitoring, and data collection and communications technologies to enable operation of multiple SMRs. We have created a fault-tolerant multi-unit SMR supervisory controller that collects and transfers information from local loops, supervise their actions, and adaptively optimize the controller parameters. The goal of this research has been to develop the methodologies and procedures for fault-tolerant supervisory control of small modular reactors. To achieve this goal, we have identified the following objectives. These objective are an ordered approach to the research: I) Development of a supervisory digital I&C system II) Fault-tolerance of the supervisory control architecture III) Automated decision making and online monitoring.« less

Thermodynamic analysis of in situ gasification-chemical looping combustion (iG-CLC) of Indian coal.

PubMed

Suresh, P V; Menon, Kavitha G; Prakash, K S; Prudhvi, S; Anudeep, A

2016-10-01

Chemical looping combustion (CLC) is an inherent CO 2 capture technology. It is gaining much interest in recent years mainly because of its potential in addressing climate change problems associated with CO 2 emissions from power plants. A typical chemical looping combustion unit consists of two reactors-fuel reactor, where oxidation of fuel occurs with the help of oxygen available in the form of metal oxides and, air reactor, where the reduced metal oxides are regenerated by the inflow of air. These oxides are then sent back to the fuel reactor and the cycle continues. The product gas from the fuel reactor contains a concentrated stream of CO 2 which can be readily stored in various forms or used for any other applications. This unique feature of inherent CO 2 capture makes the technology more promising to combat the global climate changes. Various types of CLC units have been discussed in literature depending on the type of fuel burnt. For solid fuel combustion three main varieties of CLC units exist namely: syngas CLC, in situ gasification-CLC (iG-CLC) and chemical looping with oxygen uncoupling (CLOU). In this paper, theoretical studies on the iG-CLC unit burning Indian coal are presented. Gibbs free energy minimization technique is employed to determine the composition of flue gas and oxygen carrier of an iG-CLC unit using Fe 2 O 3 , CuO, and mixed carrier-Fe 2 O 3 and CuO as oxygen carriers. The effect of temperature, suitability of oxygen carriers, and oxygen carrier circulation rate on the performance of a CLC unit for Indian coal are studied and presented. These results are analyzed in order to foresee the operating conditions at which economic and smooth operation of the unit is expected.

Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor

DOEpatents

Pennell, William E.

1977-01-01

A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annular region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve an hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings. The primary inlet openings have characteristics which limit the entry of debris and minimize the potential for debris entering the primary inlets blocking the secondary inlets from inside the modular unit.

United States and Russian Cooperation on Issues of Nuclear Nonproliferation

DTIC Science & Technology

2005-06-01

Reactors ( RERTR ) This project works with Russia to facilitate conversion of its research and test reactors from highly enriched uranium (HEU) fuel...reactor fuel purchase, accelerated RERTR activities, and accelerated Material Conversion and Consolidation implementation. 89 j. Fissile Materials

Fission Surface Power for the Exploration and Colonization of Mars

NASA Technical Reports Server (NTRS)

Houts, Mike; Porter, Ron; Gaddis, Steve; Van Dyke, Melissa; Martin, Jim; Godfroy, Tom; Bragg-Sitton, Shannon; Garber, Anne; Pearson, Boise

2006-01-01

The colonization of Mars will require abundant energy. One potential energy source is nuclear fission. Terrestrial fission systems are highly developed and have the demonstrated ability to safely produce tremendous amounts of energy. In space, fission systems not only have the potential to safely generate tremendous amounts of energy, but could also potentially be used on missions where alternatives are not practical. Programmatic risks such as cost and schedule are potential concerns with fission surface power (FSP) systems. To be mission enabling, FSP systems must be affordable and programmatic risk must be kept acceptably low to avoid jeopardizing exploration efforts that may rely on FSP. Initial FSP systems on Mars could be "workhorse" units sized to enable the establishment of a Mars base and the early growth of a colony. These systems could be nearly identical to FSP systems used on the moon. The systems could be designed to be safe, reliable, and have low development and recurring costs. Systems could also be designed to fit on relatively small landers. One potential option for an early Mars FSP system would be a 100 kWt class, NaK cooled system analogous to space reactors developed and flown under the U.S. "SNAP" program or those developed and flown by the former Soviet Union ("BUK" reactor). The systems could use highly developed fuel and materials. Water and Martian soil could be used to provide shielding. A modern, high-efficiency power conversion subsystem could be used to reduce required reactor thermal power. This, in turn, would reduce fuel burnup and radiation damage .effects by reducing "nuclear" fuels and materials development costs. A realistic, non-nuclear heated and fully integrated technology demonstration unit (TDU) could be used to reduce cost and programmatic uncertainties prior to initiating a flight program.

40 CFR 63.107 - Identification of process vents subject to this subpart.

Code of Federal Regulations, 2012 CFR

2012-07-01

... process vents associated with an air oxidation reactor, distillation unit, or reactor that is in a source.... (b) Some, or all, of the gas stream originates as a continuous flow from an air oxidation reactor... specified in paragraphs (c)(1) through (3) of this section. (1) Is directly from an air oxidation reactor...

76 FR 48184 - Exelon Nuclear, Peach Bottom Atomic Power Station, Unit 1; Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-08

... nuclear reactor facility. PBAPS Unit 1 was a high-temperature, gas-cooled reactor that was operated from... the safeguards contingency plan.'' Part 73 of 10 CFR, ``Physical Protection of Plant and Materials... physical protection system which will have capabilities for the protection of special nuclear material at...

78 FR 37591 - Entergy Nuclear Operations, Inc., Entergy Nuclear Indian Point Unit 2, LLC, Issuance of Director...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-21

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0284; Docket No. 50-247; License No. DPR-26] Entergy Nuclear Operations, Inc., Entergy Nuclear Indian Point Unit 2, LLC, Issuance of Director's Decision Notice is hereby given that the Deputy Director, Reactor Safety Programs, Office of Nuclear Reactor...

Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Dickens, Ricky; Houts, Michael; Pearson, Boise; Webster, Kenny; Gibson, Marc; Qualls, Lou; Poston, Dave; Werner, Jim; Radel, Ross

2011-01-01

The Nuclear Systems Team at NASA Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and Mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program, which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter for tests at MSFC. When tested at NASA Glenn Research Center (GRC) the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumentation (temperature, pressure, flow) for data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

NASA Astrophysics Data System (ADS)

Godfroy, T.; Dickens, R.; Houts, M.; Pearson, B.; Webster, K.; Gibson, M.; Qualls, L.; Poston, D.; Werner, J.; Radel, R.

The Nuclear Systems Team at Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter when being tested at MSFC. When tested at GRC the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumenta- tion (temperature, pressure, flow) data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

Integral Inherently Safe Light Water Reactor (I 2S-LWR)

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

Petrovic, Bojan; Memmott, Matthew; Boy, Guy

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I 2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to addressmore » the preference of some utilities in the US power market for unit power level on the order of 1 GWe.« less

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.

Production of oxalic acid from sugar beet molasses by formed nitrogen oxides.

PubMed

Gürü, M; Bilgesü, A Y; Pamuk, V

2001-03-01

Production of oxalic acid from sugar beet molasses was developed in a series of three reactors. Nitrogen oxides formed were used to manufacture oxalic acid in the second and third reactor. Parameters affecting the reaction were determined to be, air flow rate, temperature, the amount of V2O5 catalyst and the concentrations of molasses and H2SO4. The maximum yields in the second and third reactors were 78.9% and 74.6% of theoretical yield, respectively. Also, kinetic experiments were performed and the first-order rate constants were determined for the glucose consumption rate. Nitrogen oxides in off-gases from the final reactor were absorbed in water and concentrated sulphuric acid and reused in the following reactors giving slightly lower yields under similar conditions. In this novel way, it was possible to recover NO(x) and to prevent air pollution. Meanwhile, it was possible to reduce the unit cost of reactant for oxalic acid production. A maximum 77.5% and 74.1% of theoretical yield was obtained by using the absorption solutions with NO(x).

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

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

Swift, Alicia L.

There is no better time than now to close the loophole in Article IV of the Nuclear Non-proliferation Treaty (NPT) that excludes military uses of fissile material from nuclear safeguards. Several countries have declared their intention to pursue and develop naval reactor technology, including Argentina, Brazil, Iran, and Pakistan, while other countries such as China, India, Russia, and the United States are expanding their capabilities. With only a minority of countries using low enriched uranium (LEU) fuel in their naval reactors, it is possible that a state could produce highly enriched uranium (HEU) under the guise of a nuclear navymore » while actually stockpiling the material for a nuclear weapon program. This paper examines the likelihood that non-nuclear weapon states exploit the loophole to break out from the NPT and also the regional ramifications of deterrence and regional stability of expanding naval forces. Possible solutions to close the loophole are discussed, including expanding the scope of the Fissile Material Cut-off Treaty, employing LEU fuel instead of HEU fuel in naval reactors, amending the NPT, creating an export control regime for naval nuclear reactors, and forming individual naval reactor safeguards agreements.« less

Gas-Liquid Two-Phase Flows Through Packed Bed Reactors in Microgravity

NASA Technical Reports Server (NTRS)

Motil, Brian J.; Balakotaiah, Vemuri

2001-01-01

The simultaneous flow of gas and liquid through a fixed bed of particles occurs in many unit operations of interest to the designers of space-based as well as terrestrial equipment. Examples include separation columns, gas-liquid reactors, humidification, drying, extraction, and leaching. These operations are critical to a wide variety of industries such as petroleum, pharmaceutical, mining, biological, and chemical. NASA recognizes that similar operations will need to be performed in space and on planetary bodies such as Mars if we are to achieve our goals of human exploration and the development of space. The goal of this research is to understand how to apply our current understanding of two-phase fluid flow through fixed-bed reactors to zero- or partial-gravity environments. Previous experiments by NASA have shown that reactors designed to work on Earth do not necessarily function in a similar manner in space. Two experiments, the Water Processor Assembly and the Volatile Removal Assembly have encountered difficulties in predicting and controlling the distribution of the phases (a crucial element in the operation of this type of reactor) as well as the overall pressure drop.

Flowsheets and source terms for radioactive waste projections

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

Forsberg, C.W.

1985-03-01

Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF/sub 6/ conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables.

Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

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

Smith, Anthony A.

2013-07-01

The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] itmore » is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)« less

The U.S. Geological Survey's TRIGA® reactor

USGS Publications Warehouse

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

Batch-reactor microfluidic device: first human use of a microfluidically produced PET radiotracer†

PubMed Central

Miraghaie, Reza; Kotta, Kishore; Ball, Carroll E.; Zhang, Jianzhong; Buchsbaum, Monte S.; Kolb, Hartmuth C.; Elizarov, Arkadij

2013-01-01

The very first microfluidic device used for the production of 18F-labeled tracers for clinical research is reported along with the first human Positron Emission Tomography scan obtained with a microfluidically produced radiotracer. The system integrates all operations necessary for the transformation of [18F]fluoride in irradiated cyclotron target water to a dose of radiopharmaceutical suitable for use in clinical research. The key microfluidic technologies developed for the device are a fluoride concentration system and a microfluidic batch reactor assembly. Concentration of fluoride was achieved by means of absorption of the fluoride anion on a micro ion-exchange column (5 μL of resin) followed by release of the radioactivity with 45 μL of the release solution (95 ± 3% overall efficiency). The reactor assembly includes an injection-molded reactor chip and a transparent machined lid press-fitted together. The resulting 50 μL cavity has a unique shape designed to minimize losses of liquid during reactor filling and liquid evaporation. The cavity has 8 ports for gases and liquids, each equipped with a 2-way on-chip mechanical valve rated for pressure up to 20.68 bar (300 psi). The temperature is controlled by a thermoelectric heater capable of heating the reactor up to 180 °C from RT in 150 s. A camera captures live video of the processes in the reactor. HPLC-based purification and reformulation units are also integrated in the device. The system is based on “split-box architecture”, with reagents loaded from outside of the radiation shielding. It can be installed either in a standard hot cell, or as a self-shielded unit. Along with a high level of integration and automation, split-box architecture allowed for multiple production runs without the user being exposed to radiation fields. The system was used to support clinical trials of [18F]fallypride, a neuroimaging radiopharmaceutical under IND Application #109,880. PMID:23135409

Machining Test Specimens from Harvested Zion RPV Segments for Through Wall Attenuation Studies

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

Rosseel, Thomas M; Sokolov, Mikhail A; Nanstad, Randy K

2015-01-01

The decommissioning of the Zion Units 1 and 2 Nuclear Generating Station (NGS) in Zion, Illinois presents a special opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing Nuclear Power Plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, the selective procurement of materials, structures, and componentsmore » from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), the cutting of these segments into sections and blocks from the beltline and upper vertical welds and plate material, the current status of machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for chemical and microstructural (TEM, APT, SANS, and nano indention) characterization, as well as the current test plans and possible collaborative projects. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models (Rosseel et al. (2012) and Rosseel et al. (2015)).« less

40 CFR 60.701 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Synthetic Organic Chemical Manufacturing Industry (SOCMI) Reactor Processes § 60.701 Definitions. As used in... means any noncontinuous reactor process that is not characterized by steady-state conditions and in.... Reactor processes are unit operations in which one or more chemicals, or reactants other than air, are...

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

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

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

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

Kiff, Scott D.; Dazeley, Steven; Reyna, David

The current state-of-the-art in antineutrino detection is such that it is now possible to remotely monitor the operational status, power levels and fissile content of nuclear reactors in real-time. This non-invasive and incorruptible technique has been demonstrated at civilian power reactors in both Russia and the United States and has been of interest to the IAEA Novel Technologies Unit for several years. Expert's meetings were convened at IAEA headquarters in 2003 and again in 2008. The latter produced a report in which antineutrino detection was called a 'highly promising technology for safeguards applications' at nuclear reactors and several near-term goalsmore » and suggested developments were identified to facilitate wider applicability. Over the last few years, we have been working to achieve some of these goals and improvements. Specifically, we have already demonstrated the successful operation of non-toxic detectors and most recently, we are testing a transportable, above-ground detector system, which is fully contained within a standard 6 meter ISO container. If successful, such a system could allow easy deployment at any reactor facility around the world. As well, our previously demonstrated ability to remotely monitor the data and respond in real-time to reactor operational changes could allow the verification of operator declarations without the need for costly site-visits. As the global nuclear power industry expands around the world, the burden on maintaining operational histories and safeguarding inventories will increase greatly. Such a system for providing remote data to verify operator's declarations could greatly reduce the need for frequent site inspections while still providing a robust warning of anomalies requiring further investigation.« less

Commercial Superconducting Electron Linac for Radioisotope Production

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

Grimm, Terry Lee; Boulware, Charles H.; Hollister, Jerry L.

2015-08-13

The majority of radioisotopes used in the United States today come from foreign suppliers or are generated parasitically in large government accelerators and nuclear reactors. Both of these restrictions limit the availability of radioisotopes and discourage the development and evaluation of new isotopes and for nuclear medicine, science, and industry. Numerous studies have been recommending development of dedicated accelerators for production of radioisotopes for over 20 years (Institute of Medicine, 1995; Reba, et al, 2000; National Research Council, 2007; NSAC 2009). The 2015 NSAC Long Range Plan for Isotopes again identified electron accelerators as an area for continued research andmore » development. Recommendation 1(c) from the 2015 NSAC Isotope report specifically identifies electron accelerators for continued funding for the purpose of producing medical and industrial radioisotopes. Recognizing the pressing need for new production methods of radioisotopes, the United States Congress passed the American Medical Isotope Production Act of 2012 to develop a domestic production of 99Mo and to eliminate the use of highly enriched uranium (HEU) in the production of 99Mo. One of the advantages of high power electron linear accelerators (linacs) is they can create both proton- and neutron-rich isotopes by generating high energy x-rays that knock out protons or neutrons from stable atoms or by fission of uranium. This allows for production of isotopes not possible in nuclear reactors. Recent advances in superconducting electron linacs have decreased the size and complexity of these systems such that they are economically competitive with nuclear reactors and large, high energy accelerators. Niowave, Inc. has been developing a radioisotope production facility based on a superconducting electron linac with liquid metal converters.« less

78 FR 79506 - Tennessee Valley Authority: Exemption From Requirements To Revise Combined License Application To...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-30

..., Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington DC, 20555- 0001; telephone: 301...) application for two units of Westinghouse Electric Company's AP1000 advanced pressurized water reactors to be... Bellefonte Nuclear Plant, Units 3 and 4 (BLN 3&4) COL application on January 28, 2008. On September 29, 2010...

Subcritical unity for the Argonaut reactor (in Portuguese)

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

Mongiovi, G.; Aghina, L.O.B.

1971-04-01

tubetype fuel elements aiming at the construction of a subcritical unit employing the internal thermal column of an Argonaut reactor as a source. The results confirmed the feasibility of the use of natural UO/sub 2/ for the proposed arrangement as long as one has a strong source or a subcritical unit diameter greater than 100 cm. (INIS)

An Introduction to the Issues

ERIC Educational Resources Information Center

Primack, Joel

1975-01-01

The reactor safety controversy is reviewed in light of the United States Atomic Energy Commission's Reactor Safety Study and the Report to the American Physical Society by the Study Group on Light Water Reactor Safety. Areas of agreement and disagreement are identified and implications for national policy are explored. (BT)

60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

NASA Astrophysics Data System (ADS)

van Duysen, J. C.; Meric de Bellefon, G.

2017-02-01

The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

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

Benoit, Timothy; Hlotke, John Daniel; Yacout, Abdellatif

2017-07-05

This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generatedmore » during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.« less

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.

A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

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

Wheeler, Timothy A.; Liao, Huafei

2014-12-01

United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constitutedmore » a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.« less

A Monte Carlo program to calculate the exposure rate from airborne radioactive gases inside a nuclear reactor containment building.

PubMed

Sherbini, S; Tamasanis, D; Sykes, J; Porter, S W

1986-12-01

A program was developed to calculate the exposure rate resulting from airborne gases inside a reactor containment building. The calculations were performed at the location of a wall-mounted area radiation monitor. The program uses Monte Carlo techniques and accounts for both the direct and scattered components of the radiation field at the detector. The scattered component was found to contribute about 30% of the total exposure rate at 50 keV and dropped to about 7% at 2000 keV. The results of the calculations were normalized to unit activity per unit volume of air in the containment. This allows the exposure rate readings of the area monitor to be used to estimate the airborne activity in containment in the early phases of an accident. Such estimates, coupled with containment leak rates, provide a method to obtain a release rate for use in offsite dose projection calculations.

Increased occupational radiation doses: nuclear fuel cycle.

PubMed

Bouville, André; Kryuchkov, Victor

2014-02-01

The increased occupational doses resulting from the Chernobyl nuclear reactor accident that occurred in Ukraine in April 1986, the reactor accident of Fukushima that took place in Japan in March 2011, and the early operations of the Mayak Production Association in Russia in the 1940s and 1950s are presented and discussed. For comparison purposes, the occupational doses due to the other two major reactor accidents (Windscale in the United Kingdom in 1957 and Three Mile Island in the United States in 1979) and to the main plutonium-producing facility in the United States (Hanford Works) are also covered but in less detail. Both for the Chernobyl nuclear reactor accident and the routine operations at Mayak, the considerable efforts made to reconstruct individual doses from external irradiation to a large number of workers revealed that the recorded doses had been overestimated by a factor of about two.Introduction of Increased Occupational Exposures: Nuclear Industry Workers. (Video 1:32, http://links.lww.com/HP/A21).

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

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

J. Rempe; D. Knudson; J. Daw

2014-01-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation.more » To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.« less

Fission Surface Power Technology Development Update

NASA Technical Reports Server (NTRS)

Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

2011-01-01

Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power conversion unit with electrical controls, and a heat rejection system with a multi-panel radiator assembly. Testing is planned at the Glenn Research Center Vacuum Facility 6 starting in 2012, with vacuum and liquid-nitrogen cold walls to provide simulation of operationally relevant environments. A nominal two-year test campaign is planned including a Phase 1 reactor simulator and power conversion test followed by a Phase 2 integrated system test with radiator panel heat rejection. The testing is expected to demonstrate the readiness and availability of fission surface power as a viable power system option for NASA's exploration needs. In addition to surface power, technology development work within this project is also directly applicable to in-space fission power and propulsion systems.

Removal of gasoline volatile organic compounds via air biofiltration

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

Miller, R.S.; Saberiyan, A.G.; Esler, C.T.

1995-12-31

Volatile organic compounds (VOCs) generated by vapor extraction and air-stripping systems can be biologically treated in an air biofiltration unit. An air biofilter consists of one or more beds of packing material inoculated with heterotrophic microorganisms capable of degrading the organic contaminant of concern. Waste gases and oxygen are passed through the inoculated packing material, where the microorganisms will degrade the contaminant and release CO{sub 2} + H{sub 2}O. Based on data obtained from a treatability study, a full-scale unit was designed and constructed to be used for treating gasoline vapors generated by a vapor-extraction and groundwater-treatment system at amore » site in California. The unit is composed of two cylindrical reactors with a total packing volume of 3 m{sup 3}. Both reactors are packed with sphagnum moss and inoculated with hydrocarbon-degrading microorganisms of Pseudomonas and Arthrobacter spp. The two reactors are connected in series for air-flow passage. Parallel lines are used for injection of water, nutrients, and buffer to each reactor. Data collected during the startup program have demonstrated an air biofiltration unit with high organic-vapor-removal efficiency.« less

Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

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

Kulesza, J.A.; Fero, A.H.; Rouden, J.

2011-07-01

A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the {sup 93}Nb (n,n') {sup 93m}Nb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

Fukushima Daiichi Muon Imaging

NASA Astrophysics Data System (ADS)

Miyadera, Haruo

2015-10-01

Japanese government announced cold-shutdown condition of the reactors at Fukushima Daiichi by the end of 2011, and mid- and long-term roadmap towards decommissioning has been drawn. However, little is known for the conditions of the cores because access to the reactors has been limited by the high radiation environment. The debris removal from the Unit 1 - 3 is planned to start as early as 2020, but the dismantlement is not easy without any realistic information of the damage to the cores, and the locations and amounts of the fuel debris. Soon after the disaster of Fukushima Daiichi, several teams in the US and Japan proposed to apply muon transmission or scattering imagings to provide information of the Fukushima Daiichi reactors without accessing inside the reactor building. GEANT4 modeling studies of Fukushima Daiichi Unit 1 and 2 showed clear superiority of the muon scattering method over conventional transmission method. The scattering method was demonstrated with a research reactor, Toshiba Nuclear Critical Assembly (NCA), where a fuel assembly was imaged with 3-cm resolution. The muon scattering imaging of Fukushima Daiichi was approved as a national project and is aiming at installing muon trackers to Unit 2. A proposed plan includes installation of muon trackers on the 2nd floor (operation floor) of turbine building, and in front of the reactor building. Two 7mx7m detectors were assembled at Toshiba and tested.

MTR WING, TRA604. ONE OF THE LABORATORY UNITS ALONG THE ...

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

MTR WING, TRA-604. ONE OF THE LABORATORY UNITS ALONG THE SOUTH SIDE WALL. NOTE SINK, CABINET, TABLE, AND HOOD UNITS. DUCT ABOVE RECEIVES CONTAMINATED AIR AND SENDS IT TO FAN HOUSE AND STACK. NOTE PARTITION WALL BEHIND WORK UNITS. THE HEALTH PHYSICS LAB WAS SIMILARLY EQUIPPED. WINDOW AT LEFT EDGE OF VIEW. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON ORIGINAL NEGATIVE. INL NEGATIVE NO. 4225. Unknown Photographer, 2/13/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Chemical Reactors.

ERIC Educational Resources Information Center

Kenney, C. N.

1980-01-01

Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

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

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

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

John Darrell Bess

2008-06-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, suchmore » as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.« less

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

NASA Astrophysics Data System (ADS)

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

2004-02-01

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

Detection of on-surface objects with an underground radiography detector system using cosmic-ray muons

NASA Astrophysics Data System (ADS)

Fujii, Hirofumi; Hara, Kazuhiko; Hayashi, Kohei; Kakuno, Hidekazu; Kodama, Hideyo; Nagamine, Kanetada; Sato, Kazuyuki; Sato, Kotaro; Kim, Shin-Hong; Suzuki, Atsuto; Takahashi, Kazuki; Takasaki, Fumihiko

2017-05-01

We have developed a compact muon radiography detector to investigate the status of the nuclear debris in the Fukushima Daiichi Reactors. Our previous observation showed that a large portion of the Unit-1 Reactor fuel had fallen to floor level. The detector must be located underground to further investigate the status of the fallen debris. To investigate the performance of muon radiography in such a situation, we observed 2 m cubic iron blocks located on the surface of the ground through different lengths of ground soil. The iron blocks were imaged and their corresponding iron density was derived successfully.

Solid Polymer Electrolyte Fuel Cell Technology Program

NASA Technical Reports Server (NTRS)

1980-01-01

Work is reported on phase 5 of the Solid Polymer Electrolyte (SPE) Fuel Cell Technology Development program. The SPE fuel cell life and performance was established at temperatures, pressures, and current densities significantly higher than those previously demonstrated in sub-scale hardware. Operation of single-cell Buildup No. 1 to establish life capabilities of the full-scale hardware was continued. A multi-cell full-scale unit (Buildup No. 2) was designed, fabricated, and test evaluated laying the groundwork for the construction of a reactor stack. A reactor stack was then designed, fabricated, and successfully test-evaluated to demonstrate the readiness of SPE fuel cell technology for future space applications.

Microbial community structures in an integrated two-phase anaerobic bioreactor fed by fruit vegetable wastes and wheat straw.

PubMed

Wang, Chong; Zuo, Jiane; Chen, Xiaojie; Xing, Wei; Xing, Linan; Li, Peng; Lu, Xiangyang; Li, Chao

2014-12-01

The microbial community structures in an integrated two-phase anaerobic reactor (ITPAR) were investigated by 16S rDNA clone library technology. The 75L reactor was designed with a 25L rotating acidogenic unit at the top and a 50L conventional upflow methanogenic unit at the bottom, with a recirculation connected to the two units. The reactor had been operated for 21 stages to co-digest fruit/vegetable wastes and wheat straw, which showed a very good biogas production and decomposition of cellulosic materials. The results showed that many kinds of cellulose and glycan decomposition bacteria related with Bacteroidales, Clostridiales and Syntrophobacterales were dominated in the reactor, with more bacteria community diversities in the acidogenic unit. The methanogens were mostly related with Methanosaeta, Methanosarcina, Methanoculleus, Methanospirillum and Methanobacterium; the predominating genus Methanosaeta, accounting for 40.5%, 54.2%, 73.6% and 78.7% in four samples from top to bottom, indicated a major methanogenesis pathway by acetoclastic methanogenesis in the methanogenic unit. The beta diversity indexes illustrated a more similar distribution of bacterial communities than that of methanogens between acidogenic unit and methanogenic unit. The differentiation of methanogenic community composition in two phases, as well as pH values and volatile fatty acid (VFA) concentrations confirmed the phase separation of the ITPAR. Overall, the results of this study demonstrated that the special designing of ITPAR maintained a sufficient number of methanogens, more diverse communities and stronger syntrophic associations among microorganisms, which made two phase anaerobic digestion of cellulosic materials more efficient. Copyright © 2014. Published by Elsevier B.V.

Development of the Packed Bed Reactor ISS Flight Experiment

NASA Technical Reports Server (NTRS)

Patton, Martin O.; Bruzas, Anthony E.; Rame, Enrique; Motil, Brian J.

2012-01-01

Packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a leading candidate as a potential unit operation in support of long duration human space exploration. On earth, this type of reactor accounts for approximately 80% of all the reactors used in the chemical process industry today. Development of this technology for space exploration is truly crosscutting with many other potential applications (e.g., in-situ chemical processing of planetary materials and transport of nutrients through soil). NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. The first model developed by Motil et al., (2003) is based on a modified Ergun equation. This model was demonstrated at moderate gas and liquid flow rates, but extension to the lower flow rates expected in many advanced life support systems must be validated. The other model, developed by Guo et al., (2004) is based on Darcy s (1856) law for two-phase flow. This model has been validated for a narrow range of flow parameters indirectly (without full instrumentation) and included test points where the flow was not fully developed. The flight experiment presented will be designed with removable test sections to test the hydrodynamic models. The experiment will provide flexibility to test additional beds with different types of packing in the future. One initial test bed is based on the VRA (Volatile Removal Assembly), a packed bed reactor currently on ISS whose behavior in micro-gravity is not fully understood. Improving the performance of this system through an accurate model will increase our ability to purify water in the space environment.

76 FR 77021 - Notice of Availability of Combined License Applications

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-09

... searching on Docket ID NRC-2008-0170 (William States Lee III Nuclear Station Units 1 and 2), NRC-2008-0231...://www.nrc.gov/reactors/new-reactors/col.html . FOR FURTHER INFORMATION CONTACT: Donald Habib, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: (301) 415-1035...

76 FR 71608 - Notice of Availability of Combined License Applications

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-18

... searching on Docket ID NRC-2008-0170 (William States Lee III Nuclear Station Units 1 and 2), NRC-2008-0231...://www.nrc.gov/reactors/new-reactors/col.html . FOR FURTHER INFORMATION CONTACT: Donald Habib, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: (301) 415-1035...

76 FR 75566 - Notice of Availability of Combined License Applications

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-02

....regulations.gov by searching on Docket ID NRC-2008-0170 (William States Lee III Nuclear Station Units 1 and 2... available at http://www.nrc.gov/reactors/new-reactors/col.html . FOR FURTHER INFORMATION CONTACT: Donald Habib, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone...

76 FR 72725 - Notice of Availability of Combined License Applications

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-25

... searching on Docket ID NRC-2008-0170 (William States Lee III Nuclear Station Units 1 and 2), NRC-2008-0231...://www.nrc.gov/reactors/new-reactors/col.html . FOR FURTHER INFORMATION CONTACT: Donald Habib, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: (301) 415-1035...

76 FR 11521 - Prairie Island Nuclear Generating Plant, Unit 1, Northern States Power Company-Minnesota; Notice...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-02

..., Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001..., Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2011-4557 Filed 3-1... NUCLEAR REGULATORY COMMISSION [Docket No. 50-282; NRC-2011-0040] Prairie Island Nuclear Generating...

Atom probe tomography characterizations of high nickel, low copper surveillance RPV welds irradiated to high fluences

NASA Astrophysics Data System (ADS)

Miller, M. K.; Powers, K. A.; Nanstad, R. K.; Efsing, P.

2013-06-01

The Ringhals Units 3 and 4 reactors in Sweden are pressurized water reactors (PWRs) designed and supplied by Westinghouse Electric Company, with commercial operation in 1981 and 1983, respectively. The reactor pressure vessels (RPVs) for both reactors were fabricated with ring forgings of SA 508 class 2 steel. Surveillance blocks for both units were fabricated using the same weld wire heat, welding procedures, and base metals used for the RPVs. The primary interest in these weld metals is because they have very high nickel contents, with 1.58 and 1.66 wt.% for Unit 3 and Unit 4, respectively. The nickel content in Unit 4 is the highest reported nickel content for any Westinghouse PWR. Although both welds contain less than 0.10 wt.% copper, the weld metals have exhibited high irradiation-induced Charpy 41-J transition temperature shifts in surveillance testing. The Charpy impact 41-J shifts and corresponding fluences are 192 °C at 5.0 × 1023 n/m2 (>1 MeV) for Unit 3 and 162 °C at 6.0 × 1023 n/m2 (>1 MeV) for Unit 4. These relatively low-copper, high-nickel, radiation-sensitive welds relate to the issue of so-called late-blooming nickel-manganese-silicon phases. Atom probe tomography measurements have revealed ˜2 nm-diameter irradiation-induced precipitates containing manganese, nickel, and silicon, with phosphorus evident in some of the precipitates. However, only a relatively few number of copper atoms are contained within the precipitates. The larger increase in the transition temperature shift in the higher copper weld metal from the Ringhals R3 Unit is associated with copper-enriched regions within the manganese-nickel-silicon-enriched precipitates rather than changes in their size or number density.

Characterization of reactive CaCO 3 crystallization in a fluidized bed reactor as a central process of direct air capture

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

Burhenne, Luisa; Giacomin, Caroline; Follett, Trevor

A laboratory-scale, fluidized-bed pellet reactor (BPR) was used to investigate a CaCO 3 crystallization process for the recovery of CO 2 in a Direct Air Capture (DAC) process. The BPR performance was validated against data from a pilot-scale unit. Subsequently, the pellet growth under process-relevant conditions was studied over a period of 144 h. The experimental results with the BPR, containing a bed of pellets sized between 0.65 and 0.84 mm, have shown that a calcium retention of 80% can be achieved at a fluidization velocity of 60 m h -1 and a calcium loading rate of 3 mol hmore » -1. This result is consistent with calcium retention observed at pilot scale operation and hence, results from the BPR are considered representative for the pilot scale unit. Starting with a bed of pellets sized between 0.15 and 0.5 mm, the average pellet growth rate, G, at the reactor bottom increased from 8.1E-10 to 11E–10 m s -1 at the onset and decreased to 4.9E–10 m s -1 over the course of a 144 h test. The calcium retention over the course the test showed the same trend (initial increase and final decrease) as the pellet growth rate. A theoretical bed growth model was developed and validated against data from the pilot scale and benchtop pellet reactors. The model was used to calculate the bed porosity and total pellet surface area in each control volume. Lastly, the pellet surface area growth at the bottom of the reactor reproduced the pellet growth and retention data trends.« less

The νGeN experiment at the Kalinin Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Belov, V.; Brudanin, V.; Egorov, V.; Filosofov, D.; Fomina, M.; Gurov, Yu.; Korotkova, L.; Lubashevskiy, A.; Medvedev, D.; Pritula, R.; Rozova, I.; Rozov, S.; Sandukovsky, V.; Timkin, V.; Yakushev, E.; Yurkowski, J.; Zhitnikov, I.

2015-12-01

The ν GeN is new experiment at the Kalinin Nuclear Power Plant (KNPP) for detection of coherent Neutrino-Ge Nucleus elastic scattering. Recent neutrino and Dark Matter search experiments have revolutionized the detection of rear events, and rear events with low energies, in particular. Experiments have achieved sensitivities on the level of several events per hundred kg of detector material per day with energy thresholds from few hundred eV. This opens up a new unique possibility for experimental detection of neutrino-nucleus coherent scattering that has been considered to be impossible so far. The νGeN project uses low threshold high-purity Ge-detectors (HPGe) developed by JINR (Dubna, Russia) in collaboration with BSI (Baltic Scientific Instruments, Riga, Latvia) for creation of a setup designated for first observation of neutrino coherent scattering on Ge. As a powerful neutrino source the experiment will use electron antineutrinos from one of the power-generating units (reactor unit #3) of the KNPP. The coherent neutrino scattering will be observed using a differential method that compares 1) the spectra measured at the reactor operation and shut-down periods; 2) the spectra measured at different distances from the reactor core during the reactor operation. For a setup placed at a 10 m distance from the center of reactor core and with an energy threshold of 350 eV up to tens of events corresponding to neutrino coherent scattering on Ge are expected to be detected per day in the constructed setup with four HPGe low-energy-threshold detectors (~ 400 grams each). The setup sensitivity will be even more increased by using new detectors with total mass up to 5 kg.

Characterization of reactive CaCO 3 crystallization in a fluidized bed reactor as a central process of direct air capture

DOE PAGES

Burhenne, Luisa; Giacomin, Caroline; Follett, Trevor; ...

2017-10-25

A laboratory-scale, fluidized-bed pellet reactor (BPR) was used to investigate a CaCO 3 crystallization process for the recovery of CO 2 in a Direct Air Capture (DAC) process. The BPR performance was validated against data from a pilot-scale unit. Subsequently, the pellet growth under process-relevant conditions was studied over a period of 144 h. The experimental results with the BPR, containing a bed of pellets sized between 0.65 and 0.84 mm, have shown that a calcium retention of 80% can be achieved at a fluidization velocity of 60 m h -1 and a calcium loading rate of 3 mol hmore » -1. This result is consistent with calcium retention observed at pilot scale operation and hence, results from the BPR are considered representative for the pilot scale unit. Starting with a bed of pellets sized between 0.15 and 0.5 mm, the average pellet growth rate, G, at the reactor bottom increased from 8.1E-10 to 11E–10 m s -1 at the onset and decreased to 4.9E–10 m s -1 over the course of a 144 h test. The calcium retention over the course the test showed the same trend (initial increase and final decrease) as the pellet growth rate. A theoretical bed growth model was developed and validated against data from the pilot scale and benchtop pellet reactors. The model was used to calculate the bed porosity and total pellet surface area in each control volume. Lastly, the pellet surface area growth at the bottom of the reactor reproduced the pellet growth and retention data trends.« less

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

NASA Astrophysics Data System (ADS)

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

2005-02-01

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

The status of ABWR-II development

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

Hiroyuki, Okada; Hideya Kitamura; Kumiaki, Moriya

This paper reports on the current development status of the ABWR-II project, a next generation reactor design based on the ABWR. In the early 90's, a program to develop the next generation reactor for the 21. century was launched, at a time when the first ABWR was still under construction. At the initial stage of this project, development of a 'user friendly' plant design was the primary objective. Thus, the main focus was placed on selecting a design with features promoting ease of operation and maintenance. Meanwhile, the circumstances surrounding the Japanese nuclear power industry changed. The delay of FBRmore » development and the deregulation of the power generation market have significantly boosted the role of light water reactors, and accelerated the need to improve LWR economics. For these reasons, economic competitiveness became an overriding objective in the development of the ABWR-II, with no less importance placed on achieving the highest standards of safety. Several new features were adopted to enhance economic performance: 1700 MW electric output, large fuel bundles, simplified MSIV, large capacity SRV. An output of 1700 MWe was selected for compatibility with the Japanese power grid, and with consideration of current reactor pressure vessel manufacturing capability. Large fuel bundles will contribute to a shortened refueling outage period and a reduction of CRDs. For enhanced safety, the reference design implements a modified ECCS with four subdivision RHR, a diversified power source incorporating gas turbine generators (GTG), an advanced RCIC (ARCIC) and passive heat removal systems consisting of a passive containment cooling system (PCCS) and a passive reactor cooling system (PRCS). The modified ECCS configuration also enables on-line maintenance. While current reactors rely on complex accident management (AM) procedures, implemented by operators in the event of a serious accident, the ABWR-II incorporated severe accident countermeasures at the design stage, to eliminate the need of operator induced AM procedures. The ABWR-II represents one of the most promising and reliable options for the future replacement of older units, without incurring excessive R and D costs. (authors)« less

The Fukushima Nuclear Disaster and the U.S. Customs and Border Protection Response

NASA Astrophysics Data System (ADS)

McCormick, Kathy

2013-10-01

On 3/11/11, the reactors at the Fukushima Nuclear Plant in Japan were damaged by a magnitude 9.0 earthquake. Of the six reactors at the site, three were in operation prior to the event, and were automatically shut-down during the earthquake. Emergency cooling systems came online and were subsequently destroyed by a tsunami generated by the earthquake. For the operating reactors, all the reactor cores were exposed, resulting in overheating and the release of steam and hydrogen gas to the containment vessels, several of which subsequently exploded, releasing radioactivity into the atmosphere. The cores of the operating reactors melted down, and radioactive water was released to the ocean in cooling efforts. The primary radiation concerns in the United States from the disaster were radioactive plumes driven by westerly winds and contaminated commercial products and travelers. In the United States, one of the primary governmental organizations to respond to the disaster was U.S. Customs and Border Protection (CBP), which has responsibility to oversee the safety and security of cargo and travelers entering the United States. This talk will describe the various types of radioactive commodities and events encountered by CBP in the U.S. from the Fukushima disaster. Thanks to the CBP Teleforensics Center for their assistance with this presentation.

Current biotechnological developments in Belgium.

PubMed

Masschelein, C A; Callegari, J P; Laurent, M; Simon, J P; Taeymans, D

1989-01-01

In recent years, actions have been undertaken by the Belgian government to promote process innovation and technical diversification. Research programs are initiated and coordinated by the study committee for biotechnology setup within the Institute for Scientific Research in Industry and Agriculture (IRSIA). As a result of this action, the main areas where biotechnological processes are developed or commercially exploited include plant genetics, protein engineering, hybridoma technology, biopesticides, production by genetic engineering of vaccines and drugs, monoclonal detection of human and animal deseases, process reactors for aerobic and anaerobic wastewater treatment, and genetic modification of yeast and bacteria as a base for biomass and energy. Development research also includes new fermentation technologies principally based on immobilization of microorganisms, reactor design, and optimization of unit operations involved in downstream processing. Food, pharmaceutical, and chemical industries are involved in genetic engineering and biotechnology and each of these sectors is overviewed in this paper.

Dual benefit robotics programs at Sandia National Laboratories

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

Jones, A.T.

Sandia National Laboratories has one of the largest integrated robotics laboratories in the United States. Projects include research, development, and application of one-of-a-kind systems, primarily for the Department of Energy (DOE) complex. This work has been underway for more than 10 years. It began with on-site activities that required remote operation, such as reactor and nuclear waste handling. Special purpose robot systems were developed using existing commercial manipulators and fixtures and programs designed in-house. These systems were used in applications such as servicing the Sandia pulsed reactor and inspecting remote roof bolts in an underground radioactive waste disposal facility. Inmore » the beginning, robotics was a small effort, but with increasing attention to the use of robots for hazardous operations, efforts now involve a staff of more than 100 people working in a broad robotics research, development, and applications program that has access to more than 30 robotics systems.« less

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Technical Reports Server (NTRS)

Martin, James; Salvail, Pat

2003-01-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Astrophysics Data System (ADS)

Martin, James; Salvail, Pat

2004-02-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final ``wet in''. A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/-1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to <10-10 std cc/sec helium and vacuum conditioned at 250 °C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 °C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

TOPAZ II Anti-Criticality Device Rapid Prototype

NASA Astrophysics Data System (ADS)

Campbell, Donald R.; Otting, William D.

1994-07-01

The Ballistic Missile Defense Organization (BMDO) has been working on a Nuclear Electric Propulsion Space Test Project (NEPSTP) using an existing Russian Topaz II reactor system to power the NEPSTP satellite. Safety investigations have shown that it will be possible to safely launch the Topaz II system in the United States with some modification to preclude water flooded criticality. A ``fuel-out'' water subcriticality concept was selected by the Los Alamos National Laboratory (LANL) as the baseline concept. A fuel-out anti-criticality device (ACD) conceptual design was developed by Rockwell. The concept functions to hold the fuel from the four centermost thermionic fuel elements (TFEs) outside the reactor during launch and reliably inserts the fuel into the reactor once the operational orbit is achieved. A four-tenths scale ACD rapid prototype model, fabricated from the CATIA solids design model, clearly shows in three dimensions the relative size and spatial relationship of the ACD components.

Nuclear reactor

DOEpatents

Pennell, William E.; Rowan, William J.

1977-01-01

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

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

NONE

1993-09-15

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

Development and Deployment Assessment of a Melt-Down Proof Modular Micro Reactor (MDP-MMR)

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

Hawari, Ayman I.; Venneri, Francesco

The objective of this project is to perform feasibility assessment and technology gap analysis and establish a development roadmap for an innovative and highly compact Micro Modular Reactor (MMR) concept that integrates power production, power conversion and electricity generation in a single unit. The MMR is envisioned to use fully ceramic micro-encapsulated (FCM) fuel, a particularly robust form of TRISO fuel, and to be gas-cooled (e.g., He or CO 2) and capable of generating power in the range of 10 to 40 MW-thermal. It is designed to be absolutely melt-down proof (MDP) under all circumstances including complete loss of coolantmore » scenarios with no possible release of radioactive material, to be factory produced, to have a cycle length of greater than 20 years, and to be highly proliferation resistant. In addition, it will be transportable, retrievable and suitable for use in remote areas. As such, the MDP-MMR will represent a versatile reactor concept that is suitable for use in various applications including electricity generation, process heat utilization and propulsion.« less

VERA Core Simulator Methodology for PWR Cycle Depletion

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

Kochunas, Brendan; Collins, Benjamin S; Jabaay, Daniel

2015-01-01

This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclearmore » reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.« less

The Nuclear Renaissance — Implications on Quantitative Nondestructive Evaluations

NASA Astrophysics Data System (ADS)

Matzie, Regis A.

2007-03-01

The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not even have access to electricity. With this increased energy demand and the high and volatile price of fossil fuels, nuclear energy is experiencing resurgence. This so-called nuclear renaissance is broad based, reaching across Asia, the United States, Europe, as well as selected countries in Africa and South America. Some countries, such as Italy, that have actually turned away from nuclear energy are reconsidering the advisability of this design. This renaissance provides the opportunity to deploy more advanced reactor designs that are operating today, with improved safety, economy, and operations. In this keynote address, I will briefly present three such advanced reactor designs in whose development Westinghouse is participating. These designs include the advanced passive PWR, AP1000, which recently received design certification for the US Nuclear Regulatory Commission; the Pebble Bed Modular reactor (PBMR) which is being demonstrated in South Africa; and the International Reactor Innovative and Secure (IRIS), which was showcased in the US Department of Energy's recently announced Global Nuclear Energy Partnership (GNEP), program. The salient features of these designs that impact future requirements on quantitative nondestructive evaluations will be discussed. Such features as reactor vessel materials, operating temperature regimes, and new geometric configurations will be described, and mention will be made of the impact on quantitative nondestructive evaluation (NDE) approaches.

Modelling of sequential groundwater treatment with zero valent iron and granular activated carbon.

PubMed

Bayer, Peter; Finkel, Michael

2005-06-01

Multiple contaminant mixtures in groundwater may not efficiently be treated by a single technology if contaminants possess rather different properties with respect to sorptivity, solubility, and degradation potential. An obvious choice is to use sequenced units of the generally accepted treatment materials zero valent iron (ZVI) and granular activated carbon (GAC). However, as the results of this modelling study suggest, the required dimensions of both reactor units may strongly differ from those expected on the grounds of a contaminant-specific design. This is revealed by performing an analysis for a broad spectrum of design alternatives through numerical experiments for selected patterns of contaminant mixtures consisting of monochlorobenzene, tetrachloroethylene, trichloroethylene (TCE), cis-1,2-dichloroethylene (cis-DCE), and vinyl chloride (VC). It is shown that efficient treatment can be achieved only if competitive sorption effects in the GAC unit as well as the formation of intermediate products in the ZVI unit are carefully taken into account. Cost-optimal designs turned out to vary extremely depending on the prevailing conditions concerning contaminant concentrations, branching ratios, and unit costs of both reactor materials. Where VC is the critical contaminant, due to high initial concentration or extensive production as an intermediate, two options are cost-effective: an oversized ZVI unit with an oversized GAC unit or a pure GAC reactor.

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

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

None

2013-09-25

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

ScienceCinema

None

2018-01-16

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

Phased Development of Accident Tolerant Fue

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

Bragg-Sitton, Shannon M.; Carmack, W. Jon

2016-09-01

The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modelingmore » predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology readiness levels (TRL) defined for nuclear fuel development. TRL 1–3 corresponds to the “proof-of-concept” stage (Phase 1), TRL 4–6 to “proof-of-principle” (Phase 2), and TRL 7–9 to “proof-of-performance” (Phase 3). This paper will provide an overview of the anticipated activities within each phase of development and will provide an update on the current ATF development status.« less

Development of New Transportation/Storage Cask System for Use by DOE Russian Research Reactor Fuel Return Program

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

Michael Tyacke; Frantisek Svitak; Jiri Rychecky

2010-04-01

The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian supplied high-enriched uranium (HEU) fuel currently stored at Russian-designed research reactors throughout the world to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions for these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design,more » licensing, testing, and delivery of this new cask system are the results of a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: (1) Introduction/Background; (2) VPVR/M Cask Description; (3) Ancillary Equipment, (4) Cask Licensing; (5) Cask Demonstration and Operations; (6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, (7) Summary and Conclusions.« less

Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

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

Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky

2007-10-01

The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing,more » testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.« less

Design and testing of the reactor-internal hydraulic control rod drive for the nuclear heating plant

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

Batheja, P.; Meier, W.J.; Rau, P.J.

A hydraulically driven control rod is being developed at Kraftwerk Union for integration in the primary system of a small nuclear district heating reactor. An elaborate test program, under way for --3 yr, was initiated with a plexiglass rig to understand the basic principles. A design specification list was prepared, taking reactor boundary conditions and relevant German rules and regulations into account. Subsequently, an atmospheric loop for testing of components at 20 to 90/sup 0/C was erected. The objectives involved optimization of individual components such as a piston/cylinder drive unit, electromagnetic valves, and an ultrasonic position indication system as wellmore » as verification of computer codes. Based on the results obtained, full-scale components were designed and fabricated for a prototype test rig, which is currently in operation. Thus far, all atmospheric tests in this rig have been completed. Investigations under reactor temperature and pressure, followed by endurance tests, are under way. All tests to date have shown a reliable functioning of the hydraulic drive, including a novel ultrasonic position indication system.« less

Treating contaminated organics using the DETOX process

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

Elsberry, K.D.; Dhooge, P.M.

1993-05-01

Waste matrices containing organics, radionuclides, and metals pose difficult problems in waste treatment and disposal when the organic compounds and/or metals are considered to be hazardous. This paper describes the results of bench-scale studies of DETOX applied to the components of liquid mixed wastes, with the goal of establishing parameters for designing a prototype waste treatment unit. Apparent organic reaction rate orders and the dependence of apparent reaction rate on solution composition and the contact area were measured for vacuum pump oil scintillation fluids, and trichloroethylene. Reaction rate was superior in chloride-based solutions and was proportional to the contact areamore » above about 2% w/w loading of organic. Oxidations in a 4-liter volume, mixed bench-top reactor have given destruction efficiencies of 99.9999 + % for common organics. Reaction rates achieved in the mixed bench-top reactor were one to two orders of magnitude greater than had been achieved in unmixed reactions; a thoroughly mixed reactor should be capable of oxidizing 10 to 100 + grams of organic per liter-hour. Results are also presented on the solvation efficiency of DETOX for mercury, cerium, and neodymium, and for removal/destruction of organics sorbed on vermiculite. The next stage of development will be converting the bench-top unit to continuous processing.« less

A Lab-on-Chip Design for Miniature Autonomous Bio-Chemoprospecting Planetary Rovers

NASA Astrophysics Data System (ADS)

Santoli, S.

The performance of the so-called ` Lab-on-Chip ' devices, featuring micrometre size components and employed at present for carrying out in a very fast and economic way the extremely high number of sequence determinations required in genomic analyses, can be largely improved as to further size reduction, decrease of power consumption and reaction efficiency through development of nanofluidics and of nano-to-micro inte- grated systems. As is shown, such new technologies would lead to robotic, fully autonomous, microwatt consumption and complete ` laboratory on a chip ' units for accurate, fast and cost-effective astrobiological and planetary exploration missions. The theory and the manufacturing technologies for the ` active chip ' of a miniature bio/chemoprospecting planetary rover working on micro- and nanofluidics are investigated. The chip would include micro- and nanoreactors, integrated MEMS (MicroElectroMechanical System) components, nanoelectronics and an intracavity nanolaser for highly accurate and fast chemical analysis as an application of such recently introduced solid state devices. Nano-reactors would be able to strongly speed up reaction kinetics as a result of increased frequency of reactive collisions. The reaction dynamics may also be altered with respect to standard macroscopic reactors. A built-in miniature telemetering unit would connect a network of other similar rovers and a central, ground-based or orbiting control unit for data collection and transmission to an Earth-based unit through a powerful antenna. The development of the ` Lab-on-Chip ' concept for space applications would affect the economy of space exploration missions, as the rover's ` Lab-on-Chip ' development would link space missions with the ever growing terrestrial market and business concerning such devices, largely employed in modern genomics and bioinformatics, so that it would allow the recoupment of space mission costs.

Revised FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004) 2018-SR-02-1

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

Erika Bailey

2011-10-27

The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in Decembermore » 1960 and criticality was achieved in August 1963. The reactor was tested at low power during the first couple years of operation. Power ascension testing above 1 MW commenced in December 1965 immediately following the receipt of a high-power operating license. In October 1966 during power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant flow to some fuel subassemblies. Two subassemblies started to melt and the reactor was manually shut down. No abnormal releases to the environment occurred. Forty-two months later after the cause had been determined, cleanup completed, and the fuel replaced, Fermi 1 was restarted. However, in November 1972, PRDC made the decision to decommission Fermi 1 as the core was approaching its burn-up limit. The fuel and blanket subassemblies were shipped off-site in 1973. Following that, the secondary sodium system was drained and sent off-site. The radioactive primary sodium was stored on-site in storage tanks and 55 gallon (gal) drums until it was shipped off-site in 1984. The initial decommissioning of Fermi 1 was completed in 1975. Effective January 23, 1976, DPR-9 was transferred to the Detroit Edison Company (DTE) as a 'possession only' license (DTE 2010a). This report details the confirmatory activities performed during the second Oak Ridge Institute for Science and Education (ORISE) site visit to Fermi 1 in November 2010. The survey was strategically planned during a Unit 2 (Fermi 2) outage to take advantage of decreased radiation levels that were observed and attributed to Fermi 2 from the operating unit during the first site visit. However, during the second visit there were elevated radiation levels observed and attributed to the partially dismantled Fermi 1 reactor vessel and a waste storage box located on the 3rd floor of the Fermi 1 Turbine Building. Confirmatory surveys (unshielded) performed directly in the line of sight of these areas were affected. The objective of the confirmatory survey was to verify that the final radiological conditions were accurately and adequately described in Final Status Survey (FSS) documentation, relative to the established release criteria. This objective was achieved by performing document reviews, as well as independent measurements and sampling. Specifically, documentation of the planning, implementation, and results of the FSS were evaluated; side-by-side FSS measurement and source comparisons were performed; site areas were evaluated relative to appropriate FSS classification; and areas were assessed for residual, undocumented contamination.« less

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

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

Smith, P.; Sheetz, S.

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case ofmore » Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL Technical Advisory Committee (TAC) ran a Monte Carlo N-Particle (MCNP) model of a basic SFR for comparison. A 600MWth core surrounded by a lithium blanket produced approximately 1,000 grams of tritium annually with a 13% enriched, 6 year core. This is similar results to a mid-1990’s study where the Fast Flux Test Facility (FFTF), a 400 MWth reactor at the Idaho National Laboratory (INL), could produce about 1,000 grams with an external lithium target. Normalized to the LWRs values, comparative tritium production for an SFR could be approximately 0.31 g-T/kg LEU.« less

Computer simulation of the NASA water vapor electrolysis reactor

NASA Technical Reports Server (NTRS)

Bloom, A. M.

1974-01-01

The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.

9 CFR 147.16 - Procedure for the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment).

Code of Federal Regulations, 2014 CFR

2014-01-01

... mycoplasma reactors by in vivo bio-assay (enrichment). 147.16 Section 147.16 Animals and Animal Products... the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment). This procedure has been shown... publications: (a) Bigland, C. H. and A. J. DaMassa, “A Bio-Assay for Mycoplasma Gallisepticum.” In: United...

9 CFR 147.16 - Procedure for the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment).

Code of Federal Regulations, 2012 CFR

2012-01-01

... mycoplasma reactors by in vivo bio-assay (enrichment). 147.16 Section 147.16 Animals and Animal Products... the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment). This procedure has been shown... publications: (a) Bigland, C. H. and A. J. DaMassa, “A Bio-Assay for Mycoplasma Gallisepticum.” In: United...

9 CFR 147.16 - Procedure for the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment).

Code of Federal Regulations, 2013 CFR

2013-01-01

... mycoplasma reactors by in vivo bio-assay (enrichment). 147.16 Section 147.16 Animals and Animal Products... the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment). This procedure has been shown... publications: (a) Bigland, C. H. and A. J. DaMassa, “A Bio-Assay for Mycoplasma Gallisepticum.” In: United...

9 CFR 147.16 - Procedure for the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment).

Code of Federal Regulations, 2010 CFR

2010-01-01

... mycoplasma reactors by in vivo bio-assay (enrichment). 147.16 Section 147.16 Animals and Animal Products... the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment). This procedure has been shown... publications: (a) Bigland, C. H. and A. J. DaMassa, “A Bio-Assay for Mycoplasma Gallisepticum.” In: United...

Main steam-line break core shroud loading calculations for BWRs

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

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1995-12-31

In July 1994, the U.S. Nuclear regulatory Commission sent out Generic Letter 94-03 to all boiling water reactors in the United States, informing them of intergranular stress corrosion cracking of core shrouds found in 2 reactors. The letter directed all to perform safety analysis of the BWR units. Penn State performed scoping calculations to determine the forces experienced by the core shroud during a main-stream line break transient.

Nonproliferation and Threat Reduction Assistance: U.S. Programs in the Former Soviet Union

DTIC Science & Technology

2009-07-31

seeks to help Russia reconfigure its large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious...opting instead for the construction of fast breeder reactors that could burn plutonium directly for energy production. The United States might not fund...this effort, as many in the United States argue that breeder reactors , which produce more plutonium than they consume, would undermine

Nonproliferation and Threat Reduction Assistance: U.S. Programs in the Former Soviet Union

DTIC Science & Technology

2011-04-26

large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious diseases. The Global Threat Reduction...indicated that it may not pursue the MOX program to eliminate its plutonium, opting instead for the construction of fast breeder reactors that could...burn plutonium directly for energy production. The United States might not fund this effort, as many in the United States argue that breeder reactors

Unit mechanisms of fission gas release: Current understanding and future needs

DOE PAGES

Tonks, Michael; Andersson, David; Devanathan, Ram; ...

2018-03-01

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. Here, this basic understanding of the fission gas behavior mechanisms has the potentialmore » to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Unit mechanisms of fission gas release: Current understanding and future needs

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

Tonks, Michael; Andersson, David; Devanathan, Ram

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. Here, this basic understanding of the fission gas behavior mechanisms has the potentialmore » to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Unit mechanisms of fission gas release: Current understanding and future needs

NASA Astrophysics Data System (ADS)

Tonks, Michael; Andersson, David; Devanathan, Ram; Dubourg, Roland; El-Azab, Anter; Freyss, Michel; Iglesias, Fernando; Kulacsy, Katalin; Pastore, Giovanni; Phillpot, Simon R.; Welland, Michael

2018-06-01

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

Nuclear Security: Action May Be Needed to Reassess the Security of NRC-Licensed Research Reactors. Report to the Ranking Member, Subcommittee on National Security and Foreign Affairs, Committee on Oversight and Government Reform, House of Representatives. GAO-08-403

ERIC Educational Resources Information Center

Aloise, Gene

2008-01-01

There are 37 research reactors in the United States, mostly located on college campuses. Of these, 33 reactors are licensed and regulated by the Nuclear Regulatory Commission (NRC). Four are operated by the Department of Energy (DOE) and are located at three national laboratories. Although less powerful than commercial nuclear power reactors,…

DENSITY CONTROL IN A REACTOR

DOEpatents

Marshall, J. Jr.

1961-10-24

A reactor is described in which natural-uranium bodies are located in parallel channels which extend through the graphite mass in a regular lattice. The graphite mass has additional channels that are out of the lattice and contain no uranium. These additional channels decrease in number per unit volume of graphite from the center of the reactor to the exterior and have the effect of reducing the density of the graphite more at the center than at the exterior, thereby spreading neutron activity throughout the reactor. (AEC)

Simulation of an integrated system for the production of methane and single cell protein from biomass

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

Thomas, M.V.

1989-01-01

A numerical model was developed to simulate the operation of an integrated system for the production of methane and single-cell algal protein from a variety of biomass energy crops or waste streams. Economic analysis was performed at the end of each simulation. The model was capable of assisting in the determination of design parameters by providing relative economic information for various strategies. Three configurations of anaerobic reactors were simulated. These included fed-bed reactors, conventional stirred tank reactors, and continuously expanding reactors. A generic anaerobic digestion process model, using lumped substrate parameters, was developed for use by type-specific reactor models. Themore » generic anaerobic digestion model provided a tool for the testing of conversion efficiencies and kinetic parameters for a wide range of substrate types and reactor designs. Dynamic growth models were used to model the growth of algae and Eichornia crassipes was modeled as a function of daily incident radiation and temperature. The growth of Eichornia crassipes was modeled for the production of biomass as a substrate for digestion. Computer simulations with the system model indicated that tropical or subtropical locations offered the most promise for a viable system. The availability of large quantities of digestible waste and low land prices were found to be desirable in order to take advantage of the economies of scale. Other simulations indicated that poultry and swine manure produced larger biogas yields than cattle manure. The model was created in a modular fashion to allow for testing of a wide variety of unit operations. Coding was performed in the Pascal language for use on personal computers.« less

New functional biocarriers for enhancing the performance of a hybrid moving bed biofilm reactor-membrane bioreactor system.

PubMed

Deng, Lijuan; Guo, Wenshan; Ngo, Huu Hao; Zhang, Xinbo; Wang, Xiaochang C; Zhang, Qionghua; Chen, Rong

2016-05-01

In this study, new sponge modified plastic carriers for moving bed biofilm reactor (MBBR) was developed. The performance and membrane fouling behavior of a hybrid MBBR-membrane bioreactor (MBBR-MBR) system were also evaluated. Comparing to the MBBR with plastic carriers (MBBR), the MBBR with sponge modified biocarriers (S-MBBR) showed better effluent quality and enhanced nutrient removal at HRTs of 12h and 6h. Regarding fouling issue of the hybrid systems, soluble microbial products (SMP) of the MBR unit greatly influenced membrane fouling. The sponge modified biocarriers could lower the levels of SMP in mixed liquor and extracellular polymeric substances in activated sludge, thereby mitigating cake layer and pore blocking resistances of the membrane. The reduced SMP and biopolymer clusters in membrane cake layer were also observed. The results demonstrated that the sponge modified biocarriers were capable of improving overall MBBR performance and substantially alleviated membrane fouling of the subsequent MBR unit. Copyright © 2016 Elsevier Ltd. All rights reserved.

Allothermal steam gasification of biomass in cyclic multi-compartment bubbling fluidized-bed gasifier/combustor - new reactor concept.

PubMed

Iliuta, Ion; Leclerc, Arnaud; Larachi, Faïçal

2010-05-01

A new reactor concept of allothermal cyclic multi-compartment fluidized bed steam biomass gasification is proposed and analyzed numerically. The concept combines space and time delocalization to approach an ideal allothermal gasifier. Thermochemical conversion of biomass in periodic time and space sequences of steam biomass gasification and char/biomass combustion is simulated in which the exothermic combustion compartments provide heat into an array of interspersed endothermic steam gasification compartments. This should enhance unit heat integration and thermal efficiency and procure N(2)-free biosyngas with recourse neither to oxygen addition in steam gasification nor contact between flue and syngas. The dynamic, one-dimensional, multi-component, non-isothermal model developed for this concept accounts for detailed solid and gas flow dynamics whereupon gasification/combustion reaction kinetics, thermal effects and freeboard-zone reactions were tied. Simulations suggest that allothermal operation could be achieved with switch periods in the range of a minute supporting practical feasibility for portable small-scale gasification units. Copyright 2009 Elsevier Ltd. All rights reserved.

Fluidized bed coal combustion reactor

NASA Technical Reports Server (NTRS)

Moynihan, P. I.; Young, D. L. (Inventor)

1981-01-01

A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.

Disparities in nuclear power plant performance in the United States and the Federal Republic of Germany

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

Hansen, K.F.; Winje, D.K.

This report presents data comparing the performance of light water reactors in the United States and the Federal Republic of Germany (FRG). The comparisons are made for the years 1980-1983 and include 21 Westinghouse Pressurized Water Reactors (PWRs), 22 General Electric Boiling Water Reactors (BWRs) in the US; and 6 Kraftwerk Union (KWU) PWRs and 4 KWJ BWRs in the FRG. Data on capacity losses are presented in a disaggregated form for scheduled outages, forced outages, and regulatory imposed outages. Further, within the scheduled and forced outages, the data is subdivided into losses associated with the nuclear island, the balancemore » of plant, or other causes.« less

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

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

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and instructions for both normal and abnormal operating conditions is recogmized. Corfinement of radioactive materials either by tight steel shells, tight buildings, or semi-tight structures vented through filters is considered necessary in the United States. A discussion will be given of specifications, construction, and testing of these structures. The need for emergency plans has been stressed by recent experiences in radioactive releases. The problems of such plans to cover all grades of accidents will be discussed. The theoretical consequences of releases of radioactive materials have been studied and these results will be compared with actual experience. The problem of exposures from normal and abnormal operetion of reactors is a problem of desiga and operation on one hand and the amount of damage to be expected on the other. The safeguard problem is closely related to the acceptable doses of radiouctivity which the ICRP recommend. The future of atomic energy depends upon adequate safeguards and economical design and operation. Accepted criteria are required to guide designers as to the proper balance of caution and boldness. (auth)« less

Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

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

Rempe, Joy Lynn; Knudson, Darrell Lee

2014-09-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2more » sensor survivability and data qualification efforts. This initial review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. Instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations. To encourage the potential for similar activities to be completed for qualifying data from Daiichi Units 1, 2, and 3, this report provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: primary system pressure; containment building temperature; and containment pressure. As described within this report, sensor evaluations and data qualification required implementation of various processes, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design to instruments easily removed from the TMI-2 plant for evaluations. As documented in this report, results from qualifying data for these parameters led to key insights related to TMI-2 accident progression. Hence, these selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are documented in this report to facilitate implementation of similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

Requirements to the procedure and stages of innovative fuel development

NASA Astrophysics Data System (ADS)

Troyanov, V.; Zabudko, L.; Grachyov, A.; Zhdanova, O.

2016-04-01

According to the accepted current understanding under the nuclear fuel we will consider the assembled active zone unit (Fuel assembly) with its structural elements, fuel rods, pellet column, structural materials of fuel rods and fuel assemblies. The licensing process includes justification of safe application of the proposed modifications, including design-basis and experimental justification of the modified items under normal operating conditions and in violation of normal conditions, including accidents as well. Besides the justification of modified units itself, it is required to show the influence of modifications on the performance and safety of the other Reactor Unit’ and Nuclear Plant’ elements (e.g. burst can detection system, transportation and processing operations during fuel handling), as well as to justify the new standards of fuel storage etc. Finally, the modified fuel should comply with the applicable regulations, which often becomes a very difficult task, if only because those regulations, such as the NP-082-07, are not covered modification issues. Making amendments into regulations can be considered as the only solution, but the process is complicated and requires deep grounds for amendments. Some aspects of licensing new nuclear fuel are considered the example of mixed nitride uranium -plutonium fuel application for the BREST reactor unit.

Use of liquid metals in nuclear and thermonuclear engineering, and in other innovative technologies

NASA Astrophysics Data System (ADS)

Rachkov, V. I.; Arnol'dov, M. N.; Efanov, A. D.; Kalyakin, S. G.; Kozlov, F. A.; Loginov, N. I.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

By now, a good deal of experience has been gained with using liquid metals as coolants in nuclear power installations; extensive knowledge has been gained about the physical, thermophysical, and physicochemical properties of these coolants; and the scientific principles and a set of methods and means for handling liquid metals as coolants for nuclear power installations have been elaborated. Prototype and commercialgrade sodium-cooled NPP power units have been developed, including the BOR-60, BN-350, and BN-600 power units (the Soviet Union); the Rapsodie, Phenix, and Superphenix power units (France), the EBR-II power unit (the United States); and the PFR power unit (the United Kingdom). In Russia, dedicated nuclear power installations have been constructed, including those with a lead-bismuth coolant for nuclear submarines and with sodium-potassium alloy for spacecraft (the Buk and Topol installations), which have no analogs around the world. Liquid metals (primarily lithium and its alloy with lead) hold promise for use in thermonuclear power engineering, where they can serve not only as a coolant, but also as tritium-producing medium. In this article, the physicochemical properties of liquid metal coolants, as well as practical experience gained from using them in nuclear and thermonuclear power engineering and in innovative technologies are considered, and the lines of further research works are formulated. New results obtained from investigations carried out on the Pb-Bi and Pb for the SVBR and BREST fast-neutron reactors (referred to henceforth as fast reactors) and for controlled accelerator systems are described.

Recovering tubewise power from three-dimensional nodal kinetics calculation during material relocation in an HWR

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

Kalimullah; Morris, E.E.; Yang, W.S.

1994-12-31

To analyze severe accidents in some special-purpose heavy-water reactors made of assemblies consisting of a number of coaxial tubes of aluminum-clad U-Al fuel and aluminum-clad neutron-capturing material, a mechanistic model, MARTINS, for tube beatup, melting, and molten material relocation has been developed and integrated with the DIF3D nodal hexagonal-z reactor kinetics and other phenomenological modules. The DIF3D kinetics homogenizes all materials located and computes the total power produced in an axial segment of a fuel assembly. This paper presents an approximate method, used in MARTINS, to calculate the distribution of this total nodal power into the intact fuel and capturingmore » material tubes and the meat-cladding mixtures relocating during tube disruption. The method accounts for the change in intraassembly radial power profile due to assembly geometry change with the progress of segment-by-segment disruption of different tubes. Earlier methods to recover pinwise power from nodal calculation for liquid-metal-cooled reactors and light water reactors (X-Y and hexagonal unit cells) are not practical for a disrupting assembly having material relocation. Figure 1 shows the assembly`s end view, divided into rings for modeling and analysis. A ring is a coolant subchannel plus the outer surrounding tube. The present method for distributing the nodal power consists of two parts: (a) calculation of the relative values of ring-by-ring power per unit uranium mass and power per unit mass of neutron-capturing material in a given assembly segment, and (b) normalization of these relative values such that the total power of all rings (intact tubes and U-Al-Cp meat-cladding mixtures, where Cp implies the neutron-capturing material) equals the DIF3D-calculated nodal power for the assembly axial segment.« less

Sequential dark and photo fermentation hydrogen production from hydrolyzed corn stover: A pilot test using 11 m3 reactor.

PubMed

Zhang, Quanguo; Zhang, Zhiping; Wang, Yi; Lee, Duu-Jong; Li, Gang; Zhou, Xuehua; Jiang, Danping; Xu, Bo; Lu, Chaoyang; Li, Yameng; Ge, Xumeng

2018-04-01

Pilot tests of sequential dark and photo fermentation H 2 production were for the first time conducted in a 11 m 3 reactor (3 m 3 for dark and 8 m 3 for photo compartments). A combined solar and light-emitting diode illumination system and a thermal controlling system was installed and tested. With dark fermentation unit maintained at pH 4.5 and 35 °C and photo fermentation unit at pH 7.0 and 30 °C, the overall biogas production rate using hydrolyzed corn stover as substrate reached 87.8 ± 3.8 m 3 /d with 68% H 2 content, contributed by dark unit at 7.5 m 3 -H 2 /m 3 -d and by photo unit at 4.7 m 3 /m 3 -d. Large variation was noted for H 2 production rate in different compartments of the tested units, revealing the adverse effects of poor mixing, washout, and other inhomogeneity associated with large reactor operations. Copyright © 2018 Elsevier Ltd. All rights reserved.

Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

PubMed

Zhang, Yifeng; Angelidaki, Irini

2012-05-15

A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system. Copyright © 2012 Elsevier Ltd. All rights reserved.

Technetium-99m production issues in the United Kingdom.

PubMed

Green, Christopher H

2012-04-01

Nuclear Medicine developed when it was realised that a radioisotopic substitution of Iodine-131 for the stable Iodine-127 would follow the same metabolic pathway in the body enabling the thyroid to be imaged and the thyroid uptake measured. The Iodine could be complexed with pharmaceutical substrates to enable other organs to be imaged, but its use was limited and high gamma energy and beta emission restricted the activity of each radiopharmaceutical used, leading to long acquisition times and degraded images. As a pure gamma emitter of 140 keV and with a 6-h half-life, Technetium-99m is a better radionuclide and images a wider range of bodily organs. However, its short half-life also requires it to be eluted from its mother radionuclide, Mo-99, in a generator, delivered weekly from radiopharmaceutical companies who obtain the Mo-99 in liquid form from high-flux research reactors. All went well till around 2007, when the NRU Reactor in Canada was closed and all other reactors went down for various periods for unrelated problems, leading to widespread Mo-99 shortages. Although the reactors have since recovered, they are 48 to 57 years old, and it seems that few governments have made any future provision such as building replacement reactors.

Use of a Ceramic Membrane to Improve the Performance of Two-Separate-Phase Biocatalytic Membrane Reactor.

PubMed

Ranieri, Giuseppe; Mazzei, Rosalinda; Wu, Zhentao; Li, Kang; Giorno, Lidietta

2016-03-14

Biocatalytic membrane reactors (BMR) combining reaction and separation within the same unit have many advantages over conventional reactor designs. Ceramic membranes are an attractive alternative to polymeric membranes in membrane biotechnology due to their high chemical, thermal and mechanical resistance. Another important use is their potential application in a biphasic membrane system, where support solvent resistance is highly needed. In this work, the preparation of asymmetric ceramic hollow fibre membranes and their use in a two-separate-phase biocatalytic membrane reactor will be described. The asymmetric ceramic hollow fibre membranes were prepared using a combined phase inversion and sintering technique. The prepared fibres were then used as support for lipase covalent immobilization in order to develop a two-separate-phase biocatalytic membrane reactor. A functionalization method was proposed in order to increase the density of the reactive hydroxyl groups on the surface of ceramic membranes, which were then amino-activated and treated with a crosslinker. The performance and the stability of the immobilized lipase were investigated as a function of the amount of the immobilized biocatalytst. Results showed that it is possible to immobilize lipase on a ceramic membrane without altering its catalytic performance (initial residual specific activity 93%), which remains constant after 6 reaction cycles.

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.

Technetium-99m production issues in the United Kingdom

PubMed Central

Green, Christopher H.

2012-01-01

Nuclear Medicine developed when it was realised that a radioisotopic substitution of Iodine-131 for the stable Iodine-127 would follow the same metabolic pathway in the body enabling the thyroid to be imaged and the thyroid uptake measured. The Iodine could be complexed with pharmaceutical substrates to enable other organs to be imaged, but its use was limited and high gamma energy and beta emission restricted the activity of each radiopharmaceutical used, leading to long acquisition times and degraded images. As a pure gamma emitter of 140 keV and with a 6-h half-life, Technetium-99m is a better radionuclide and images a wider range of bodily organs. However, its short half-life also requires it to be eluted from its mother radionuclide, Mo-99, in a generator, delivered weekly from radiopharmaceutical companies who obtain the Mo-99 in liquid form from high-flux research reactors. All went well till around 2007, when the NRU Reactor in Canada was closed and all other reactors went down for various periods for unrelated problems, leading to widespread Mo-99 shortages. Although the reactors have since recovered, they are 48 to 57 years old, and it seems that few governments have made any future provision such as building replacement reactors. PMID:22557795

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

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

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.« less

The UASB reactor as an alternative for the septic tank for on-site sewage treatment.

PubMed

Coelho, A L S S; do Nascimento, M B H; Cavalcanti, P F F; van Haandel, A C

2003-01-01

Although septic tanks are amply used for on site sewage treatment, these units have serious drawbacks: the removal efficiency of organic material and suspended solids is low, the units are costly and occupy a large area and operational cost is high due to the need for periodic desludging. In this paper an innovative variant of the UASB reactor is proposed as an alternative for the septic tank. This alternative has several important advantages in comparison with the conventional septic tank: (1) Although the volume of the UASB reactor was about 4 times smaller than the septic tank, its effluent quality was superior, even though small sludge particles were present, (2) desludging of the UASB reactor is unnecessary and even counterproductive, as the sludge mass guarantees proper performance, (3) the UASB reactor is easily transportable (compact and light) and therefore can be produced in series, strongly reducing construction costs and (4) since the concentration of colloids in the UASB effluent is much smaller than in the ST effluent, it is expected that the infiltration of the effluent will be much less problematic.

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-10-28

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008; and, Chris...related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two consulting and

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-07-17

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...planned nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008...contracting between U.S. firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-12-23

reactors deployed” in the UAE. Some Members of Congress had welcomed the UAE government’s stated commitments not to pursue proliferation-sensitive...for the planned nuclear reactor or on handling spent reactor fuel. (...continued) May...firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two

Nonproliferation and Threat Reduction Assistance: U.S, Programs in the Former Soviet Union

DTIC Science & Technology

2008-03-26

reconfigure its large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious diseases. For FY2004, the Bush...program to eliminate its plutonium, opting instead for the construction of fast breeder reactors that could burn plutonium directly for energy production...The United States might not fund this effort, as many in the United States argue that breeder reactors , which produce more plutonium than they

Nodal weighting factor method for ex-core fast neutron fluence evaluation

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

Chiang, R. T.

The nodal weighting factor method is developed for evaluating ex-core fast neutron flux in a nuclear reactor by utilizing adjoint neutron flux, a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV, the unit fission source, and relative assembly nodal powers. The method determines each nodal weighting factor for ex-core neutron fast flux evaluation by solving the steady-state adjoint neutron transport equation with a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV as the adjoint source, by integrating the unit fission source with a typical fission spectrum to the solved adjointmore » flux over all energies, all angles and given nodal volume, and by dividing it with the sum of all nodal weighting factors, which is a normalization factor. Then, the fast neutron flux can be obtained by summing the various relative nodal powers times the corresponding nodal weighting factors of the adjacent significantly contributed peripheral assembly nodes and times a proper fast neutron attenuation coefficient over an operating period. A generic set of nodal weighting factors can be used to evaluate neutron fluence at the same location for similar core design and fuel cycles, but the set of nodal weighting factors needs to be re-calibrated for a transition-fuel-cycle. This newly developed nodal weighting factor method should be a useful and simplified tool for evaluating fast neutron fluence at selected locations of interest in ex-core components of contemporary nuclear power reactors. (authors)« less

The effects of electric power industry restructuring on the safety of nuclear power plants in the United States

NASA Astrophysics Data System (ADS)

Butler, Thomas S.

Throughout the United States the electric utility industry is restructuring in response to federal legislation mandating deregulation. The electric utility industry has embarked upon an extraordinary experiment by restructuring in response to deregulation that has been advocated on the premise of improving economic efficiency by encouraging competition in as many sectors of the industry as possible. However, unlike the telephone, trucking, and airline industries, the potential effects of electric deregulation reach far beyond simple energy economics. This dissertation presents the potential safety risks involved with the deregulation of the electric power industry in the United States and abroad. The pressures of a competitive environment on utilities with nuclear power plants in their portfolio to lower operation and maintenance costs could squeeze them to resort to some risky cost-cutting measures. These include deferring maintenance, reducing training, downsizing staff, excessive reductions in refueling down time, and increasing the use of on-line maintenance. The results of this study indicate statistically significant differences at the .01 level between the safety of pressurized water reactor nuclear power plants and boiling water reactor nuclear power plants. Boiling water reactors exhibited significantly more problems than did pressurized water reactors.

Unit mechanisms of fission gas release: Current understanding and future needs

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

Tonks, Michael; Andersson, David; Devanathan, Ram

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel properties and, once the gas is released into the gap between the fuel and cladding, lowering gap thermal conductivity and increasing gap pressure. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are being applied to provide unprecedented understanding of the unit mechanisms that define the fission product behavior. In this article, existing research on the basic mechanisms behind the various stages of fission gas releasemore » during normal reactor operation are summarized and critical areas where experimental and simulation work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior during reactor operation and to design fuels that have improved fission product retention. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

NASA Astrophysics Data System (ADS)

Tyapkov, V. F.

2014-07-01

The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

40 CFR 60.611 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-07-01

... From the Synthetic Organic Chemical Manufacturing Industry (SOCMI) Air Oxidation Unit Processes § 60... given them. Air Oxidation Reactor means any device or process vessel in which one or more organic... compounds. Ammoxidation and oxychlorination reactions are included in this definition. Air Oxidation Reactor...

40 CFR 60.611 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-07-01

... From the Synthetic Organic Chemical Manufacturing Industry (SOCMI) Air Oxidation Unit Processes § 60... given them. Air Oxidation Reactor means any device or process vessel in which one or more organic... compounds. Ammoxidation and oxychlorination reactions are included in this definition. Air Oxidation Reactor...

40 CFR 60.611 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-07-01

... From the Synthetic Organic Chemical Manufacturing Industry (SOCMI) Air Oxidation Unit Processes § 60... given them. Air Oxidation Reactor means any device or process vessel in which one or more organic... compounds. Ammoxidation and oxychlorination reactions are included in this definition. Air Oxidation Reactor...

40 CFR 60.611 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... From the Synthetic Organic Chemical Manufacturing Industry (SOCMI) Air Oxidation Unit Processes § 60... given them. Air Oxidation Reactor means any device or process vessel in which one or more organic... compounds. Ammoxidation and oxychlorination reactions are included in this definition. Air Oxidation Reactor...

Model-Based Sensor Placement for Component Condition Monitoring and Fault Diagnosis in Fossil Energy Systems

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

Mobed, Parham; Pednekar, Pratik; Bhattacharyya, Debangsu

Design and operation of energy producing, near “zero-emission” coal plants has become a national imperative. This report on model-based sensor placement describes a transformative two-tier approach to identify the optimum placement, number, and type of sensors for condition monitoring and fault diagnosis in fossil energy system operations. The algorithms are tested on a high fidelity model of the integrated gasification combined cycle (IGCC) plant. For a condition monitoring network, whether equipment should be considered at a unit level or a systems level depends upon the criticality of the process equipment, its likeliness to fail, and the level of resolution desiredmore » for any specific failure. Because of the presence of a high fidelity model at the unit level, a sensor network can be designed to monitor the spatial profile of the states and estimate fault severity levels. In an IGCC plant, besides the gasifier, the sour water gas shift (WGS) reactor plays an important role. In view of this, condition monitoring of the sour WGS reactor is considered at the unit level, while a detailed plant-wide model of gasification island, including sour WGS reactor and the Selexol process, is considered for fault diagnosis at the system-level. Finally, the developed algorithms unify the two levels and identifies an optimal sensor network that maximizes the effectiveness of the overall system-level fault diagnosis and component-level condition monitoring. This work could have a major impact on the design and operation of future fossil energy plants, particularly at the grassroots level where the sensor network is yet to be identified. In addition, the same algorithms developed in this report can be further enhanced to be used in retrofits, where the objectives could be upgrade (addition of more sensors) and relocation of existing sensors.« less

Expert assessments of the cost of light water small modular reactors

PubMed Central

Abdulla, Ahmed; Azevedo, Inês Lima; Morgan, M. Granger

2013-01-01

Analysts and decision makers frequently want estimates of the cost of technologies that have yet to be developed or deployed. Small modular reactors (SMRs), which could become part of a portfolio of carbon-free energy sources, are one such technology. Existing estimates of likely SMR costs rely on problematic top-down approaches or bottom-up assessments that are proprietary. When done properly, expert elicitations can complement these approaches. We developed detailed technical descriptions of two SMR designs and then conduced elicitation interviews in which we obtained probabilistic judgments from 16 experts who are involved in, or have access to, engineering-economic assessments of SMR projects. Here, we report estimates of the overnight cost and construction duration for five reactor-deployment scenarios that involve a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants vary by more than a factor of 2.5. Expert judgments about likely SMR costs display an even wider range. Median estimates for a 45 megawatts-electric (MWe) SMR range from $4,000 to $16,300/kWe and from $3,200 to $7,100/kWe for a 225-MWe SMR. Sources of disagreement are highlighted, exposing the thought processes of experts involved with SMR design. There was consensus that SMRs could be built and brought online about 2 y faster than large reactors. Experts identify more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable. PMID:23716682

Expert assessments of the cost of light water small modular reactors.

PubMed

Abdulla, Ahmed; Azevedo, Inês Lima; Morgan, M Granger

2013-06-11

Analysts and decision makers frequently want estimates of the cost of technologies that have yet to be developed or deployed. Small modular reactors (SMRs), which could become part of a portfolio of carbon-free energy sources, are one such technology. Existing estimates of likely SMR costs rely on problematic top-down approaches or bottom-up assessments that are proprietary. When done properly, expert elicitations can complement these approaches. We developed detailed technical descriptions of two SMR designs and then conduced elicitation interviews in which we obtained probabilistic judgments from 16 experts who are involved in, or have access to, engineering-economic assessments of SMR projects. Here, we report estimates of the overnight cost and construction duration for five reactor-deployment scenarios that involve a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants vary by more than a factor of 2.5. Expert judgments about likely SMR costs display an even wider range. Median estimates for a 45 megawatts-electric (MWe) SMR range from $4,000 to $16,300/kWe and from $3,200 to $7,100/kWe for a 225-MWe SMR. Sources of disagreement are highlighted, exposing the thought processes of experts involved with SMR design. There was consensus that SMRs could be built and brought online about 2 y faster than large reactors. Experts identify more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable.

Modeling phosphorus removal and recovery from anaerobic digester supernatant through struvite crystallization in a fluidized bed reactor.

PubMed

Rahaman, Md Saifur; Mavinic, Donald S; Meikleham, Alexandra; Ellis, Naoko

2014-03-15

The cost associated with the disposal of phosphate-rich sludge, the stringent regulations to limit phosphate discharge into aquatic environments, and resource shortages resulting from limited phosphorus rock reserves, have diverted attention to phosphorus recovery in the form of struvite (MAP: MgNH4PO4·6H2O) crystals, which can essentially be used as a slow release fertilizer. Fluidized-bed crystallization is one of the most efficient unit processes used in struvite crystallization from wastewater. In this study, a comprehensive mathematical model, incorporating solution thermodynamics, struvite precipitation kinetics and reactor hydrodynamics, was developed to illustrate phosphorus depletion through struvite crystal growth in a continuous, fluidized-bed crystallizer. A thermodynamic equilibrium model for struvite precipitation was linked to the fluidized-bed reactor model. While the equilibrium model provided information on supersaturation generation, the reactor model captured the dynamic behavior of the crystal growth processes, as well as the effect of the reactor hydrodynamics on the overall process performance. The model was then used for performance evaluation of the reactor, in terms of removal efficiencies of struvite constituent species (Mg, NH4 and PO4), and the average product crystal sizes. The model also determined the variation of species concentration of struvite within the crystal bed height. The species concentrations at two extreme ends (inlet and outlet) were used to evaluate the reactor performance. The model predictions provided a reasonably good fit with the experimental results for PO4-P, NH4-N and Mg removals. Predicated average crystal sizes also matched fairly well with the experimental observations. Therefore, this model can be used as a tool for performance evaluation and process optimization of struvite crystallization in a fluidized-bed reactor. Crown Copyright © 2013. Published by Elsevier Ltd. All rights reserved.

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

NASA Astrophysics Data System (ADS)

Fic, Adam; Składzień, Jan; Gabriel, Michał

2015-03-01

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

Fuel processors for fuel cell APU applications

NASA Astrophysics Data System (ADS)

Aicher, T.; Lenz, B.; Gschnell, F.; Groos, U.; Federici, F.; Caprile, L.; Parodi, L.

The conversion of liquid hydrocarbons to a hydrogen rich product gas is a central process step in fuel processors for auxiliary power units (APUs) for vehicles of all kinds. The selection of the reforming process depends on the fuel and the type of the fuel cell. For vehicle power trains, liquid hydrocarbons like gasoline, kerosene, and diesel are utilized and, therefore, they will also be the fuel for the respective APU systems. The fuel cells commonly envisioned for mobile APU applications are molten carbonate fuel cells (MCFC), solid oxide fuel cells (SOFC), and proton exchange membrane fuel cells (PEMFC). Since high-temperature fuel cells, e.g. MCFCs or SOFCs, can be supplied with a feed gas that contains carbon monoxide (CO) their fuel processor does not require reactors for CO reduction and removal. For PEMFCs on the other hand, CO concentrations in the feed gas must not exceed 50 ppm, better 20 ppm, which requires additional reactors downstream of the reforming reactor. This paper gives an overview of the current state of the fuel processor development for APU applications and APU system developments. Furthermore, it will present the latest developments at Fraunhofer ISE regarding fuel processors for high-temperature fuel cell APU systems on board of ships and aircrafts.

MODULAR CORE UNITS FOR A NEUTRONIC REACTOR

DOEpatents

Gage, J.F. Jr.; Sherer, D.B.

1964-04-01

A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

Reactivity control assembly for nuclear reactor. [LMFBR

DOEpatents

Bollinger, L.R.

1982-03-17

This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

Nuclear reactors built, being built, or planned, 1991

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

Simpson, B.

1992-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

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

Apparatus and method for continuous production of materials

DOEpatents

Chang, Chih-hung; Jin, Hyungdae

2014-08-12

Embodiments of a continuous-flow injection reactor and a method for continuous material synthesis are disclosed. The reactor includes a mixing zone unit and a residence time unit removably coupled to the mixing zone unit. The mixing zone unit includes at least one top inlet, a side inlet, and a bottom outlet. An injection tube, or plurality of injection tubes, is inserted through the top inlet and extends past the side inlet while terminating above the bottom outlet. A first reactant solution flows in through the side inlet, and a second reactant solution flows in through the injection tube(s). With reference to nanoparticle synthesis, the reactant solutions combine in a mixing zone and form nucleated nanoparticles. The nucleated nanoparticles flow through the residence time unit. The residence time unit may be a single conduit, or it may include an outer housing and a plurality of inner tubes within the outer housing.

75 FR 5357 - In the Matter of Entergy Nuclear Operations, Inc., et al.; Order Extending the Effectiveness of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-02

...). Pilgrim is a boiling water nuclear reactor that is owned by Entergy Nuclear and operated by ENO. The... Generating Unit No. 1 (IP1). IP1 is a pressurized water nuclear reactor that is owned by ENIP2 and maintained... nuclear reactors that are owned by ENIP2 and ENIP3, respectively, and operated by ENO. The facilities are...

Low cost silicon solar array project silicon materials task: Establishment of the feasibility of a process capable of low-cost, high volume production of silane (step 1) and the pyrolysis of silane to semiconductor-grade silicon (step 2)

NASA Technical Reports Server (NTRS)

Breneman, W. C.; Farrier, E. G.; Rexer, J.

1977-01-01

Extended operation of a small process-development unit routinely produced high quality silane in 97+% yield from dichlorosilane. The production rate was consistent with design loadings for the fractionating column and for the redistribution reactor. A glass fluid-bed reactor was constructed for room temperature operation. The behavior of a bed of silcon particles was observed as a function of various feedstocks, component configurations, and operating conditions. For operating modes other than spouting, the bed behaved in an erratic and unstable manner. A method was developed for casting molten silicon powder into crack-free solid pellets for process evaluation. The silicon powder was melted and cast into thin walled quartz tubes that sacrificially broke on cooling.

International nuclear safety concerns. Hearing before the Subcommittee on Energy, Nuclear Proliferation, and Government Processes of the Committee on Governmental Affairs, United States Senate, Ninety-Ninth Congress, Second Session, May 8, 1986

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

Not Available

1986-01-01

Allan I. Mendelowitz of the General Accounting Office (GAO) and James R. Shea of the Nuclear Regulatory Commission were the principal witnesses at a hearing on international concerns about reactor safety. The hearing focused on the Soviet accident at Chernobyl as a demonstration that safety matters are a legitimate area of international concern. Among the issues under discussion were safety standards and inspection procedures of the International Atomic Energy Agency (IAEA). Estimates developed by the GAO show that developing countries, which lack a strong technical base or nuclear background to handle emergencies, will have half the reactors in the worldmore » by the year 2000. Mendelowitz and Shea, together with supporting testimony from others in their agencies described international efforts to improve safeguards.« less

Commander Lousma adjusts MLR controls on middeck

NASA Image and Video Library

1982-03-30

STS003-22-127 (22-30 March 1982) --- Astronaut Jack R. Lousma, STS-3 commander, wearing communications kit assembly (assy) mini-headset, adjusts controls on Monodisperse Latex Reactor (MLR) experiment located in forward middeck lockers MF57H and MF57K. To reach MLR support electronics assy controls, Lousma squeezes in between forward lockers and Development Flight Instrument (DFI) unit on starboard bulkhead. Photo credit: NASA

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

ORNL Named as Part of IAES Research Reactor Hub

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

None

The International Atomic Energy Agency (IAEA) has named ORNL and Idaho National Laboratory part of an International Centre based on Research Reactors. The designation makes the United States one of only three countries identified for unique capabilities and excellence in nuclear research.

Lunar base thermoelectric power station study

NASA Technical Reports Server (NTRS)

Determan, William; Frye, Patrick; Mondt, Jack; Fleurial, Jean-Pierre; Johnson, Ken; Stapfer, G.; Brooks, Michael D.; Heshmatpour, Ben

2006-01-01

Under NASA's Project Prometheus, the Nuclear Systems Program, the Jet Propulsion Laboratory, Pratt & Whitney Rocketdyne, and Teledyne Energy Systems have teamed with a number of universities, under the Segmented Thermoelectric Multicouple Converter (STMC) program, to develop the next generation of advanced thermoelectric converters for space reactor power systems. Work on the STMC converter assembly has progressed to the point where the lower temperature stage of the segmented multicouple converter assembly is ready for laboratory testing and the upper stage materials have been identified and their properties are being characterized. One aspect of the program involves mission application studies to help define the potential benefits from the use of these STMC technologies for designated NASA missions such as the lunar base power station where kilowatts of power are required to maintain a permanent manned presence on the surface of the moon. A modular 50 kWe thermoelectric power station concept was developed to address a specific set of requirements developed for this mission. Previous lunar lander concepts had proposed the use of lunar regolith as in-situ radiation shielding material for a reactor power station with a one kilometer exclusion zone radius to minimize astronaut radiation dose rate levels. In the present concept, we will examine the benefits and requirements for a hermetically-sealed reactor thermoelectric power station module suspended within a man-made lunar surface cavity. The concept appears to maximize the shielding capabilities of the lunar regolith while minimizing its handling requirements. Both thermal and nuclear radiation levels from operation of the station, at its 100-m exclusion zone radius, were evaluated and found to be acceptable. Site preparation activities are reviewed and well as transport issues for this concept. The goal of the study was to review the entire life cycle of the unit to assess its technical problems and technology needs in all areas to support the development, deployment, operation and disposal of the unit.

Enhancing nitrogen removal from low carbon to nitrogen ratio wastewater by using a novel sequencing batch biofilm reactor.

PubMed

Zou, Jinte; Li, Jun; Ni, Yongjiong; Wei, Su

2016-12-01

Removing nitrogen from wastewater with low chemical oxygen demand/total nitrogen (COD/TN) ratio is a difficult task due to the insufficient carbon source available for denitrification. Therefore, in the present work, a novel sequencing batch biofilm reactor (NSBBR) was developed to enhance the nitrogen removal from wastewater with low COD/TN ratio. The NSBBR was divided into two units separated by a vertical clapboard. Alternate feeding and aeration was performed in the two units, which created an anoxic unit with rich substrate content and an aeration unit deficient in substrate simultaneously. Therefore, the utilization of the influent carbon source for denitrification was increased, leading to higher TN removal compared to conventional SBBR (CSBBR) operation. The results show that the CSBBR removed up to 76.8%, 44.5% and 10.4% of TN, respectively, at three tested COD/TN ratios (9.0, 4.8 and 2.5). In contrast, the TN removal of the NSBBR could reach 81.9%, 60.5% and 26.6%, respectively, at the corresponding COD/TN ratios. Therefore, better TN removal performance could be achieved in the NSBBR, especially at low COD/TN ratios (4.8 and 2.5). Furthermore, it is easy to upgrade a CSBBR into an NSBBR in practice. Copyright © 2016. Published by Elsevier B.V.

Returning HEU Fuel from the Czech Republic to Russia

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

Michael Tyacke; Dr. Igor Bolshinsky

In December 1999, representatives from the United States, Russian Federation, and International Atomic Energy Agency began working on a program to return Russian supplied, highly enriched, uranium fuel stored at foreign research reactors to Russia. Now, under the Global Threat Reduction Initiative’s Russian Research Reactor Fuel Return Program, this effort has repatriated over 800 kg of highly enriched uranium to Russia from over 10 countries. In May 2004, the “Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning Cooperation for the Transfer of Russian Produced Research Reactor Nuclear Fuel to themore » Russian Federation” was signed. This agreement provides legal authority for the Russian Research Reactor Fuel Return Program and establishes parameters whereby eligible countries may return highly enriched uranium spent and fresh fuel assemblies and other fissile materials to Russia. On December 8, 2007, one of the largest shipments of highly enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together. In February 2003, Russian Research Reactor Fuel Return Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their highly enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This article discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

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

Camacho-Bunquin, Jeffrey; Shou, Heng; Marshall, Christopher L.

An integrated atomic layer deposition synthesis-catalysis (I-ALD-CAT) tool was developed. It combines an ALD manifold in-line with a plug-flow reactor system for the synthesis of supported catalytic materials by ALD and immediate evaluation of catalyst reactivity using gas-phase probe reactions. The I-ALD-CAT delivery system consists of 12 different metal ALD precursor channels, 4 oxidizing or reducing agents, and 4 catalytic reaction feeds to either of the two plug-flow reactors. The system can employ reactor pressures and temperatures in the range of 10{sup −3} to 1 bar and 300–1000 K, respectively. The instrument is also equipped with a gas chromatograph andmore » a mass spectrometer unit for the detection and quantification of volatile species from ALD and catalytic reactions. In this report, we demonstrate the use of the I-ALD-CAT tool for the synthesis of platinum active sites and Al{sub 2}O{sub 3} overcoats, and evaluation of catalyst propylene hydrogenation activity.« less

Low cost silicon solar array project: Feasibility of low-cost, high-volume production of silane and pyrolysis of silane to semiconductor-grade silicon

NASA Technical Reports Server (NTRS)

Breneman, W. C.

1978-01-01

Silicon epitaxy analysis of silane produced in the Process Development Unit operating in a completely integrated mode consuming only hydrogen and metallurgical silicon resulted in film resistivities of up to 120 ohms cm N type. Preliminary kinetic studies of dichlorosilane disproportionation in the liquid phase have shown that 11.59% SiH4 is formed at equilibrium after 12 minutes contact time at 56 C. The fluid-bed reactor was operated continuously for 48 hours with a mixture of one percent silane in helium as the fluidizing gas. A high silane pyrolysis efficiency was obtained without the generation of excessive fines. Gas flow conditions near the base of the reactor were unfavorable for maintaining a bubbling bed with good heat transfer characteristics. Consequently, a porous agglomerate formed in the lower portion of the reactor. Dense coherent plating was obtained on the silicon seed particles which had remained fluidizied throughout the experiment.

On-board diesel autothermal reforming for PEM fuel cells: Simulation and optimization

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

Cozzolino, Raffaello, E-mail: raffaello.cozzolino@unicusano.it; Tribioli, Laura

2015-03-10

Alternative power sources are nowadays the only option to provide a quick response to the current regulations on automotive pollutant emissions. Hydrogen fuel cell is one promising solution, but the nature of the gas is such that the in-vehicle conversion of other fuels into hydrogen is necessary. In this paper, autothermal reforming, for Diesel on-board conversion into a hydrogen-rich gas suitable for PEM fuel cells, has investigated using the simulation tool Aspen Plus. A steady-state model has been developed to analyze the fuel processor and the overall system performance. The components of the fuel processor are: the fuel reforming reactor,more » two water gas shift reactors, a preferential oxidation reactor and H{sub 2} separation unit. The influence of various operating parameters such as oxygen to carbon ratio, steam to carbon ratio, and temperature on the process components has been analyzed in-depth and results are presented.« less

Integrated horizontal-flow anaerobic and radial-flow aerobic reactors for the removal of organic matter and nitrogen from domestic sewage.

PubMed

Vieira, L G T; Fazolo, A; Zaiat, M; Foresti, E

2003-01-01

This paper presents the conception and discusses the results obtained from the operation of an integrated biological anaerobic/aerobic/anaerobic system composed of horizontal-flow anaerobic and radial-flow aerobic reactors for domestic sewage treatment. The performance of a horizontal-flow anaerobic immobilized biomass reactor, with five stages,followed by a radial-flow aerobic immobilized biomass reactor was evaluated along 22 weeks. After the 14th week, the last stage of the HAIB reactor was used as a denitrifying unit. Polyurethane foam cubic matrices with 1-cm sides were used as support for biomass immobilization in all the units. The influent domestic sewage presented mean chemical oxygen demand of 365 +/- 71 mg. 1(-1) and the temperature was 23 +/- 3degrees C. The integrated system achieved COD removal efficiency of 90% while the maximum ammonium removal efficiency was 97% in the aerobic post-treatment unit. The nitrification process was found to be better represented by first-order reactions in series model. The apparent first-order kinetic coefficient for nitrate formation was about 50 times higher than that estimated for the nitrite formation. The denitrification process was well represented by a Monod-type kinetic model. The maximum specific denitrifying rate and the half-saturation coefficient were 2.9 x 10(-4) mg NO(3)(-)-N mg(-1) VSS h(-1) and 19.4 mg NO(3)(-)-N 1(-1), respectively.

MACHINING TEST SPECIMENS FROM HARVESTED ZION RPV SEGMENTS

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

Nanstad, Randy K; Rosseel, Thomas M; Sokolov, Mikhail A

2015-01-01

The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials,more » structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].« less

Advanced Coal Liquefaction Research and Development Facility, Wilsonville, Alabama. Run 262 with Black Thunder subbituminous coal: Technical progress report

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

Not Available

This report presents the results of Run 262 performed at the Advanced Coal Liquefaction R&D Facility in Wilsonville, Alabama. The run started on July 10, 1991 and continued until September 30, 1991, operating in the Close-Coupled Integrated Two-Stage Liquefaction mode processing Black Thunder Mine subbituminous coal (Wyodak-Anderson seam from Wyoming Powder River Basin). A dispersed molybdenum catalyst was evaluated for its performance. The effect of the dispersed catalyst on eliminating solids buildup was also evaluated. Half volume reactors were used with supported Criterion 324 1/16`` catalyst in the second stage at a catalyst replacement rate of 3 lb/ton of MFmore » coal. The hybrid dispersed plus supported catalyst system was tested for the effect of space velocity, second stage temperature, and molybdenum concentration. The supported catalyst was removed from the second stage for one test period to see the performance of slurry reactors. Iron oxide was used as slurry catalyst at a rate of 2 wt % MF coal throughout the run (dimethyl disulfide (DMDS) was used as the sulfiding agent). The close-coupled reactor unit was on-stream for 1271.2 hours for an on-stream factor of 89.8% and the ROSE-SR unit was on-feed for 1101.6 hours for an on-stream factor of 90.3% for the entire run.« less

Individual dose due to radioactivity accidental release from fusion reactor.

PubMed

Nie, Baojie; Ni, Muyi; Wei, Shiping

2017-04-05

As an important index shaping the design of fusion safety system, evaluation of public radiation consequences have risen as a hot topic on the way to develop fusion energy. In this work, the comprehensive public early dose was evaluated due to unit gram tritium (HT/HTO), activated dust, activated corrosion products (ACPs) and activated gases accidental release from ITER like fusion reactor. Meanwhile, considering that we cannot completely eliminate the occurrence likelihood of multi-failure of vacuum vessel and tokamak building, we conservatively evaluated the public radiation consequences and environment restoration after the worst hypothetical accident preliminarily. The comparison results show early dose of different unit radioactivity release under different conditions. After further performing the radiation consequences, we find it possible that the hypothetical accident for ITER like fusion reactor would result in a level 6 accident according to INES, not appear level 7 like Chernobyl or Fukushima accidents. And from the point of environment restoration, we need at least 69 years for case 1 (1kg HTO and 1000kg dust release) and 34-52years for case 2 (1kg HTO and 10kg-100kg dust release) to wait the contaminated zone drop below the general public safety limit (1mSv per year) before it is suitable for human habitation. Copyright © 2016 Elsevier B.V. All rights reserved.

A simple cost-effective manometric respirometer: design and application in wastewater biomonitoring

NASA Astrophysics Data System (ADS)

Rahman, Mohammad Shahidur; Islam, M. Akhtarul

2015-09-01

Application of respirometric tools in wastewater engineering fields is still not getting familiarity and acceptance by academy or industry in developing countries as compared to the use of conventional biochemical oxygen demand (BOD) approach. To justify the applicability of respirometry, a low-cost respirometric device suitable for monitoring biodegradation process in wastewater has been developed. This device contains six independently operating reactors placed in a temperature control unit for the bioassay of five wastewater samples simultaneously (along with one blank). Each reactor is equipped with a magnetic stirrer for the continuous agitation of the test sample. Six manometers, linked with the individual reactors, measure the pressure and volume changes in the headspace gas phase of the reactor. Working formulae have been derived to convert the `volume-change in gas phase' data to `the oxygen depletion in the whole liquid-gas system' data. The performance of the device has been tested with glucose-glutamic acid standard solution and found satisfactory. Conventional BOD test and the respirometric measurements were performed simultaneously and it is found that in addition to measuring the BOD of the sample, this device gives oxygen uptake profile for further analysis to determine the biokinetic coefficients. Additionally, in some cases, following a specific test protocol, the respirometer can indirectly estimate the carbon dioxide evolved during biodegradation process for calculating respiratory activity parameter such as respiratory quotient. It is concluded that the device can be used in the laboratories associated with the activated sludge plants and also for teaching and research purposes in developing countries.

Steady State Advanced Tokamak (SSAT): The mission and the machine

NASA Astrophysics Data System (ADS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the U.S. National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new 'Steady State Advanced Tokamak' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO.

Feasibility Study of the Geotextile Waste Filtration Unit.

DTIC Science & Technology

2000-02-10

Treatment Module 3-32 Figure 3-20. THE SCHEMATIC OF THE MOVING BED BIOFILM REACTOR ( MBBR ) 3൪ Figure 4-1. The Original Distributed Concept for WFUs...Moving Bed Biofilm Reactor ( MBBR ) process appears to be one of the most feasible processes available to meet Force Provider liquid waste stream...Moving Bed Biofilm Reactor ( MBBR ) process was then examined.31 In this system, both activated sludge and fixed-film processes occur in a bioreactor

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

Light Water Reactor Sustainability Program: Integrated Program Plan

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

None, None

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 60%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to grow by about 24% from 2013 to 2040 . At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license, for a total of 60 years of operation (the oldest commercial plants in the Unitedmore » States reached their 40th anniversary in 2009). Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity for 40- and 60-year license periods. If current operating nuclear power plants do not operate beyond 60 years (and new nuclear plants are not built quickly enough to replace them), the total fraction of generated electrical energy from nuclear power will rapidly decline. That decline will be accelerated if plants are shut down before 60 years of operation. Decisions on extended operation ultimately rely on economic factors; however, economics can often be improved through technical advancements. The U.S. Department of Energy Office of Nuclear Energy’s 2010 Research and Development Roadmap (2010 Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: 1. Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; 2. Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; 3. Develop sustainable nuclear fuel cycles; and 4. Understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.« less

Can we Plan. The political economy of commercial nuclear energy policy in the United States

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

Campbell, J.L. Jr.

1984-01-01

The dissertation is an analysis of the commercial nuclear energy sector's decline in the United States. The research attempts to reconcile the debate between Weberian-institutional and Marxist political theory about the state's inability to successfully plan industrial development in advanced capitalist countries. Synthesizing these views, the central hypothesis guiding the research is that the greater the state's relative autonomy from political and economic constraints in an institutional sense, i.e., the greater its insulation from the contradictions of capitalism and democracy, the greater its planning capacity and the more successful it will be in directing industrial performance. The research examines onemore » industrial sector, commercial nuclear energy, and draws two major comparison. First, the French and US nuclear industries are compared, since the state's relative autonomy is much greater in the former than in the latter. This comparison is developed to identify policy areas where nuclear planning has succeeded in France but failed in America. Four areas are identified: reactor standardization, waste management, reactor safety, and financing. Second, looking particularly at the US, the policy areas are compared to analyze the development of policy and its effects on the sector's performance and to determine the degree to which planning was undermined by the structural constraints characteristic of a state with low relative autonomy.« less

Control of autothermal reforming reactor of diesel fuel

NASA Astrophysics Data System (ADS)

Dolanc, Gregor; Pregelj, Boštjan; Petrovčič, Janko; Pasel, Joachim; Kolb, Gunther

2016-05-01

In this paper a control system for autothermal reforming reactor for diesel fuel is presented. Autothermal reforming reactors and the pertaining purification reactors are used to convert diesel fuel into hydrogen-rich reformate gas, which is then converted into electricity by the fuel cell. The purpose of the presented control system is to control the hydrogen production rate and the temperature of the autothermal reforming reactor. The system is designed in such a way that the two control loops do not interact, which is required for stable operation of the fuel cell. The presented control system is a part of the complete control system of the diesel fuel cell auxiliary power unit (APU).

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

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

Corradini, M. L.; Peko, D.; Farmer, M.

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.« less

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

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

Corradini, M. L.

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.« less

Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

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

Clarno, Kevin T.

2016-01-30

The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, andmore » at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.« less

Characterization of 14C in Swedish light water reactors.

PubMed

Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

2008-08-01

This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

AQUABOX 50 AND MARABU PACKED BIOLOGICAL REACTOR SYSTEM TECHNOLOGY EVALUATION, STADTWERKE DUESSELDORF AG SITE, DUESSELDORF, GERMANY

EPA Science Inventory

This ITER summarizes the results of an evaluation of the AQUABOX 50 and MARABU Packed Biological Reactor technologies. The evaluation was conducted under a bilateral agreement between the United States (U.S.) Environmental Protection Agency (EPA) Superfund Innovative Technology ...

U.S. Nuclear Cooperation With India: Issues for Congress

DTIC Science & Technology

2010-02-24

Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor Export Deal,” Chennai Business Line Online, July 10, 2009...agreements that covered reactors producing more than 5 MW thermal or special nuclear material connected therewith. 123 United States General Accounting

Analytical design and performance studies of nuclear furnace tests of small nuclear light bulb models

NASA Technical Reports Server (NTRS)

Latham, T. S.; Rodgers, R. J.

1972-01-01

Analytical studies were continued to identify the design and performance characteristics of a small-scale model of a nuclear light bulb unit cell suitable for testing in a nuclear furnace reactor. Emphasis was placed on calculating performance characteristics based on detailed radiant heat transfer analyses, on designing the test assembly for ease of insertion, connection, and withdrawal at the reactor test cell, and on determining instrumentation and test effluent handling requirements. In addition, a review of candidate test reactors for future nuclear light bulb in-reactor tests was conducted.

Direct liquefaction Proof-of-Concept facility. Final technical progress report

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

Comolli, A.G.; Lee, L.K.; Pradhan, V.R.

1995-08-01

This report presents the results of work which included extensive modifications to HRI`s existing 3 ton per day Process Development Unit (PDU) and completion of the first PDU run. The 58-day Run 1 demonstrated scale-up of the Catalytic Two-Stage Liquefaction (CTSL Process) on Illinois No. 6 coal to produce distillate liquid products at a rate of up to 5 barrels per to of moisture-ash-free coal. The Kerr McGee Rose-SR unit from Wilsonville was redesigned and installed next to the US Filter installation to allow a comparison of the two solids removal systems. Also included was a new enclosed reactor tower,more » upgraded computer controls and a data acquisition system, an alternate power supply, a newly refurbished reactor, an in-line hydrotreater, interstage sampling system, coal handling unit, a new ebullating pump, load cells and improved controls and remodeled preheaters. Distillate liquid yields of 5 barrels/ton of moisture ash free coal were achieved. Coal slurry recycle rates were reduced from the 2--2.5 to 1 ratio demonstrated at Wilsonville to as low as 0.9 to 1. Coal feed rates were increased during the test by 50% while maintaining process performance at a marginally higher reactor severity. Sulfur in the coal was reduced from 4 wt% to ca. 0.02 wt% sulfur in the clean distillate fuel product. More than 3,500 gallons of distillate fuels were collected for evaluation and upgrading studies. The ROSE-SR Process was operated for the first time with a pentane solvent in a steady-state model. The energy rejection of the ash concentrate was consistently below prior data, being as low as 12%, allowing improved liquid yields and recovery.« less

Turning the tide of public opinion on nuclear power

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

Cohen, B.L.

1997-04-01

Until the early 1970s, the tide of public opinion in the United States was strongly in favor of nuclear power. New power plants were coming on line frequently, and in 1973-74, there were close to 40 new orders per year for new reactors in the United States. Official government projections estimated 1000 operating reactors by the year 2000. Fuel reprocessing, plutonium recycle, and breeder reactor development were also proceeding smoothly and rapidly. But, in the mid-1970s, the tide suddenly turned against the nuclear industry. How did this come about? In the late 1960s, energetic and idealistic young people who hadmore » never experienced economic insecurity or World Wars came of age. Environmentalism was an attractive outlet for their activity in most of the Western world. In the United States, opposition to the Vietnam War, in which these young people had a personal stake, was even more popular at first, but by the early 1970s, Vietnam was winding down, and they turned also to Environmentalism. Numerous environmental groups started up, aided heavily by the favorable connotation of the very word {open_quotes}environmentalist{close_quotes} in the public mind. Their organizational experience, political savvy, and media connections gained from their antiwar protests were powerful assets. But the groups needed specific targets to attack, and they soon found that nuclear power was well-suited for that purpose. Here was a new technology, coming on at a very rapid pace. To the public, radiation was highly mysterious, and people were well aware that it could be dangerous. And, the word danger had taken on a new meaning. Previous generations were well acquainted with death and were much less averse to risk-taking than the generation of the 1970s.« less

Nuclear reactors built, being built, or planned 1993

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

Not Available

1993-08-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical datamore » that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.« less

Design and Test Plans for a Non-Nuclear Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Mason, Lee; Palac, Donald; Gibson, Marc; Houts, Michael; Warren, John; Werner, James; Poston, David; Qualls, Arthur Lou; Radel, Ross; Harlow, Scott

2012-01-01

A joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) team is developing concepts and technologies for affordable nuclear Fission Power Systems (FPSs) to support future exploration missions. A key deliverable is the Technology Demonstration Unit (TDU). The TDU will assemble the major elements of a notional FPS with a non-nuclear reactor simulator (Rx Sim) and demonstrate system-level performance in thermal vacuum. The Rx Sim includes an electrical resistance heat source and a liquid metal heat transport loop that simulates the reactor thermal interface and expected dynamic response. A power conversion unit (PCU) generates electric power utilizing the liquid metal heat source and rejects waste heat to a heat rejection system (HRS). The HRS includes a pumped water heat removal loop coupled to radiator panels suspended in the thermal-vacuum facility. The basic test plan is to subject the system to realistic operating conditions and gather data to evaluate performance sensitivity, control stability, and response characteristics. Upon completion of the testing, the technology is expected to satisfy the requirements for Technology Readiness Level 6 (System Demonstration in an Operational and Relevant Environment) based on the use of high-fidelity hardware and prototypic software tested under realistic conditions and correlated with analytical predictions.

Design and Test Plans for a Non-Nuclear Fission Power System Technology Demonstration Unit

NASA Astrophysics Data System (ADS)

Mason, L.; Palac, D.; Gibson, M.; Houts, M.; Warren, J.; Werner, J.; Poston, D.; Qualls, L.; Radel, R.; Harlow, S.

A joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) team is developing concepts and technologies for affordable nuclear Fission Power Systems (FPSs) to support future exploration missions. A key deliverable is the Technology Demonstration Unit (TDU). The TDU will assemble the major elements of a notional FPS with a non-nuclear reactor simulator (Rx Sim) and demonstrate system-level performance in thermal vacuum. The Rx Sim includes an electrical resistance heat source and a liquid metal heat transport loop that simulates the reactor thermal interface and expected dynamic response. A power conversion unit (PCU) generates electric power utilizing the liquid metal heat source and rejects waste heat to a heat rejection system (HRS). The HRS includes a pumped water heat removal loop coupled to radiator panels suspended in the thermal-vacuum facility. The basic test plan is to subject the system to realistic operating conditions and gather data to evaluate performance sensitivity, control stability, and response characteristics. Upon completion of the testing, the technology is expected to satisfy the requirements for Technology Readiness Level 6 (System Demonstration in an Operational and Relevant Environment) based on the use of high-fidelity hardware and prototypic software tested under realistic conditions and correlated with analytical predictions.

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

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

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE PAGES

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald; ...

2014-05-01

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

Design of an integrated fuel processor for residential PEMFCs applications

NASA Astrophysics Data System (ADS)

Seo, Yu Taek; Seo, Dong Joo; Jeong, Jin Hyeok; Yoon, Wang Lai

KIER has been developing a novel fuel processing system to provide hydrogen rich gas to residential PEMFCs system. For the effective design of a compact hydrogen production system, each unit process for steam reforming and water gas shift, has a steam generator and internal heat exchangers which are thermally and physically integrated into a single packaged hardware system. The newly designed fuel processor (prototype II) showed a thermal efficiency of 78% as a HHV basis with methane conversion of 89%. The preferential oxidation unit with two staged cascade reactors, reduces, the CO concentration to below 10 ppm without complicated temperature control hardware, which is the prerequisite CO limit for the PEMFC stack. After we achieve the initial performance of the fuel processor, partial load operation was carried out to test the performance and reliability of the fuel processor at various loads. The stability of the fuel processor was also demonstrated for three successive days with a stable composition of product gas and thermal efficiency. The CO concentration remained below 10 ppm during the test period and confirmed the stable performance of the two-stage PrOx reactors.

Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

NASA Astrophysics Data System (ADS)

Honaiser, Eduardo Henrique Rangel

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

Chemistry experience in the primary heat transport circuits of Kraftwerk Union pressurized water reactors

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

Riess, R.

Chosen for this description of the selected Kraftwerk Union (KWU) pressurized water reactor units were Obrigheim (KWO, 345 MW(e)), Stade (KKS, 662 (MW(e)), Borselle (KCB, 477 MW(e)), and Biblis (KWB-A, 1204 MW(e)). The experience at these plants shows that with a special startup procedure and a proper chemical control of the primary heat transport system that influences general corrosion, selective types of corrosion, corrosion product activity transport and resulting contamination, and radiation-induced decomposition, KWU units have no basic problems.

Optimal Refueling Pattern Search for a CANDU Reactor Using a Genetic Algorithm

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

Quang Binh, DO; Gyuhong, ROH; Hangbok, CHOI

2006-07-01

This paper presents the results from the application of genetic algorithms to a refueling optimization of a Canada deuterium uranium (CANDU) reactor. This work aims at making a mathematical model of the refueling optimization problem including the objective function and constraints and developing a method based on genetic algorithms to solve the problem. The model of the optimization problem and the proposed method comply with the key features of the refueling strategy of the CANDU reactor which adopts an on-power refueling operation. In this study, a genetic algorithm combined with an elitism strategy was used to automatically search for themore » refueling patterns. The objective of the optimization was to maximize the discharge burn-up of the refueling bundles, minimize the maximum channel power, or minimize the maximum change in the zone controller unit (ZCU) water levels. A combination of these objectives was also investigated. The constraints include the discharge burn-up, maximum channel power, maximum bundle power, channel power peaking factor and the ZCU water level. A refueling pattern that represents the refueling rate and channels was coded by a one-dimensional binary chromosome, which is a string of binary numbers 0 and 1. A computer program was developed in FORTRAN 90 running on an HP 9000 workstation to conduct the search for the optimal refueling patterns for a CANDU reactor at the equilibrium state. The results showed that it was possible to apply genetic algorithms to automatically search for the refueling channels of the CANDU reactor. The optimal refueling patterns were compared with the solutions obtained from the AUTOREFUEL program and the results were consistent with each other. (authors)« less

Visible-Light-Responsive Photocatalysis: Ag-Doped TiO2 Catalyst Development and Reactor Design Testing

NASA Technical Reports Server (NTRS)

Coutts, Janelle L.; Hintze, Paul E.; Meier, Anne; Shah, Malay G.; Devor, Robert W.; Surma, Jan M.; Maloney, Phillip R.; Bauer, Brint M.; Mazyck, David W.

2016-01-01

In recent years, the alteration of titanium dioxide to become visible-light-responsive (VLR) has been a major focus in the field of photocatalysis. Currently, bare titanium dioxide requires ultraviolet light for activation due to its band gap energy of 3.2 eV. Hg-vapor fluorescent light sources are used in photocatalytic oxidation (PCO) reactors to provide adequate levels of ultraviolet light for catalyst activation; these mercury-containing lamps, however, hinder the use of this PCO technology in a spaceflight environment due to concerns over crew Hg exposure. VLR-TiO2 would allow for use of ambient visible solar radiation or highly efficient visible wavelength LEDs, both of which would make PCO approaches more efficient, flexible, economical, and safe. Over the past three years, Kennedy Space Center has developed a VLR Ag-doped TiO2 catalyst with a band gap of 2.72 eV and promising photocatalytic activity. Catalyst immobilization techniques, including incorporation of the catalyst into a sorbent material, were examined. Extensive modeling of a reactor test bed mimicking air duct work with throughput similar to that seen on the International Space Station was completed to determine optimal reactor design. A bench-scale reactor with the novel catalyst and high-efficiency blue LEDs was challenged with several common volatile organic compounds (VOCs) found in ISS cabin air to evaluate the system's ability to perform high-throughput trace contaminant removal. The ultimate goal for this testing was to determine if the unit would be useful in pre-heat exchanger operations to lessen condensed VOCs in recovered water thus lowering the burden of VOC removal for water purification systems.

78 FR 53482 - Entergy Operations, Inc., River Bend Station, Unit 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-29

... facility consists of a boiling-water reactor located in West Feliciana Parish, Louisiana. 2.0 Request... Containment Leakage Testing for Water- Cooled Power Reactors,'' requires that components which penetrate containment be periodically leak tested at the ``P a, '' defined as the ``calculated peak containment internal...

Fuel element concept for long life high power nuclear reactors

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

75 FR 76498 - Firstenergy Nuclear Operating Company, Davis-Besse Nuclear Power Station; Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-08

... Company, Davis-Besse Nuclear Power Station; Environmental Assessment And Finding of No Significant Impact... operation of the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS), located in Ottawa County, Ohio. In... the reactor coolant pressure boundary of light-water nuclear power reactors provide adequate margins...

Oxygen transfer and uptake, nutrient removal, and energy footprint of parallel full-scale IFAS and activated sludge processes.

PubMed

Rosso, Diego; Lothman, Sarah E; Jeung, Matthew K; Pitt, Paul; Gellner, W James; Stone, Alan L; Howard, Don

2011-11-15

Integrated fixed-film activated sludge (IFAS) processes are becoming more popular for both secondary and sidestream treatment in wastewater facilities. These processes are a combination of biofilm reactors and activated sludge processes, achieved by introducing and retaining biofilm carrier media in activated sludge reactors. A full-scale train of three IFAS reactors equipped with AnoxKaldnes media and coarse-bubble aeration was tested using off-gas analysis. This was operated independently in parallel to an existing full-scale activated sludge process. Both processes achieved the same percent removal of COD and ammonia, despite the double oxygen demand on the IFAS reactors. In order to prevent kinetic limitations associated with DO diffusional gradients through the IFAS biofilm, this systems was operated at an elevated dissolved oxygen concentration, in line with the manufacturer's recommendation. Also, to avoid media coalescence on the reactor surface and promote biofilm contact with the substrate, high mixing requirements are specified. Therefore, the air flux in the IFAS reactors was much higher than that of the parallel activated sludge reactors. However, the standardized oxygen transfer efficiency in process water was almost same for both processes. In theory, when the oxygen transfer efficiency is the same, the air used per unit load removed should be the same. However, due to the high DO and mixing requirements, the IFAS reactors were characterized by elevated air flux and air use per unit load treated. This directly reflected in the relative energy footprint for aeration, which in this case was much higher for the IFAS system than activated sludge. Copyright © 2011 Elsevier Ltd. All rights reserved.

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.

MicroChannel Reactors for ISRU Applications Using Nanofabricated Catalysts

NASA Astrophysics Data System (ADS)

Carranza, Susana; Makel, Darby B.; Vander Wal, Randall L.; Berger, Gordon M.; Pushkarev, Vladimir V.

2006-01-01

With the new direction of NASA to emphasize the exploration of the Moon, Mars and beyond, quick development and demonstration of efficient systems for In-Situ Resources Utilization (ISRU) is more critical and timely than ever before. Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. This paper presents current development of miniaturized chemical processing systems that combine microchannel reactor design with nanofabricated catalysts. Carbon nanotubes (CNT) are used to produce a nanostructure within microchannel reactors, as support for catalysts. By virtue of their nanoscale dimensions, nanotubes geometrically restrict the catalyst particle size that can be supported upon the tube walls. By confining catalyst particles to sizes smaller than the CNT diameter, a more uniform catalyst particle size distribution may be maintained. The high dispersion permitted by the vast surface area of the nanoscale material serves to retain the integrity of the catalyst by reducing sintering or coalescence. Additionally, catalytic efficiency increases with decreasing catalyst particle size (reflecting higher surface area per unit mass) while chemical reactivity frequently is enhanced at the nanoscale. Particularly significant is the catalyst exposure. Rather than being confined within a porous material or deposited upon a 2-d surface, the catalyst is fully exposed to the reactant gases by virtue of the nanofabricated support structure. The combination of microchannel technology with nanofabricated catalysts provides a synergistic effect, enhancing both technologies with the potential to produce much more efficient systems than either technology alone. The development of highly efficient microchannel reactors will be applicable to multiple ISRU programs. By selection of proper nanofabricated catalysts, the microchannel reactors can be designed for the processes that generate the most benefit for each mission, from early demonstration missions to long term settlements.

Evaluation Metrics Applied to Accident Tolerant Fuels

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

Shannon M. Bragg-Sitton; Jon Carmack; Frank Goldner

2014-10-01

The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuelsmore » and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels 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 U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being readied for insertion in fiscal year 2015. This paper provides a brief summary of the proposed evaluation process that would be used to evaluate and prioritize the candidate accident tolerant fuel concepts currently under development.« less

Compact, Lightweight Adsorber and Sabatier Reactor for CO2 Capture and Reduction for Consumable and Propellant Production

NASA Technical Reports Server (NTRS)

Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Busby, Stacy A.; Abney, Morgan B.; Perry, Jay L.; Knox, James C.

2012-01-01

The utilization of CO2 to produce (or recycle) life support consumables, such as O2 and H2O, and to generate propellant fuels is an important aspect of NASA's concept for future, long duration planetary exploration. One potential approach is to capture and use CO2 from the Martian atmosphere to generate the consumables and propellant fuels. Precision Combustion, Inc. (PCI), with support from NASA, continues to develop its regenerable adsorber technology for capturing CO2 from gaseous atmospheres (for cabin atmosphere revitalization and in-situ resource utilization applications) and its Sabatier reactor for converting CO2 to methane and water. Both technologies are based on PCI's Microlith(R) substrates and have been demonstrated to reduce size, weight, and power consumption during CO2 capture and methanation process. For adsorber applications, the Microlith substrates offer a unique resistive heating capability that shows potential for short regeneration time and reduced power requirements compared to conventional systems. For the Sabatier applications, the combination of the Microlith substrates and durable catalyst coating permits efficient CO2 methanation that favors high reactant conversion, high selectivity, and durability. Results from performance testing at various operating conditions will be presented. An effort to optimize the Sabatier reactor and to develop a bench-top Sabatier Development Unit (SDU) will be discussed.

Development Status for a Combined Solid Oxide Co-Electrolyzer and Carbon Formation Reactor System for Oxygen Regeneration

NASA Technical Reports Server (NTRS)

Green, Robert D.; Matter, Paul H.; Holt, Chris; Beachy, Michael; Gaydos, James; Farmer, Serene C.; Setlock, John

2016-01-01

A critical component in spacecraft life support loop closure is the removal of carbon dioxide (CO2, produced by the crew) from the cabin atmosphere and chemical reduction of this CO2 to recover the oxygen. In 2015, we initiated development of an oxygen recovery system for life support applications consisting of a solid oxide co-electrolyzer (SOCE) and a carbon formation reactor (CFR). The SOCE electrolyzes a combined stream of carbon dioxide (CO2) and water (H2O) gas mixtures to produce synthesis gas (e.g., CO and H2 gas) and pure dry oxygen as separate products. This SOCE is being developed from a NASA GRC solid oxide fuel cell and stack design originally developed for aeronautics long-duration power applications. The CFR, being developed by pHMatter LLC, takes the CO and H2 output from the SOCE, and converts it primarily to solid carbon (C(s)) and H2O and CO2. Although the solid carbon accumulates in the CFR, the innovative design allows easy removal of the carbon product, requiring minimal crew member (CM) time and low resupply mass (1.0 kg/year/CM) for replacement of the solid carbon catalyst, a significant improvement over previous Bosch reactor approaches. In this work, we will provide a status of our Phase I efforts in the development and testing of both the SOCE and CFR prototype units, along with an initial assessment of the combined SOCE-CFR system, including a mass and power projections, along with an estimate of the oxygen recovery rate.

ATOMIC POWER PLANT

DOEpatents

Daniels, F.

1957-11-01

This patent relates to neutronic reactor power plants and discloses a design of a reactor utilizing a mixture of discrete units of a fissionable material, such as uranium carbide, a neutron moderator material, such as graphite, to carry out the chain reaction. A liquid metal, such as bismuth, is used as the coolant and is placed in the reactor chamber with the fissionable and moderator material so that it is boiled by the heat of the reaction, the boiling liquid and vapors passing up through the interstices between the discrete units. The vapor and flue gases coming off the top of the chamber are passed through heat exchangers, to produce steam, for example, and thence through condensers, the condensed coolant being returned to the chamber by gravity and the non- condensible gases being carried off through a stack at the top of the structure.

Development and Application of Laser Peening System for PWR Power Plants

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

Masaki Yoda; Itaru Chida; Satoshi Okada

2006-07-01

Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion crackingmore » (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described. (authors)« less

Nuclear pursuits

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

Not Available

1993-05-01

This table lists quantities of warheads (in stockpile, peak number per year, total number built, number of known test explosions), weapon development milestones (developers of the atomic bomb and hydrogen bomb, date of first operational ICBM, first nuclear-powered naval SSN in service, first MIRVed missile deployed), and testing milestones (first fission test, type of boosted fission weapon, multistage thermonuclear test, number of months from fission bomb to multistage thermonuclear bomb, etc.), and nuclear infrastructure (assembly plants, plutonium production reactors, uranium enrichment plants, etc.). Countries included in the tally are the United States, Soviet Union, Britain, France, and China.

LASER APPLICATIONS AND OTHER TOPICS IN QUANTUM ELECTRONICS: On the possibility of development of a photochemical unit based on an NH3 laser with an intracavity reactor

NASA Astrophysics Data System (ADS)

Apatin, V. M.; Belokurov, A. N.; Makarov, Grigorii N.; Mendoza, P.; Petin, A. N.; Pigul'skii, S. V.; Rios, I.; Ryabov, Evgenii A.

2006-03-01

The possibility of developing a photochemical setup on the basis of an optically pumped ammonia laser with an intracavity photoreactor is proved. The obtained values of the cavity 'implication' factor γ are comparable with those of intracavity systems based on a CO2 laser. The conditions for achieving the maximum energy in the focusing cavity are determined and the ways to control the shape of its caustic are indicated.

Evaluation of biogas production potential by dry anaerobic digestion of switchgrass--animal manure mixtures.

PubMed

Ahn, H K; Smith, M C; Kondrad, S L; White, J W

2010-02-01

Anaerobic digestion is a biological method used to convert organic wastes into a stable product for land application with reduced environmental impacts. The biogas produced can be used as an alternative renewable energy source. Dry anaerobic digestion [>15% total solid (TS)] has an advantage over wet digestion (<10% TS) because it allows for the use of a smaller volume of reactor and because it reduces wastewater production. In addition, it produces a fertilizer that is easier to transport. Performance of anaerobic digestion of animal manure-switchgrass mixture was evaluated under dry (15% TS) and thermophilic conditions (55 degrees C). Three different mixtures of animal manure (swine, poultry, and dairy) and switchgrass were digested using batch-operated 1-L reactors. The swine manure test units showed 52.9% volatile solids (VS) removal during the 62-day trial, while dairy and poultry manure test units showed 9.3% and 20.2%, respectively. Over the 62 day digestion, the swine manure test units yielded the highest amount of methane 0.337 L CH4/g VS, while the dairy and poultry manure test units showed very poor methane yield 0.028 L CH4/g VS and 0.002 L CH4/g VS, respectively. Although dairy and poultry manure performed poorly, they may still have high potential as biomass for dry anaerobic digestion if appropriate designs are developed to prevent significant volatile fatty acid (VFA) accumulation and pH drop.

The in-depth safety assessment (ISA) pilot projects in Ukraine.

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

Kot, C. A.

1998-02-10

Ukraine operates pressurized water reactors of the Soviet-designed type, VVER. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs). After approval of the SARS by the Ukrainian Nuclear Regulatory Authority, the plants will be granted longer-term operating licenses. In September 1995, the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine issued a new contents requirement for the safety analysis reports of VVERs in Ukraine. It contains requirements in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. Themore » DBA requirements are an expanded version of the older SAR requirements. The last two requirements, on PRA and BDBA, are new. The US Department of Energy (USDOE), through the International Nuclear Safety Program (INSP), has initiated an assistance and technology transfer program to Ukraine to assist their nuclear power stations in developing a Western-type technical basis for the new SARS. USDOE sponsored In-Depth Safety Assessments (ISAs) have been initiated at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1. USDOE/INSP have structured the ISA program in such a way as to provide maximum assistance and technology transfer to Ukraine while encouraging and supporting the Ukrainian plants to take the responsibility and initiative and to perform the required assessments.« less

Nuclear Safeguards and the International Atomic Energy Agency

DTIC Science & Technology

1995-01-01

designed to use HEU, the tite safeguards agreement already in place for cen- RERTR (Reduced Enrichment for Research and trifuge facilities (which allows only...notice in- types. Many such reactors have been converted. spections, such as provided for under the Hexa- (See discussion below on the RERTR program...the Schumer amendment to the United States was developing suitable alternate 19Some believe that the suspension of the RERTR program may have been a

Central Asia: Regional Developments and Implications for U.S. Interests

DTIC Science & Technology

2008-07-10

United Press International, December 13, 2005. Organization (SCO; see above, Regional Tensions) that stated that “as large - scale military operations...reserves, and Kazakhstan and Uzbekistan have been among the world’s top producers of low enriched uranium. Kazakhstan had a fast breeder reactor at Aktau...Asia, Afghanistan, and eventually Pakistan and India.72 All the states of the region possess large - scale resources that could contribute to the region

Evaluation of decentralized treatment of sewage employing Upflow Septic Tank/Baffled Reactor (USBR) in developing countries.

PubMed

Sabry, Tarek

2010-02-15

A new concept for a low-cost modified septic tank, named Upflow Septic Tank/Baffled Reactor (USBR), was constructed and tested in a small village in Egypt. During almost one year of continuous operation and monitoring, this system was found to have very satisfactory removal results, where the average results of COD, BOD, and TSS removal efficiencies were 84%, 81%, and 89%, respectively, and the results of the experiment proved that the second compartment (Anaerobic Baffled Reactor) was the main treatment unit in removing the pollutants during the start-up period and at the very early steady-state stage. However, after this period and during the steady-state operation conditions, the second compartment served as a polishing step. Also, it was observed that the USBR system was not affected by the imposed shock loads at the peak flow and organic periods. The results showed that the system is slightly influenced by the drop in the temperature. Decreasing in BOD and COD removal by factor of 9% was observed, when temperature decreases from the average of 35 degrees C in summer time (for the first 127 days) to the average of 22 degrees C in winter time (between day 252 and day 280). Whereas, the TSS removals were not affected by the drop in temperature. The results of the sewage flow variations during one year of operation were compared with Goodrich Formula to see the applicability of this equation in rural developing countries. MAIN FINDING OF THE WORK: The Upflow Septic Tank/Baffled Reactor system could become a promising alternative to the conventional treatment plants in rural developing countries.

75 FR 76051 - Northern States Power Company-Minnesota, Prairie Island Nuclear Generating Plant, Units 1 and 2...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-07

..., 2010 (Agencywide Documents Access and Management System Accession Nos. ML093280883 and ML101480083... systems for light-water nuclear power reactors,'' and appendix K to 10 CFR part 50, ``ECCS Evaluation... core cooling system (ECCS) for reactors fueled with zircaloy or ZIRLO\\TM\\ cladding. In addition...

75 FR 16517 - Dominion Nuclear Connecticut, Inc.; Millstone Power Station, Unit Nos 1, 2, and 3; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-01

.... Borchardt, NRC, to M. S. Fertel, Nuclear Energy Institute, ADAMS Accession No. ML091410309). The licensee's... effect. The facility consists of one boiling water reactor and two pressurized water reactors located in... public. The supplemental January 12, 2010, letter contains, as an attachment, an environmental assessment...

40 CFR 63.1579 - What definitions apply to this subpart?

Code of Federal Regulations, 2014 CFR

2014-07-01

... regeneration of catalyst in situ in any one of several reactors (e.g., 4 or 5 separate reactors) that can be..., wet injection, or caustic injection control device that treats (in-situ) the catalytic reforming unit...) at specified intervals or at the owner's or operator's convenience for in situ catalyst regeneration...

40 CFR 63.1579 - What definitions apply to this subpart?

Code of Federal Regulations, 2013 CFR

2013-07-01

... regeneration of catalyst in situ in any one of several reactors (e.g., 4 or 5 separate reactors) that can be..., wet injection, or caustic injection control device that treats (in-situ) the catalytic reforming unit...) at specified intervals or at the owner's or operator's convenience for in situ catalyst regeneration...

40 CFR 63.1579 - What definitions apply to this subpart?

Code of Federal Regulations, 2012 CFR

2012-07-01

... regeneration of catalyst in situ in any one of several reactors (e.g., 4 or 5 separate reactors) that can be..., wet injection, or caustic injection control device that treats (in-situ) the catalytic reforming unit...) at specified intervals or at the owner's or operator's convenience for in situ catalyst regeneration...

40 CFR 141.720 - Inactivation toolbox components.

Code of Federal Regulations, 2014 CFR

2014-07-01

... treatment unit process with a measurable disinfectant residual level and a liquid volume. Under this... through reactor validation testing, as described in paragraph (d)(2) of this section. The UV dose values....0 12 11 143 (vii) 3.5 15 15 163 (viii) 4.0 22 22 186 (2) Reactor validation testing. Systems must...

Pyrolysis of cassava rhizome in a counter-rotating twin screw reactor unit.

PubMed

Sirijanusorn, Somsak; Sriprateep, Keartisak; Pattiya, Adisak

2013-07-01

A counter-rotating twin screw reactor unit was investigated for its behaviour in the pyrolysis of cassava rhizome biomass. Several parameters such as pyrolysis temperature in the range of 500-700°C, biomass particle size of <0.6mm, the use of sand as heat transfer medium, nitrogen flow rate of 4-10 L/min and nitrogen pressure of 1-3 bar were thoroughly examined. It was found that the pyrolysis temperature of 550°C could maximise the bio-oil yield (50 wt.%). The other optimum parameters for maximising the bio-oil yield were the biomass particle size of 0.250-0.425 mm, the nitrogen flow rate of 4 L/min and the nitrogen pressure of 2 bar. The use of the heat transfer medium could increase the bio-oil yield to a certain extent. Moreover, the water content of bio-oil produced with the counter-rotating twin screw reactor was relatively low, whereas the solids content was relatively high, compared to some other reactor configurations. Copyright © 2013 Elsevier Ltd. All rights reserved.

Safeguards Challenges for Pebble-Bed Reactors (PBRs):Peoples Republic of China (PRC)

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

Forsberg, Charles W.; Moses, David Lewis

2009-11-01

The Peoples Republic of China (PRC) is operating the HTR-10 pebble-bed reactor (PBR) and is in the process of building a prototype PBR plant with two modular reactors (250-MW(t) per reactor) feeding steam to a single turbine-generator. It is likely to be the first modular hightemperature reactor to be ready for commercial deployment in the world because it is a highpriority project for the PRC. The plant design features multiple modular reactors feeding steam to a single turbine generator where the number of modules determines the plant output. The design and commercialization strategy are based on PRC strengths: (1) amore » rapidly growing electric market that will support low-cost mass production of modular reactor units and (2) a balance of plant system based on economics of scale that uses the same mass-produced turbine-generator systems used in PRC coal plants. If successful, in addition to supplying the PRC market, this strategy could enable China to be the leading exporter of nuclear reactors to developing countries. The modular characteristics of the reactor match much of the need elsewhere in the world. PBRs have major safety advantages and a radically different fuel. The fuel, not the plant systems, is the primary safety system to prevent and mitigate the release of radionuclides under accident conditions. The fuel consists of small (6-cm) pebbles (spheres) containing coatedparticle fuel in a graphitized carbon matrix. The fuel loading per pebble is small (~9 grams of low-enriched uranium) and hundreds of thousands of pebbles are required to fuel a nuclear plant. The uranium concentration in the fuel is an order of magnitude less than in traditional nuclear fuels. These characteristics make the fuel significantly less attractive for illicit use (weapons production or dirty bomb); but, its unusual physical form may require changes in the tools used for safeguards. This report describes PBRs, what is different, and the safeguards challenges. A series of safeguards recommendations are made based on the assumption that the reactor is successfully commercialized and is widely deployed.« less

Physical-chemical treatment of wastes: a way to close turnover of elements in LSS

NASA Astrophysics Data System (ADS)

Kudenko, Yu A.; Gribovskaya, I. V.; Zolotukhin, I. G.

2000-05-01

"Man-plants-physical-chemical unit" system designed for space stations or terrestrial ecohabitats to close steady-state mineral, water and gas exchange is proposed. The physical-chemical unit is to mineralize all inedible plant wastes and physiological human wastes (feces, urine, gray water) by electromagnetically activated hydrogen peroxide in an oxidation reactor. The final product is a mineralized solution containing all elements balanced for plants' requirements. The solution has been successfully used in experiments to grow wheat, beans and radish. The solution was reusable: the evaporated moisture was replenished by the phytotron condensate. Sodium salination of plants was precluded by evaporating reactor-mineralized urine to sodium saturation concentration to crystallize out NaCl which can be used as food for the crew. The remaining mineralized product was brought back for nutrition of plants. The gas composition of the reactor comprises O 2, N 2, CO 2, NH 3, H 2. At the reactor's output hydrogen and oxygen were catalyzed into water, NH 3 was converted in a water trap into NH 4 and used for nutrition of plants. A special accessory at the reactor's output may produce hydrogen peroxide from intrasystem water and gas which makes possible to close gas loops between LSS components.

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

Gehin, Jess C; Godfrey, Andrew T; Evans, Thomas M

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications, including a core simulation capability called VERA-CS. A key milestone for this endeavor is to validate VERA against measurements from operating nuclear power reactors. The first step in validation against plant data is to determine the ability of VERA to accurately simulate the initial startup physics tests for Watts Bar Nuclear Power Station, Unit 1 (WBN1) cycle 1. VERA-CS calculations were performed with the Insilico code developed at ORNL using cross sectionmore » processing from the SCALE system and the transport capabilities within the Denovo transport code using the SPN method. The calculations were performed with ENDF/B-VII.0 cross sections in 252 groups (collapsed to 23 groups for the 3D transport solution). The key results of the comparison of calculations with measurements include initial criticality, control rod worth critical configurations, control rod worth, differential boron worth, and isothermal temperature reactivity coefficient (ITC). The VERA results for these parameters show good agreement with measurements, with the exception of the ITC, which requires additional investigation. Results are also compared to those obtained with Monte Carlo methods and a current industry core simulator.« less

Generation-IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

McFarlane, Harold

2008-05-01

Nuclear power technology has evolved through roughly three generations of system designs: a first generation of prototypes and first-of-a-kind units implemented during the period 1950 to 1970; a second generation of industrial power plants built from 1970 to the turn of the century, most of which are still in operation today; and a third generation of evolutionary advanced reactors which began being built by the turn of the 20^th century, usually called Generation III or III+, which incorporate technical lessons learned through more than 12,000 reactor-years of operation. The Generation IV International Forum (GIF) is a cooperative international endeavor to develop advanced nuclear energy systems in response to the social, environmental and economic requirements of the 21^st century. Six Generation IV systems under development by GIF promise to enhance the future contribution and benefits of nuclear energy. All Generation IV systems aim at performance improvement, new applications of nuclear energy, and/or more sustainable approaches to the management of nuclear materials. High-temperature systems offer the possibility of efficient process heat applications and eventually hydrogen production. Enhanced sustainability is achieved primarily through adoption of a closed fuel cycle with reprocessing and recycling of plutonium, uranium and minor actinides using fast reactors. This approach provides significant reduction in waste generation and uranium resource requirements.

LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

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

McElroy, W.N.

1985-08-01

This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old)more » was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.« less

Neutron radiation characteristics of the IVth generation reactor spent fuel

NASA Astrophysics Data System (ADS)

Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

2018-03-01

Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

Development of an NDA system for high-level waste from the Chernobyl new safe confinement construction site

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

Lee, Sang-yoon; Browne, Michael C; Rael, Carlos D

2010-01-01

In early 2009, preliminary excavation work has begun in preparation for the construction of the New Safe Confinement (NSC) at the Chernobyl Nuclear Power Plant (ChNPP) in Ukraine. The NSC is the structure that will replace the present containment structure and will confine the radioactive remains of the ChNPP Unit-4 reactor for the next 100 years. It is expected that special nuclear material (SNM) that was ejected from the Unit-4 reactor during the accident in 1986 could be uncovered and would therefore need to be safeguarded. ChNPP requested the assistance of the United States Department of Energy/National Nuclear Security Administrationmore » (NNSA) with developing a new non-destructive assay (NDA) system that is capable of assaying radioactive debris stored in 55-gallon drums. The design of the system has to be tailored to the unique circumstances and work processes at the NSC construction site and the ChNPP. This paper describes the Chernobyl Drum Assay System (CDAS), the solution devised by Los Alamos National Laboratory, Sonalysts Inc., and the ChNPP, under NNSA's International Safeguards and Engagement Program (INSEP). The neutron counter measures the spontaneous fission neutrons from the {sup 238}U, {sup 240}Pu, {sup 244}Cm in a waste drum and estimates the mass contents of the SNMs in the drum by using of isotopic compositions determined by fuel burnup. The preliminary evaluation on overall measurement uncertainty shows that the system meets design performance requirements imposed by the facility.« less

76 FR 39134 - ZIONSOLUTIONS, LLC; Zion Nuclear Power Station, Units 1 and 2 Exemption From Recordkeeping...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-05

...; Zion Nuclear Power Station, Units 1 and 2 Exemption From Recordkeeping Requirements 1.0 Background Zion Nuclear Power Station (ZNPS or Zion), Unit 1, is a Westinghouse 3250 MWt Pressurized Water Reactor which... previously applicable to the nuclear power units and associated systems, structures, and components (SSC) are...

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.

VERA Core Simulator methodology for pressurized water reactor cycle depletion

DOE PAGES

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

2017-01-12

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

Investigation of the effects of hydrogenotrophic denitrification and anammox on the improvement of the quality of the drinking water supply system.

PubMed

Khanitchaidecha, W; Koshy, P; Kamei, T; Shakya, M; Kazama, F

2013-01-01

A drinking water supply system operates at Chyasal (in the Kathmandu Valley, Nepal) for purifying the groundwater that has high levels of ammonium nitrogen (NH4-N). However, high NO3-N concentrations were seen in the water after treatment. To further improve the quality of the drinking water, two types of attached growth reactors were developed for the purification system: (i) a hydrogenotrophic denitrification (HD reactor) and (ii) a concurrent reactor with anammox and hydrogenotrophic denitrification (AnHD reactor). For the HD reactor fed by water containing NO3-N, the denitrification efficiency was high (95-98%) for all NO3-N feed rates (20-40 mg/L). The nitrite-nitrogen (NO2-N) and nitrate-nitrogen (NO3-N) concentrations in the effluent were ∼0.5 mg/L. On the other hand, the AnHD reactor fed with water containing NH4-N and NO2-N was operated under varying flow rates of H2(30-70 mL/min) and intermittent supply periods (1-2 h). The efficiency of the anammox process was found to increase with decreasing H2flow rates or with increasing intermittency of the H2supply, while the efficiency of denitrification decreased under these conditions. For the optimal condition of 1.5 h intermittent H2supply, the anammox and denitrification efficiencies of the AnHD reactor reached 80% and 42%, respectively, while the concentrations of both NH4-N and NO2-N in the effluent were <1.0 mg/L, and no NO3-N was detected. From the experimental results, it is clear that both HD and AnHD reactors can function as efficient and critical units of the water purification system.

Nuclear reactors built, being built, or planned 1996

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

NONE

1997-08-01

This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables ofmore » the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.« less

Breaking America’s Dependence on Foreign…Molybdenum

PubMed Central

Einstein, Andrew J.

2009-01-01

Brief Unstructured Abstract Approximately 9 million nuclear cardiology studies performed each year in the United States employ technetium-99m, which is produced from the decay of molybdenum-99. The fragility of the worldwide technetium-99m supply chain has been underscored by current shortages caused by an unplanned shutdown of Europe’s largest reactor. The majority of the United States’ supply derives from a reactor in Canada nearing the end of its lifespan, whose planned replacements have been recently cancelled. In this article, the clinical importance of technetium-99m and our tenuous dependence on foreign supply of Molybdenum is addressed. PMID:19356583

Licensed operating reactors: Status summary report, data as of December 31, 1995. Volume 20

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

NONE

1996-06-01

The US Nuclear Regulatory Commission`s monthly summary of licensed nuclear power reactor data is based primarily on the operating data report submitted by licensees for each unit. This report is divided into two sections: the first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 availability factors, capacity factors, and forced outage rates are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensees and notes to the use of weighted averages and starting dates other than commercial operation are provided.

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.; El-Genk, Mohamed S.

2013-09-01

This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

PWR design for low doses in the United Kingdom: The present and the future

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

Zodiates, A.M.; Willcock, A.

1995-03-01

The Pressurizer Water Reactor (PWR) design chosen for adoption by Nuclear Electric plc was based on the Westinghouse Standard Nuclear Unit Power Plant System (SNUPPS). This design was developed to meet the United Kingdom (UK) requirements and those improvements are embodied in the Sizewell B plant. Nuclear Electric plc is now looking to the design of the future PWRs to be built in the UK. These PWRs will be based as replicas of the Sizewell B design, but attention will be given to reducing operator doses further. This paper details the approach in operator protection improvements incorporated at Sizewall B,more » presents the estimated annual collective dose, and identifies the approach being adopted to reduce further operator doses in future plants.« less

Of Ashes and Atoms

NASA Technical Reports Server (NTRS)

2005-01-01

This feature length DVD documentary, reviews the history of the Plum Brook Nuclear Reactor from the initial settlers of the area, through its use as a munitions facility during the second World War to the development of the nuclear facility and its use as one of the first nuclear test reactors built in the United States, and the only one built by NASA. It concludes with the beginning of the decommissioning of the facility. There is a brief review of the reactor design, and its workings. Through discussions with the NASA engineers and operators of the facility, the film reviews the work done to advance the knowledge of the effects of radiation, the properties of radiated materials, and the work to advance the state of the art in nuclear propulsion. The film shows footage of public tours, and shows actual footage of the facility in operation, and after its shutdown in 1973. The DVD was narrated by Kate Mulgrew, who leads the viewer through the history of the facility to its eventual ongoing decommissioning, and return to the state of pastoral uses.

EDF experience with {open_quotes}hot spot{close_quotes} management

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

Guio, J.M. de

1995-03-01

During the past few years, {open_quotes}hot spots{close_quotes} due to the presence of particles of metal activated during their migration through the reactor core, have been detected at several French pressurized water reactor (PWR) units. These {open_quotes}hot spots,{close_quotes} which generate very high dose rates (from about 10 Gy/h to 200 G/h) are a significant factor in increase occupational exposures during outrates. Of particular concern are the difficult cases which prolong outage duration and increase the volume of radiological waste. Confronted with this situation, Electricite de France (EDF) has set up a national research group, as part of its ALARA program, tomore » establish procedures and techniques to avoid, detect, and eliminate of hot spots. In particular, specific processes have been developed to eliminate these hot spots which are most costly in terms of occupational exposure due to the need for reactor maintenance. This paper sets out the general approach adopted at EDF so far to cope with the problem of hot spots, illustrated by experience at Blayais 3 and 4.« less

Influence of liquid medium on the activity of a low-alpha Fischer-Tropsch catalyst

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

Gormley, R.J.; Zarochak, M.F.; Deffenbaugh, P.W.

1995-12-31

The purpose of this research was to measure activity, selectivity, and the maintenance of these properties in slurry autoclave experiments with a Fischer-Tropsch (FT) catalyst that was used in the {open_quotes}FT II{close_quotes} bubble-column test, conducted at the Alternative Fuels Development Unit (AFDU) at LaPorte, Texas during May 1994. The catalyst contained iron, copper, and potassium and was formulated to produce mainly hydrocarbons in the gasoline range with lesser production of diesel-range products and wax. The probability of chain growth was thus deliberately kept low. Principal goals of the autoclave work have been to find the true activity of this catalystmore » in a stirred tank reactor, unhindered by heat or mass transfer effects, and to obtain a steady conversion and selectivity over the approximately 15 days of each test. Slurry autoclave testing of the catalyst in heavier waxes also allows insight into operation of larger slurry bubble column reactors. The stability of reactor operation in these experiments, particularly at loadings exceeding 20 weight %, suggests the likely stability of operations on a larger scale.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE PAGES

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; ...

2016-10-01

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

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

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

Modular bioreactor for the remediation of liquid streams and methods for using the same

DOEpatents

Noah, Karl S.; Sayer, Raymond L.; Thompson, David N.

1998-01-01

The present invention is directed to a bioreactor system for the remediation of contaminated liquid streams. The bioreactor system is composed of at least one and often a series of sub-units referred to as bioreactor modules. The modular nature of the system allows bioreactor systems be subdivided into smaller units and transported to waste sites where they are combined to form bioreactor systems of any size. The bioreactor modules further comprises reactor fill materials in the bioreactor module that remove the contaminants from the contaminated stream. To ensure that the stream thoroughly contacts the reactor fill materials, each bioreactor module comprises means for directing the flow of the stream in a vertical direction and means for directing the flow of the stream in a horizontal direction. In a preferred embodiment, the reactor fill comprises a sulfate reducing bacteria which is particularly useful for precipitating metals from acid mine streams.

Modular bioreactor for the remediation of liquid streams and methods for using the same

DOEpatents

Noah, K.S.; Sayer, R.L.; Thompson, D.N.

1998-06-30

The present invention is directed to a bioreactor system for the remediation of contaminated liquid streams. The bioreactor system is composed of at least one and often a series of sub-units referred to as bioreactor modules. The modular nature of the system allows bioreactor systems be subdivided into smaller units and transported to waste sites where they are combined to form bioreactor systems of any size. The bioreactor modules further comprises reactor fill materials in the bioreactor module that remove the contaminants from the contaminated stream. To ensure that the stream thoroughly contacts the reactor fill materials, each bioreactor module comprises means for directing the flow of the stream in a vertical direction and means for directing the flow of the stream in a horizontal direction. In a preferred embodiment, the reactor fill comprises a sulfate reducing bacteria which is particularly useful for precipitating metals from acid mine streams. 6 figs.

Design of a laboratory scale fluidized bed reactor

NASA Astrophysics Data System (ADS)

Wikström, E.; Andersson, P.; Marklund, S.

1998-04-01

The aim of this project was to construct a laboratory scale fluidized bed reactor that simulates the behavior of full scale municipal solid waste combustors. The design of this reactor is thoroughly described. The size of the laboratory scale fluidized bed reactor is 5 kW, which corresponds to a fuel-feeding rate of approximately 1 kg/h. The reactor system consists of four parts: a bed section, a freeboard section, a convector (postcombustion zone), and an air pollution control (APC) device system. The inside diameter of the reactor is 100 mm at the bed section and it widens to 200 mm in diameter in the freeboard section; the total height of the reactor is 1760 mm. The convector part consists of five identical sections; each section is 2700 mm long and has an inside diameter of 44.3 mm. The reactor is flexible regarding the placement and number of sampling ports. At the beginning of the first convector unit and at the end of each unit there are sampling ports for organic micropollutants (OMP). This makes it possible to study the composition of the flue gases at various residence times. Sampling ports for inorganic compounds and particulate matter are also placed in the convector section. All operating parameters, reactor temperatures, concentrations of CO, CO2, O2, SO2, NO, and NO2 are continuously measured and stored at selected intervals for further evaluation. These unique features enable full control over the fuel feed, air flows, and air distribution as well as over the temperature profile. Elaborate details are provided regarding the configuration of the fuel-feeding systems, the fluidized bed, the convector section, and the APC device. This laboratory reactor enables detailed studies of the formation mechanisms of OMP, such as polychlorinated dibenzo-p-dioxins (PCDDs), polychlorinated dibenzofurans (PCDFs), poly-chlorinated biphenyls (PCBs), and polychlorinated benzenes (PCBzs). With this system formation mechanisms of OMP occurring in both the combustion and postcombustion zones can be studied. Other advantages are memory effect minimization and the reduction of experimental costs compared to full scale combustors. Comparison of the combustion parameters and emission data from this 5 kW laboratory scale reactor with full scale combustors shows good agreement regarding emission levels and PCDD/PCDF congener patterns. This indicates that the important formation and degradation reactions of OMP in the reactor are the same formation mechanisms as in full scale combustors.

Hydrogen Purification and Recycling for an Integrated Oxygen Recovery System Architecture

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Greenwood, Zachary; Wall, Terry; Miller, Lee; Wheeler, Ray

2016-01-01

The United States Atmosphere Revitalization life support system on the International Space Station (ISS) performs several services for the crew including oxygen generation, trace contaminant control, carbon dioxide (CO2) removal, and oxygen recovery. Oxygen recovery is performed using a Sabatier reactor developed by Hamilton Sundstrand, wherein CO2 is reduced with hydrogen in a catalytic reactor to produce methane and water. The water product is purified in the Water Purification Assembly and recycled to the Oxygen Generation Assembly (OGA) to provide O2 to the crew. This architecture results in a theoretical maximum oxygen recovery from CO2 of approximately 54% due to the loss of reactant hydrogen in Sabatier-produced methane that is currently vented outside of ISS. Plasma Methane Pyrolysis technology (PPA), developed by Umpqua Research Company, provides the capability to further close the Atmosphere Revitalization oxygen loop by recovering hydrogen from Sabatier-produced methane. A key aspect of this technology approach is to purify the hydrogen from the PPA product stream which includes acetylene, unreacted methane and byproduct water and carbon monoxide. In 2015, four sub-scale hydrogen separation systems were delivered to NASA for evaluation. These included two electrolysis single-cell hydrogen purification cell stacks developed by Sustainable Innovations, LLC, a sorbent-based hydrogen purification unit using microwave power for sorbent regeneration developed by Umpqua Research Company, and a LaNi4.6Sn0.4 metal hydride produced by Hydrogen Consultants, Inc. Here we report the results of these evaluations, discuss potential architecture options, and propose future work.

Hydrogen Purification and Recycling for an Integrated Oxygen Recovery System Architecture

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Greenwood, Zachary; Wall, Terry; Nur, Mononita; Wheeler, Richard R., Jr.; Preston, Joshua; Molter, Trent

2016-01-01

The United States Atmosphere Revitalization life support system on the International Space Station (ISS) performs several services for the crew including oxygen generation, trace contaminant control, carbon dioxide (CO2) removal, and oxygen recovery. Oxygen recovery is performed using a Sabatier reactor developed by Hamilton Sundstrand, wherein CO2 is reduced with hydrogen in a catalytic reactor to produce methane and water. The water product is purified in the Water Purification Assembly and recycled to the Oxygen Generation Assembly (OGA) to provide O2 to the crew. This architecture results in a theoretical maximum oxygen recovery from CO2 of approx.54% due to the loss of reactant hydrogen in Sabatier-produced methane that is currently vented outside of ISS. Plasma Pyrolysis Assembly (PPA) technology, developed by Umpqua Research Company, provides the capability to further close the Atmosphere Revitalization oxygen loop by recovering hydrogen from Sabatier-produced methane. A key aspect of this technology approach is the need to purify the hydrogen from the PPA product stream which includes acetylene, unreacted methane and byproduct water and carbon monoxide. In 2015, four sub-scale hydrogen separation systems were delivered to NASA for evaluation. These included two electrolysis single-cell hydrogen purification cell stacks developed by Sustainable Innovations, LLC, a sorbent-based hydrogen purification unit using microwave power for sorbent regeneration developed by Umpqua Research Company, and a LaNi4.6Sn0.4 metal hydride produced by Hydrogen Consultants, Inc. Here we report the results of these evaluations to-date, discuss potential architecture options, and propose future work.

Research and proposal on selective catalytic reduction reactor optimization for industrial boiler.

PubMed

Yang, Yiming; Li, Jian; He, Hong

2017-08-24

The advanced computational fluid dynamics (CFD) software STAR-CCM+ was used to simulate a denitrification (De-NOx) project for a boiler in this paper, and the simulation result was verified based on a physical model. Two selective catalytic reduction (SCR) reactors were developed: reactor 1 was optimized and reactor 2 was developed based on reactor 1. Various indicators, including gas flow field, ammonia concentration distribution, temperature distribution, gas incident angle, and system pressure drop were analyzed. The analysis indicated that reactor 2 was of outstanding performance and could simplify developing greatly. Ammonia injection grid (AIG), the core component of the reactor, was studied; three AIGs were developed and their performances were compared and analyzed. The result indicated that AIG 3 was of the best performance. The technical indicators were proposed for SCR reactor based on the study. Flow filed distribution, gas incident angle, and temperature distribution are subjected to SCR reactor shape to a great extent, and reactor 2 proposed in this paper was of outstanding performance; ammonia concentration distribution is subjected to ammonia injection grid (AIG) shape, and AIG 3 could meet the technical indicator of ammonia concentration without mounting ammonia mixer. The developments above on the reactor and the AIG are both of great application value and social efficiency.

77 FR 40092 - License Amendment To Increase the Maximum Reactor Power Level, Florida Power & Light Company, St...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-06

... Increase the Maximum Reactor Power Level, Florida Power & Light Company, St. Lucie, Units 1 and 2 AGENCY... amendment for Renewed Facility Operating License Nos. DPR-67 and NPF-16, issued to Florida Power & Light... St. Lucie County, Florida. The proposed license amendment would increase the maximum thermal power...

78 FR 79501 - Tennessee Valley Authority, Exemption From the Requirement To Submit an Annual Update to the...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-30

..., Maryland 20852. FOR FURTHER INFORMATION CONTACT: Anthony Minarik, Office of New Reactors, U.S. Nuclear... advanced pressurized water reactors to be constructed and operated at the Bellefonte site, located near the... 052000-15). The NRC docketed the Bellefonte Nuclear Plant, Units 3 and 4 (BLN 3&4) COL application on...

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-05-14

fuel for future civilian light water reactors deployed” in the UAE. The agreement also states that future cooperation may encompass training...planned nuclear reactor . (...continued) May 4, 2008; and, Chris Stanton and Ivan...already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two consulting and advisory services contracts related to the establishment

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.

The Case of Nuclear Propulsion

NASA Technical Reports Server (NTRS)

Koroteev, Anatoly S.; Ponomarev-Stepnoi, Nicolai N.; Smetannikov, Vladimir P.; Gafarov, Albert A.; Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Martin, James; Bragg-Sitton, Shannon; Dickens, Ricky

2003-01-01

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 utilized. Successful utilization will simultaneously develop the infrastructure and experience necessary for developing even higher power and performance systems. To be successful, development programs must devise strategies for rapidly converting paper reactor concepts into actual flight hardware. One approach to accomplishing this is to design highly testable systems, and to structure the program to contain frequent, significant hardware milestones. This paper discusses ongoing efforts in Russia and the United States aimed at enabling near-term utilization of space fission systems.

Safety and Regulatory Issues of the Thorium Fuel Cycle

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

Ade, Brian; Worrall, Andrew; Powers, Jeffrey

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2),more » add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.« less

Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant.

PubMed

Sohrabi, M; Ghasemi, M; Amrollahi, R; Khamooshi, C; Parsouzi, Z

2013-05-01

Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well as beyond design-basis accidents. Such scenarios are planned to be studied in the near future, for this reactor.

UV Disinfection System for Cabin Air

NASA Astrophysics Data System (ADS)

Lim, Soojung

Ultraviolet (UV) radiation is commonly used for disinfection of water. As a result of advancements made in the last 10-15 years, the analysis and design of UV disinfection systems for water is well developed. UV disinfection is also used for disinfection of air; however, despite the fact the UV-air systems have a longer record of application than UV-water systems, the methods used to analyze and design UV-air disinfection systems remain quite empirical. It is well-established that the effectiveness of UV-air systems is strongly affected by the type of microorganisms, the irradiation level/type (lamp power and wavelength), duration of irradiation (exposure time), air movement pattern (mixing degree), and relative humidity. This paper will describe ongoing efforts to evaluate, design and test a UV-air system based on first principles. Specific issues to be addressed in this work will include laboratory measurements of relevant kinetics (i.e., UV dose-response behavior) and numerical simulations designed to represent fluid mechanics and the radiation intensity field. UV dose-response behavior of test microorganism was measured using a laboratory (bench-scale) system. Target microorganisms (e.g., bacterial spores) were first applied to membrane filters at sub-monolayer coverage. The filters were then transferred to an environmental chamber at fixed relative humidity (RH) and allowed to equilibrate with their surroundings. Microorganisms were then subjected to UV exposure under a collimated beam. The experiment was repeated at RH values ranging from 20% to 100%. UV dose-response behavior was observed to vary with RH. For example, at 100% RH, a UV dose of 20 mJ/cm2 accomplished 90% (1 log10 units) of the B. subtilis spore inactivation, whereas 99 % (2 log10 units) inactivation was accomplished at this same UV dose under 20% RH conditions. However, at higher doses, the result was opposite of that in low dose. Reactor behavior is simulated using an integrated application of computational fluid dynamics (CFD) and radiation intensity field models. These simulations followed a Lagrangian approach, wherein the UV radiation intensity field was mapped onto simulated particle trajectories for prediction of the UV dose delivered to each particle. By repeating these calculations for a large number of simulated particle trajectories, an estimate of the UV dose distribution delivered by the reactor can be made. In turn, these dose distribution estimates are integrated with the UV dose-response behavior described above to yield an estimate of microbial inactivation accomplished by the reactor. This modeling approach has the advantage of allowing simulation of many reactor configurations in a relatively short period of time. Moreover, by following this approach of "numerical prototyping," it is possible to "build" and analyze several virtual reactors before the construction of a physical prototype. As such, this procedure allows effective development of efficient reactors.

Design of the Sandia-Israel 20-kW reflux heat-pipe solar receiver/reactor

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

Diver, R.B.; Ginn, W.C.

1987-09-01

This report describes the design and fabrication of a 20-kW sodium reflux heat-pipe solar receiver/reactor for CO/sub 2/ reforming of methane. This project is part of a bilateral agreement between the United States and Israel. Under the terms of the agreement the solar receiver/reactor has been designed and built by Sandia National Laboratories for testing in the 7-meter solar furnace facility at the Weizmann Institute of Science in Rehovot, Israel. 16 refs., 11 figs., 2 tabs.

Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

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

Stoller, R.E.

1987-12-01

The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and alteredmore » mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.« less

Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration

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

J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster

2011-05-31

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topicalmore » areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus on meeting two of the eight needs outlined in the recently published 'Technology Roadmap on Instrumentation, Control, and Human-Machine Interface (ICHMI) to Support DOE Advanced Nuclear Energy Programs' which was created 'to provide a systematic path forward for the integration of new ICHMI technologies in both near-term and future nuclear power plants and the reinvigoration of the U.S. nuclear ICHMI community and capabilities.' The research consortium is led by The University of Tennessee (UT) and is focused on three interrelated topics: Topic 1 (simulator development and measurement sensitivity analysis) is led by Dr. Mike Doster with Dr. Paul Turinsky of North Carolina State University (NCSU). Topic 2 (multivariate autonomous control of modular reactors) is led by Dr. Belle Upadhyaya of the University of Tennessee (UT) and Dr. Robert Edwards of Penn State University (PSU). Topic 3 (monitoring, diagnostics, and prognostics system development) is led by Dr. Wes Hines of UT. Additionally, South Carolina State University (SCSU, Dr. Ken Lewis) participated in this research through summer interns, visiting faculty, and on-campus research projects identified throughout the grant period. Lastly, Westinghouse Science and Technology Center (Dr. Mario Carelli) was a no-cost collaborator and provided design information related to the IRIS demonstration platform and defining needs that may be common to other SMR designs. The results of this research are reported in a six-volume Final Report (including the Executive Summary, Volume 1). Volumes 2 through 6 of the report describe in detail the research and development under the topical areas. This volume serves to introduce the overall NERI-C project and to summarize the key results. Section 2 provides a summary of the significant contributions of this project. A list of all the publications under this project is also given in Section 2. Section 3 provides a brief summary of each of the five volumes (2-6) of the report. The contributions of SCSU are described in Section 4, including a summary of undergraduate research experience. The project management organizational chart is provided as Figure 1. Appendices A, B, and C contain the reports on the summer research performed at the University of Tennessee by undergraduate students from South Carolina State University.« less

Development Status of the Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M.; Pearson, Jon Boise; Godfoy, Thomas

2012-01-01

This paper summarizes the progress that has been made in the development of the Fission Power System Technology Demonstration Unit (TDU). The reactor simulator core and Annular Linear Induction Pump have been fabricated and assembled into a test loop at the NASA Marshall Space Flight Center. A 12 kWe Power Conversion Unit (PCU) is being developed consisting of two 6 kWe free-piston Stirling engines. The two 6 kWe engines have been fabricated by Sunpower Inc. and are currently being tested separately prior to integration into the PCU. The Facility Cooling System (FCS) used to reject convertor waste heat has been assembled and tested at the NASA Glenn Research Center (GRC). The structural elements, including a Buildup Assembly Platform (BAP) and Upper Truss Structure (UTS) have been fabricated, and will be used to test cold-end components in thermal vacuum prior to TDU testing. Once all components have been fully tested at the subsystem level, they will be assembled into an end-to-end system and tested in thermal vacuum at GRC.

Development Status of the Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M; Pearson, Jon Boise; Godfroy, Thomas

2012-01-01

This paper summarizes the progress that has been made in the development of the Fission Power System Technology Demonstration Unit (TDU). The reactor simulator core and Annular Linear Induction Pump have been fabricated and assembled into a test loop at the NASA Marshall Space Flight Center. A 12 kWe Power Conversion Unit (PCU) is being developed consisting of two 6 kWe free-piston Stirling engines. The two 6 kWe engines have been fabricated by Sunpower Inc. and are currently being tested separately prior to integration into the PCU. The Facility Cooling System (FCS) used to reject convertor waste heat has been assembled and tested at the NASA Glenn Research Center (GRC). The structural elements, including a Buildup Assembly Platform (BAP) and Upper Truss Structure (UTS) have been fabricated, and will be used to test cold-end components in thermal vacuum prior to TDU testing. Once all components have been fully tested at the subsystem level, they will be assembled into an end-to-end system and tested in thermal vacuum at NASA GRC.

An innovative sewage sludge reduction by using a combined mesophilic anaerobic and thermophilic aerobic process with thermal-alkaline treatment and sludge recirculation.

PubMed

Cho, Hyun Uk; Park, Sang Kyu; Ha, Jeong Hyub; Park, Jong Moon

2013-11-15

Lab-scale High Efficiency Digestion (HED) systems containing a Mesophilic Anaerobic Reactor (MAR), Thermophilic Aerobic Reactor (TAR), liquid/solid separation unit, and thermal-alkaline treatment were developed to evaluate the efficiencies of sludge reduction and methane production. The HED process was divided into three phases to examine the influence of sludge pretreatment and pretreated sludge recirculation using TCOD and VSS reduction, COD solubilization, and methane production. The VSS removal with a solid/liquid separation unit, sludge recirculation, and thermal-alkaline treatment drastically increased up to 95% compared to the feed concentration. In addition, the results of COD solubilization and VSS/TSS showed that the solubilization of cells and organic matters by the thermal-alkaline treatment was highly increased, which was also consistent with the SEM images. In particular, the methane production rate increased 24-fold when the feed sludge and recirculated sludge were pretreated together. Collectively, the HED experiments performed with sludge recirculation and thermal-alkaline treatment demonstrated that the HED systems can be successfully employed for highly efficient sewage sludge reduction and methane gas production. Copyright © 2013 Elsevier Ltd. All rights reserved.

Upgrading the fuel-handling machine of the Novovoronezh nuclear power plant unit no. 5

NASA Astrophysics Data System (ADS)

Terekhov, D. V.; Dunaev, V. I.

2014-02-01

The calculation of safety parameters was carried out in the process of upgrading the fuel-handling machine (FHM) of the Novovoronezh nuclear power plant (NPP) unit no. 5 based on the results of quantitative safety analysis of nuclear fuel transfer operations using a dynamic logical-and-probabilistic model of the processing procedure. Specific engineering and design concepts that made it possible to reduce the probability of damaging the fuel assemblies (FAs) when performing various technological operations by an order of magnitude and introduce more flexible algorithms into the modernized FHM control system were developed. The results of pilot operation during two refueling campaigns prove that the total reactor shutdown time is lowered.

Staged Catalytic Partial Oxidation (SCPO) System - The State of Art Integrated Short Contact Time Hydrogen Generator

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

Ke Liu; Jin Ki Hong; Wei Wei

Research and development on hydrogen and syngas production have great potential in addressing the following challenges in energy arena: (1) produce more clean fuels to meet the increasing demands for clean liquid and gaseous fuels for transportation and electricity generation, (2) increase the efficiency of energy utilization for fuels and electricity production, and (3) eliminate the pollutants and decouple the link between energy utilization and greenhouse gas emissions in end-use systems [Song, 2006, Liu, Song & Subramani 2009]. In this project, GE Global Research (GEGR) collaborated with Argonne National Laboratory (ANL) and the University of Minnesota (UoMn), developed and demonstratedmore » a low cost, compact staged catalytic partial oxidation (SCPO) technology for distributed hydrogen generation. GEGR analyzed different reforming system designs, and developed the SCPO reforming system which is a unique technology staging and integrating 3 different short contact time catalysts in a single, compact reactor: catalytic partial oxidation (CPO), steam methane reforming (SMR) and water-gas shift (WGS). This integration is demonstrated via the fabrication of a prototype scale unit of each key technology. Approaches for key technical challenges of the program includes: · Analyzed different system designs · Designed the SCPO hydrogen production system · Developed highly active and sulfur tolerant CPO catalysts · Designed and built different pilot-scale reactors to demonstrate each key technology · Evaluated different operating conditions · Quantified the efficiency and cost of the system · Developed process design package (PDP) for 1500 kg H2/day distributed H2 production unit. SCPO met the Department of Energy (DOE) and GE’s cost and efficiency targets for distributed hydrogen production.« less

Development of Technical Basis for Burnup Credit Regulatory Guidance in the United States

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

Parks, Cecil V; Wagner, John C; Mueller, Don

2011-01-01

In the United States (U.S.) there has been and continues to be considerable interest in the increased use of burnup credit as part of the safety basis for SNF systems and this interest has motivated numerous technical studies related to the application of burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission initiated a burnup credit research program, with support from the Oak Ridge National Laboratory, to develop regulatory guidance and the supporting technical basis for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Themore » objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details.« less

Development of position measurement unit for flying inertial fusion energy target

NASA Astrophysics Data System (ADS)

Tsuji, R.; Endo, T.; Yoshida, H.; Norimatsu, T.

2016-03-01

We have reported the present status in the development of a position measurement unit (PMU) for a flying inertial fusion energy (IFE) target. The PMU, which uses Arago spot phenomena, is designed to have a measurement accuracy smaller than 1 μm. By employing divergent, pulsed orthogonal laser beam illumination, we can measure the time and the target position at the pulsed illumination. The two-dimensional Arago spot image is compressed into one-dimensional image by a cylindrical lens for real-time processing. The PMU are set along the injection path of the flying target. The local positions of the target in each PMU are transferred to the controller and analysed to calculate the target trajectory. Two methods are presented to calculate the arrival time and the arrival position of the target at the reactor centre.

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

NONE

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 60%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to grow by about 24% from 2013 to 2040 . At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license, for a total of 60 years of operation (the oldest commercial plants in the Unitedmore » States reached their 40th anniversary in 2009). Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity for 40- and 60-year license periods. If current operating nuclear power plants do not operate beyond 60 years (and new nuclear plants are not built quickly enough to replace them), the total fraction of generated electrical energy from nuclear power will rapidly decline. That decline will be accelerated if plants are shut down before 60 years of operation. Decisions on extended operation ultimately rely on economic factors; however, economics can often be improved through technical advancements. The U.S. Department of Energy Office of Nuclear Energy's 2010 Research and Development Roadmap (2010 Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: 1. Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; 2. Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; 3. Develop sustainable nuclear fuel cycles; and 4. Understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.« less

Biomass-derived Syngas Utilization for Fuels and Chemicals - Final Report

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

Dayton, David C

2010-03-24

Executive Summary The growing gap between petroleum production and demand, mounting environmental concerns, and increasing fuel prices have stimulated intense interest in research and development (R&D) of alternative fuels, both synthetic and bio-derived. Currently, the most technically defined thermochemical route for producing alternative fuels from lignocellulosic biomass involves gasification/reforming of biomass to produce syngas (carbon monoxide [CO] + hydrogen [H2]), followed by syngas cleaning, Fischer-Tropsch synthesis (FTS) or mixed alcohol synthesis, and some product upgrading via hydroprocessing or separation. A detailed techno-economic analysis of this type of process has recently been published [1] and it highlights the need for technicalmore » breakthroughs and technology demonstration for gas cleanup and fuel synthesis. The latter two technical barrier areas contribute 40% of the total thermochemical ethanol cost and 70% of the production cost, if feedstock costs are factored out. Developing and validating technologies that reduce the capital and operating costs of these unit operations will greatly reduce the risk for commercializing integrated biomass gasification/fuel synthesis processes for biofuel production. The objective of this project is to develop and demonstrate new catalysts and catalytic processes that can efficiently convert biomass-derived syngas into diesel fuel and C2-C4 alcohols. The goal is to improve the economics of the processes by improving the catalytic activity and product selectivity, which could lead to commercialization. The project was divided into 4 tasks: Task 1: Reactor Systems: Construction of three reactor systems was a project milestone. Construction of a fixed-bed microreactor (FBR), a continuous stirred tank reactor (CSTR), and a slurry bubble column reactor (SBCR) were completed to meet this milestone. Task 2: Iron Fischer-Tropsch (FT) Catalyst: An attrition resistant iron FT catalyst will be developed and tested. Task 3: Chemical Synthesis: Promising process routes will be identified for synthesis of selected chemicals from biomass-derived syngas. A project milestone was to select promising mixed alcohol catalysts and screen productivity and performance in a fixed bed micro-reactor using bottled syngas. This milestone was successfully completed in collaboration withour catalyst development partner. Task 4: Modeling, Engineering Evaluation, and Commercial Assessment: Mass and energy balances of conceptual commercial embodiment for FT and chemical synthesis were completed.« less

Emerging nuclear programs in Asia: The Phillipines, Thailand, Indonesia, and Pakistan

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

Williams, M.L.

This article is a review of the potential for nuclear energy development in the developing nations of Pakistan, Indonesia, Thailand, and the Philippines. In each country, there is a substantial need for new generating capacity, and each is exploring the idea of having nuclear energy supply a meaningful portion of this new capacity. Of the four countries, only Pakistan is currently a nuclear operator, and one vintage CANDU plant in operation and the Chashma unit under construction. Thailand and Indonesia have ambitious plans to have 12 reactors in service by the year 2015.

A Global Need, a Global Resource Nuclear Power and the New Millennium: In Contrast to the 25 Billion Tons of Carbon Dioxide Emitted into the Atmosphere Each Year as Fossil Fuel Waste, the Spent Fuel Produced Yearly from All the World's Reactors Would Fit inside a Two-Story Structure Built on a Basketball Court

ERIC Educational Resources Information Center

Baird, Stephen L.

2004-01-01

The technological literacy standards were developed to act as a beacon for educators to guide them in their quest to develop a population of technically literate citizens who possess the skills, abilities, and knowledge necessary to actively and constructively participate in the democratic, technologically dependent society of the United States.…

76 FR 70067 - Taxpayers and Ratepayers United, et al.; Environmental Impacts of Severe Reactor and Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-10

... environmental issues raised in the Fukushima Task Force Report. The NRC is not instituting a public comment... Reactor Safety in the 21st Century: The Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident'' (Fukushima Task Force Report, ADAMS Accession No. ML111861807), dated July 12, 2011, as...

75 FR 30451 - Duke Energy Carolinas, LLC; Duke Energy Carolinas, LLC; William States Lee III Combined License...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-01

... Energy Carolinas, LLC; William States Lee III Combined License Application; Notice of Intent To Conduct a... environmental review of the William States Lee III Nuclear Station, Units 1 and 2 combined licenses application...-licensing/col/lee.html '' to `` http://www.nrc.gov/reactors/new-reactors/col/lee.html ''. Dated at Rockville...

Modeling Geometric Arrangements of TiO2-Based Catalyst Substrates and Isotropic Light Sources to Enhance the Efficiency of a Photocatalystic Oxidation (PCO) Reactor

NASA Technical Reports Server (NTRS)

Richards, Jeffrey T.; Levine, Lanfang H.; Husk, Geoffrey K.

2011-01-01

The closed confined environments of the ISS, as well as in future spacecraft for exploration beyond LEO, provide many challenges to crew health. One such challenge is the availability of a robust, energy efficient, and re-generable air revitalization system that controls trace volatile organic contaminants (VOCs) to levels below a specified spacecraft maximum allowable concentration (SMAC). Photocatalytic oxidation (PCO), which is capable of mineralizing VOCs at room temperature and of accommodating a high volumetric flow, is being evaluated as an alternative trace contaminant control technology. In an architecture of a combined air and water management system, placing a PCO unit before a condensing heat exchanger for humidity control will greatly reduce the organic load into the humidity condensate loop ofthe water processing assembly (WPA) thereby enhancing the life cycle economics ofthe WPA. This targeted application dictates a single pass efficiency of greater than 90% for polar VOCs. Although this target was met in laboratory bench-scaled reactors, no commercial or SBIR-developed prototype PCO units examined to date have achieved this goal. Furthermore, the formation of partial oxidation products (e.g., acetaldehyde) was not eliminated. It is known that single pass efficiency and partial oxidation are strongly dependent upon the contact time and catalyst illumination, hence the requirement for an efficient reactor design. The objective of this study is to maximize the apparent contact time and illuminated catalyst surface area at a given reactor volume and volumetric flow. In this study, a Ti02-based photocatalyst is assumed to be immobilized on porous substrate panels and illumination derived from linear isotropic light sources. Mathematical modeling using computational fluid dynamics (CFD) analyses were performed to investigate the effect of: 1) the geometry and configuration of catalyst-coated substrate panels, 2) porosity of the supporting substrate, and 3) varying the light source and spacing on contact time and illuminated catalyst area.

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

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

Kinetic modelling and microbial community assessment of anaerobic biphasic fixed film bioreactor treating distillery spent wash.

PubMed

Acharya, Bhavik K; Pathak, Hilor; Mohana, Sarayu; Shouche, Yogesh; Singh, Vasdev; Madamwar, Datta

2011-08-01

Anaerobic digestion, microbial community structure and kinetics were studied in a biphasic continuously fed, upflow anaerobic fixed film reactor treating high strength distillery wastewater. Treatment efficiency of the bioreactor was investigated at different hydraulic retention times (HRT) and organic loading rates (OLR 5-20 kg COD m⁻³ d⁻¹). Applying the modified Stover-Kincannon model to the reactor, the maximum removal rate constant (U(max)) and saturation value constant (K(B)) were found to be 2 kg m⁻³ d⁻¹ and 1.69 kg m⁻³ d⁻¹ respectively. Bacterial community structures of acidogenic and methanogenic reactors were assessed using culture-independent analyses. Sequencing of 16S rRNA genes exhibited a total of 123 distinct operational taxonomic units (OTUs) comprising 49 from acidogenic reactor and 74 (28 of eubacteria and 46 of archaea) from methanogenic reactor. The findings reveal the role of Lactobacillus sp. (Firmicutes) as dominant acid producing organisms in acidogenic reactor and Methanoculleus sp. (Euryarchaeotes) as foremost methanogens in methanogenic reactor. Copyright © 2011 Elsevier Ltd. All rights reserved.

Exploring the Deployment Potential of Small Modular Reactors

NASA Astrophysics Data System (ADS)

Abdulla, Ahmed Y.

This thesis reports the results of several investigations into the viability of an emergent technology. Due to the lack of data in such cases, and the sensitivity surrounding nuclear power, exploring the potential of small modular reactors (SMRs) proved challenging. Moreover, these reactors come in a wide range of sizes and can employ a number of technologies, which made investigating the category as a whole difficult. We started by looking at a subset of SMRs that were the most promising candidates for near to mid-term deployment: integral light water SMRs. We conducted a technically detailed elicitation of expert assessments of their capital costs and construction duration, focusing on five reactor deployment scenarios that involved a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants varied by more than a factor of 2.5. Expert judgments about likely SMR costs displayed an even wider range. There was consensus that an SMR plant's construction duration would be shorter than a large reactor's. Experts identified more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable, though these reactors do not constitute a paradigm shift when it comes to nuclear power's safety and security. Using these expert assessments of cost and construction duration, we calculated levelized cost of electricity values for four of the five scenarios. For the large plant, median levelized cost estimates ranged from 56 to 120 per MWh. Median estimates of levelized cost ranged from 77 to 240 per MWh for a 45MWe SMR, and from 65 to 120 per MWh for a 225MWe unit. We concluded that controlling construction duration is important, though not as important a factor in the analysis as capital cost, and, given the price of electricity in some parts of the U.S., it is possible to construct an argument for deploying SMRs in certain locations. We then decided to investigate the technical and institutional barriers hampering the development and deployment of a subset of six SMRs, including two light water designs and four non-light water advanced designs. We organized an invitational workshop that became an integrated assessment of various designs and of the institutional innovations required to bring SMRs to market. Some valuable insights were gleaned from the workshop: there is consensus that many of the challenges facing advanced SMRs are rooted in institutional biases in favor of the light water economy, as opposed to technical ones. The institutional factors that are judged to pose the greatest challenge to the mass deployment of SMRs are: the lack of a greenhouse gas control regime; political and financial instability; public concerns about nuclear safety and waste; and inadequate national and international institutions. When asked what factors most help promote SMR adoption in OECD and developing countries, economic factors dominate the list of characteristics that most contribute to their promotion in OECD countries but, when it comes to developing countries, institutional factors are regarded as being of highest import. Safety of design and safety in operation are judged the most important characteristic on both lists.

Testing of Liquid Metal Components for Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.; Godfroy, Thomas J.; Pearson, J. Boise

2010-01-01

The Early Flight Fission Test Facility (EFF-TF) was established by the Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations. The EFF-TF is currently supporting an effort to develop an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled (Sodium-Potassium eutectic, NaK-78) reactor design. This design was derived from the only fission system that the United States has deployed for space operation, the Systems for Nuclear Auxiliary Power (SNAP) 10A reactor, which was launched in 1965. Two prototypical components recently tested at MSFC were a pair of Stirling power conversion units that would be used in a reactor system to convert heat to electricity, and an annular linear induction pump (ALIP) that uses travelling electromagnetic fields to pump the liquid metal coolant through the reactor loop. First ever tests were conducted at MSFC to determine baseline performance of a pair of 1 kW Stirling convertors using NaK as the hot side working fluid. A special test rig was designed and constructed and testing was conducted inside a vacuum chamber at MSFC. This test rig delivered pumped NaK for the hot end temperature to the Stirlings and water as the working fluid on the cold end temperature. These test were conducted through a hot end temperature range between 400 to 550C in increments of 50 C and a cold end temperature range from 30 to 70 C in 20 C increments. Piston amplitudes were varied from 6 to 1 1mm in .5 mm increments. A maximum of 2240 Watts electric was produced at the design point of 550 hot end, 40 C cold end with a piston amplitude of 10.5mm. This power level was reached at a gross thermal efficiency of 28%. A baseline performance map was established for the pair of 1kW Stirling convertors. The performance data will then be used for design modification to the Stirling convertors. The ALIP tested at MSFC has no moving parts and no direct electrical connections to the liquid metal containing components. Pressure is developed by the interaction of the magnetic field produced by the stator and the current which flows as a result of the voltage induced in the liquid metal contained in the pump duct. Flow is controlled by variation of the voltage supplied to the pump windings. Under steady-state conditions, pump performance is measured for flow rates from 0.5-4.3 kg/s. The pressure rise developed by the pump to support these flow rates is roughly 5-65 kPa. The RMS input voltage (phase-to-phase voltage) ranges from 5-120 V, while the frequency can be varied arbitrarily up to 60 Hz. Performance is quantified at different loop temperature levels from 50 C up to 650 C, which is the peak operating temperature of the proposed AFSP reactor. The transient response of the pump is also evaluated to determine its behavior during startup and shut-down procedures.

Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

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

Bsebsu, F.M.; Abotweirat, F.; Elwaer, S.

2008-07-15

The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulicmore » design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)« less

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

Wood Storks of the Birdsville Colony and swamps of the Savannah River site

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

Coulter, M.C.

1989-08-01

The population of Wood Storks (Mycteria americana) that breeds in the United States has decreased from an estimated 20,000 pairs in 1930 to just under 5,000 pairs in 1980. The decline has prompted the United States Fish an Wildlife Service to list the United States breeding population of Wood Storks as endangered. When the US Department of Energy (USDOE) decided to restart L-Reactor on the SRS there was concern that when the reactor was restarted, cooling water flowing into the Steel Creek Delta would raise the water level and the area would become too deep for foraging storks. USDOE beganmore » consultation with the US Fish and Wildlife Service (USFWS) and agreed to develop and maintain alternative foraging habitat to replace the potential loss so the Kathwood Foraging Ponds were developed. We have examined the breeding biology of the birds at the colony to determine when the birds may have maximum food demand and we have studied foraging ecology of the birds. The objectives of the work carried out in 1989 were: to determine the locations of foraging sites of Wood Storks from the Birdsville Colony ad examine year-to-year variation in sites used; to characterize in more detail the habitat, vegetation, water quality and prey density/biomass at these sites; to observe the breeding birds to determine times when food demands at the colony are greatest; to examine the movements of storks from the rookery to foraging sites and related seasonal trends to the breeding biology; to examine the importance of the Savannah River Site Swamp (SRSS) to foraging Wood Storks; and to examine the movements of individual birds to determine the generality of the observed patterns. 28 refs., 100 figs., 78 tabs.« less

Superfund record of decision (EPA region 10): Idaho National Engineering Lab, (USDOE) Operable Unit 26 (Stationary Low-Power Reactor-1 and Boiling Water Reactor Experiment-I Burial Grounds), Idaho Falls, ID, December 1, 1995

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

NONE

1997-03-01

This document presents the selected remedial action for the Stationary Low-Power Reactor-1 (SL-1) burial ground, the Boiling Water Reactor Experiment-I (BORAX-I) burial ground, and 10 no action sites in Waste Area Group 5. Actual or threatened releases of hazardous substances from the SL-1 and BORAX-I burial grounds, if not addressed by implementing the response action selected in this Record of Decision, may present a current or potential threat to public health, welfare, or the environment. The 10 no action sites do not present a threat to human health or the environment.

The benefits of an advanced fast reactor fuel cycle for plutonium management

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

Hannum, W.H.; McFarlane, H.F.; Wade, D.C.

1996-12-31

The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium andmore » long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.« less

Development of a Toolbox Using Chemical, Physical and Biological Technologies for Decontamination of Sediments to Support Strategic Army Response to Natural Disasters

DTIC Science & Technology

2006-11-01

disinfection) was tested using soil microcosms and respirometry to determine diesel range and total organic compound degradation. These tests were...grease) such as benzo(a)pyrene were detected above chronic (long term-measured in years) screening levels. Levels of diesel and oil range organics... bioremediation , and toxicity of liquid and solid samples. The Comput-OX 4R is a 4 reactor unit with no stirring modules or temperature controlled water bath

Retrospective Study of Selected DoD Materials and Structures Research and Development Programs. Phase 1. Case History Data Collection

DTIC Science & Technology

1979-03-01

made in continuous form by reducing boron trichloride with hydrogen and depositing the elemental boron formed on an electrically heated, continuously...filament take-up unit. A stoichio- metric mixture of boron trichloride and hydrogen is introduced at the top of the reactor. These react at the surface of...fibers are tungsten wire, boron trichloride , and hydrogen gas. The fine diameter tungsten wire on which boron is deposited is an imported product and is

Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 5: Conclusions and recomendations

NASA Technical Reports Server (NTRS)

Williams, J. R.

1974-01-01

Air pollution resulting from the use of fossil fuels is discussed. Phenomena relating to the emission of CO2 such as the greenhouse effect and multiplier effect are explored. Particulate release is also discussed. The following recommendations are made for the elimination of fossil fuel combustion products in the United States: development of nuclear breeder reactors, use of solar energy systems, exploration of energy alternatives such as geothermal and fusion, and the substitution of coal for gas and oil use.

Development of an ultrahigh-temperature process for the enzymatic hydrolysis of lactose. IV. Immobilization of two thermostable beta-glycosidases and optimization of a packed-bed reactor for lactose conversion.

PubMed

Petzelbauer, Inge; Kuhn, Bernhard; Splechtna, Barbara; Kulbe, Klaus D; Nidetzky, Bernd

2002-03-20

Recombinant hyperthermostable beta-glycosidases from the archaea Sulfolobus solfataricus (Ss beta Gly) and Pyrococcus furiosus (CelB) were covalently attached onto the insoluble carriers chitosan, controlled pore glass (CPG), and Eupergit C. For each enzyme/carrier pair, the protein-binding capacity, the immobilization yield, the pH profiles for activity and stability, the activity/temperature profile, and the kinetic constants for lactose hydrolysis at 70 degrees C were determined. Eupergit C was best among the carriers in regard to retention of native-like activity and stability of Ss beta Gly and CelB over the pH range 3.0-7.5. Its protein binding capacity of approximately 0.003 (on a mass basis) was one-third times that of CPG, while immobilization yields were typically 80% in each case. Activation energies for lactose conversion by the immobilized enzymes at pH 5.5 were in the range 50-60 kJ/mol. This is compared to values of approximately 75 kJ/mol for the free enzymes. Immobilization expands the useful pH range for CelB and Ss beta Gly by approximately 1.5 pH units toward pH 3.5 and pH 4.5, respectively. A packed-bed enzyme reactor was developed for the continuous conversion of lactose in different media, including whey and milk, and operated over extended reaction times of up to 14 days. The productivities of the Eupergit C-immobilized enzyme reactor were determined at dilution rates between 1 and 12 h(-1), and using 45 and 170 g/L initial lactose. Results of kinetic modeling for the same reactor, assuming plug flow and steady state, suggest the presence of mass-transfer limitation of the reaction rate under the conditions used. Formation of galacto-oligosaccharides in the continuous packed-bed reactor and in the batch reactor using free enzyme was closely similar in regard to yield and individual saccharide components produced. Copyright 2002 John Wiley & Sons, Inc. Biotechnol Bioeng 77: 619-631, 2002; DOI 10.1002/bit.10110

Incorporating Equipment Condition Assessment in Risk Monitors for Advanced Small Modular Reactors

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

Coble, Jamie B.; Coles, Garill A.; Meyer, Ryan M.

2013-10-01

Advanced small modular reactors (aSMRs) can complement the current fleet of large light-water reactors in the USA for baseload and peak demand power production and process heat applications (e.g., water desalination, shale oil extraction, hydrogen production). The day-to-day costs of aSMRs are expected to be dominated by operations and maintenance (O&M); however, the effect of diverse operating missions and unit modularity on O&M is not fully understood. These costs could potentially be reduced by optimized scheduling, with risk-informed scheduling of maintenance, repair, and replacement of equipment. Currently, most nuclear power plants have a “living” probabilistic risk assessment (PRA), which reflectsmore » the as-operated, as-modified plant and combine event probabilities with population-based probability of failure (POF) for key components. “Risk monitors” extend the PRA by incorporating the actual and dynamic plant configuration (equipment availability, operating regime, environmental conditions, etc.) into risk assessment. In fact, PRAs are more integrated into plant management in today’s nuclear power plants than at any other time in the history of nuclear power. However, population-based POF curves are still used to populate fault trees; this approach neglects the time-varying condition of equipment that is relied on during standard and non-standard configurations. Equipment condition monitoring techniques can be used to estimate the component POF. Incorporating this unit-specific estimate of POF in the risk monitor can provide a more accurate estimate of risk in different operating and maintenance configurations. This enhanced risk assessment will be especially important for aSMRs that have advanced component designs, which don’t have an available operating history to draw from, and often use passive design features, which present challenges to PRA. This paper presents the requirements and technical gaps for developing a framework to integrate unit-specific estimates of POF into risk monitors, resulting in enhanced risk monitors that support optimized operation and maintenance of aSMRs.« less

AGC-2 Irradiation Report

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

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

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

Hybrid reactor based on combined cavitation and ozonation: from concept to practical reality.

PubMed

Gogate, P R; Mededovic-Thagard, S; McGuire, D; Chapas, G; Blackmon, J; Cathey, R

2014-03-01

The present work gives an in depth discussion related to the development of a hybrid advanced oxidation reactor, which can be effectively used for the treatment of various types of water. The reactor is based on the principle of intensifying degradation/disinfection using a combination of hydrodynamic cavitation, acoustic cavitation, ozone injection and electrochemical oxidation/precipitation. Theoretical studies have been presented to highlight the uniform distribution of the cavitational activity and enhanced generation of hydroxyl radicals in the cavitation zone, as well as higher turbulence in the main reactor zone. The combination of these different oxidation technologies have been shown to result in enhanced water treatment ability, which can be attributed to the enhanced generation of hydroxyl radicals, enhanced contact of ozone and contaminants, and the elimination of mass transfer resistances during electrochemical oxidation/precipitation. Compared to the use of individual approaches, the hybrid reactor is expected to intensify the treatment process by 5-20 times, depending on the application in question, which can be confirmed based on the literature illustrations. Also, the use of Ozonix® has been successfully proven while processing recycled fluids at commercial sites on over 750 oil and natural gas wells during hydraulic operations around the United States. The superiority of the hybrid process over conventional chemical treatments in terms of bacteria and scale reduction as well as increased water flowability and better chemical compatibility, which is a key requirement for oil and gas applications, has been established. Copyright © 2013 Elsevier B.V. All rights reserved.

Nuclear reactors built, being built, or planned, 1994

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

NONE

1995-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

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

NASA Astrophysics Data System (ADS)

Fratoni, Massimiliano

This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

Plasma core reactor simulations using RF uranium seeded argon discharges

NASA Technical Reports Server (NTRS)

Roman, W. C.

1975-01-01

An experimental investigation was conducted using the United Technologies Research Center (UTRC) 80 kW and 1.2 MW RF induction heater systems to aid in developing the technology necessary for designing a self-critical fissioning uranium plasma core reactor (PCR). A nonfissioning, steady-state RF-heated argon plasma seeded with pure uranium hexafluoride (UF6) was used. An overall objective was to achieve maximum confinement of uranium vapor within the plasma while simultaneously minimizing the uranium compound wall deposition. Exploratory tests were conducted using the 80 kW RF induction heater with the test chamber at approximately atmospheric pressure and discharge power levels on the order of 10 kW. Four different test chamber flow configurations were tested to permit selection of the configuration offering the best confinement characteristics for subsequent tests at higher pressure and power in the 1.2 MW RF induction heater facility.

Summary of the Forty-Fifth NCRP annual meeting on "the future of nuclear power worldwide: safety, health and the environment".

PubMed

Corradini, Michael

2011-01-01

The role of nuclear power as a major resource in meeting the projected growth of electric power requirements in the United States and worldwide during the 21st century is a subject of great contemporary interest. The goal of the 2009 NCRP Annual Meeting was to provide a forum for an in-depth discussion of issues related to the safety, health and environmental protection aspects of new nuclear power reactor systems and related fuel-cycle facilities such as fuel production and reprocessing strategies. The meeting was an international conference with participation of almost 400 representatives from many nations, scientific organizations, nuclear industries, and governmental agencies engaged in the development and regulatory control of advanced nuclear reactor systems and fuel-cycle operations. Highlights of the meeting are summarized in this report. Copyright © 2010 Health Physics Society

Development of flow systems by direct-milling on poly(methyl methacrylate) substrates using UV-photopolymerization as sealing process.

PubMed

Rodrigues, Eunice R G O; Lapa, Rui A S

2009-03-01

An alternative process for the design and construction of fluidic devices is presented. Several sealing processes were studied, as well as the hydrodynamic characteristics of the proposed fluidic devices. Manifolds were imprinted on polymeric substrates by direct-write milling, according to Computer Assisted Design (CAD) data. Poly(methyl methacrylate) (PMMA) was used as substrate due to its physical and chemical properties. Different bonding approaches for the imprinted channels were evaluated and UV-photopolymerization of acrylic acid (AA) was selected. The hydrodynamic characteristics of the proposed flow devices were assessed and compared to those obtained in similar flow systems using PTFE reactors and micro-pumps as propulsion units (multi-pumping approach). The applicability of the imprinted reactors was evaluated in the sequential determination of calcium and magnesium in water samples. Results obtained were in good agreement with those obtained by the reference procedure.

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE PAGES

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin; ...

2017-12-22

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

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

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

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.

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

Kupca, L.; Beno, P.

A very brief summary is provided of a primary circuit piping material properties analysis. The analysis was performed for the Bohunice V-1 reactor and the Kola-1 and -2 reactors. Assessment was performed on Bohunice V-1 archive materials and primary piping material cut from the Kola units after 100,000 hours of operation. Main research program tasks included analysis of mechanical properties, corrosion stability, and microstructural properties. Analysis results are not provided.

DEMINERALIZER BUILDING, TRA608. INSTALLATION OF SAMPLING AND OTHER INSTRUMENTS COMPLETES ...

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

DEMINERALIZER BUILDING, TRA-608. INSTALLATION OF SAMPLING AND OTHER INSTRUMENTS COMPLETES DEMINERALIZER UNITS ALONG NORTH WALL. CAMERA FACES EAST. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON THE ORIGINAL NEGATIVE. INL NEGATIVE NO. 3996A. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Study of the Dynamics of a Pressurized Reactor as Integral Unit for Naval Propulsion; STUDIO DELLA DINAMICA DI UN REATTORE PRESSURIZZATO AD UNITA INTEGRALE PER PROPULSIONE NAVALE

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

Gilardi, E.; Cimorelli, L.

1963-07-01

The dynamic behavior of an integrated, pressurizedwater reactor with natural circulation was investigated both by analog computer techniques and a simplified analytical approach. Hydraulic instabilities due to the core or riser were considered, as well as overall stability and problems arising from heavy sea conditions. (auth)

Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312

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

Bracey, William; Bondre, Jayant; Shelton, Catherine

2013-07-01

The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In Januarymore » 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to evaluate the solutions, and the alternative solutions. The complexity of the project is increasing with time (more fuel assemblies, new storage systems, deteriorating logistics infrastructure at some sites, etc.) but with the uncertainty on the final disposal path, flexibility and simplicity will be critical. (authors)« less

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

Research Program of a Super Fast Reactor

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

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

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

Fuel processing for PEM fuel cells: transport and kinetic issues of system design

NASA Astrophysics Data System (ADS)

Zalc, J. M.; Löffler, D. G.

In light of the distribution and storage issues associated with hydrogen, efficient on-board fuel processing will be a significant factor in the implementation of PEM fuel cells for automotive applications. Here, we apply basic chemical engineering principles to gain insight into the factors that limit performance in each component of a fuel processor. A system consisting of a plate reactor steam reformer, water-gas shift unit, and preferential oxidation reactor is used as a case study. It is found that for a steam reformer based on catalyst-coated foils, mass transfer from the bulk gas to the catalyst surface is the limiting process. The water-gas shift reactor is expected to be the largest component of the fuel processor and is limited by intrinsic catalyst activity, while a successful preferential oxidation unit depends on strict temperature control in order to minimize parasitic hydrogen oxidation. This stepwise approach of sequentially eliminating rate-limiting processes can be used to identify possible means of performance enhancement in a broad range of applications.

Global threat reduction initiative Russian nuclear material removal progress

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

Cummins, Kelly; Bolshinsky, Igor

2008-07-15

In December 1999 representatives from the United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) started discussing a program to return to Russia Soviet- or Russian-supplied highly enriched uranium (HEU) fuel stored at the Russian-designed research reactors outside Russia. Trilateral discussions among the United States, Russian Federation, and the International Atomic Energy Agency (IAEA) have identified more than 20 research reactors in 17 countries that have Soviet- or Russian-supplied HEU fuel. The Global Threat Reduction Initiative's Russian Research Reactor Fuel Return Program is an important aspect of the U.S. Government's commitment to cooperate with the other nationsmore » to prevent the proliferation of nuclear weapons and weapons-usable proliferation-attractive nuclear materials. To date, 496 kilograms of Russian-origin HEU have been shipped to Russia from Serbia, Latvia, Libya, Uzbekistan, Romania, Bulgaria, Poland, Germany, and the Czech Republic. The pilot spent fuel shipment from Uzbekistan to Russia was completed in April 2006. (author)« less

ALARA Council: Sharing our resources and experiences to reduce doses at Commonwealth Edison Facilities

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

Rescek, F.

1995-03-01

Commonwealth Edison Company is an investor-owned utility company supplying electricity to over three million customers (eight million people) in Chicago and northern Illinois, USA. The company operates 16 generating stations which have the capacity to produce 22,522 megawatts of electricity. Six of these generating stations, containing 12 nuclear units, supply 51% of this capacity. The 12 nuclear units are comprised of four General Electric boiling water (BWR-3) reactors, two General Electric BWR-5 reactors, and six Westinghouse four-loop pressurized water reactors (PWR). In August 1993, Commonwealth Edison created an ALARA Council with the responsibility to provide leadership and guidance that resultsmore » in an effective ALARA Culture within the Nuclear Operations Division. Unlike its predecessor, the Corporate ALARA Committee, the ALARA Council is designed to bring together senior managers from the six nuclear stations and corporate to create a collaborative effort to reduce occupational doses at Commonwealth Edison`s stations.« 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.

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

NASA Astrophysics Data System (ADS)

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

2015-11-01

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

The role of inertial fusion energy in the energy marketplace of the 21st century and beyond

NASA Astrophysics Data System (ADS)

John Perkins, L.

The viability of inertial fusion in the 21st century and beyond will be determined by its ultimate cost, complexity, and development path relative to other competing, long term, primary energy sources. We examine this potential marketplace in terms of projections for population growth, energy demands, competing fuel sources and environmental constraints (CO 2), and show that the two competitors for inertial fusion energy (IFE) in the medium and long term are methane gas hydrates and advanced, breeder fission; both have potential fuel reserves that will last for thousands of years. Relative to other classes of fusion concepts, we argue that the single largest advantage of the inertial route is the perception by future customers that the IFE fusion power core could achieve credible capacity factors, a result of its relative simplicity, the decoupling of the driver and reactor chamber, and the potential to employ thick liquid walls. In particular, we show that the size, cost and complexity of the IFE reactor chamber is little different to a fission reactor vessel of the same thermal power. Therefore, relative to fission, because of IFE's tangible advantages in safety, environment, waste disposal, fuel supply and proliferation, our research in advanced targets and innovative drivers can lead to a certain, reduced-size driver at which future utility executives will be indifferent to the choice of an advanced fission plant or an advanced IFE power plant; from this point on, we have a competitive commercial product. Finally, given that the major potential customer for energy in the next century is the present developing world, we put the case for future IFE "reservations" which could be viable propositions providing sufficient reliability and redundancy can be realized for each modular reactor unit.

DOE-GO-14154-1 OHIO FINAL report Velocys 30Sept08

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

Terry J. Mazanec

2008-09-30

The overall goal of the OHIO project was to develop a commercially viable high intensity process to produce ethylene by controlled catalytic reaction of ethane with oxygen in a microchannel reactor. Microchannel technology provides a breakthrough solution to the challenges identified in earlier development work on catalytic ethane oxidation. Heat and mass transfer limitations at the catalyst surface create destructively high temperatures that are responsible for increased production of waste products (CO, CO2, and CH4). The OHIO project focused on microscale energy and mass transfer management, designed to alleviate these transport limitations, thereby improving catalyst selectivity and saving energy-rich feedstock.more » The OHIO project evaluated ethane oxidation in small scale microchannel laboratory reactors including catalyst test units, and full commercial length single- and multi-channel reactors. Small scale catalyst and single channel results met target values for ethylene yields, demonstrating that the microchannel concept improves mass and heat transport compared to conventional reactors and results in improved ethylene yield. Earlier economic sensitivity studies of ethane oxidation processes suggested that only modest improvements were necessary to provide a system that provides significant feedstock, energy, and capital benefits compared to conventional steam ethane cracking. The key benefit derived from the OHIO process is energy savings. Ethylene production consumes more energy than any other U.S. chemical process.1 The OHIO process offers improved feedstock utilization and substantial energy savings due to a novel reaction pathway and the unique abilities of microchannel process technology to control the reaction temperature and other critical process parameters. Based on projected economic benefits of the process, the potential energy savings could reach 150 trillion Btu/yr by the year 2020, which is the equivalent of over 25 million barrels of oil.« less

Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)

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

Dearing, J.F.

The Subchannel Analysis of Blockages in Reactor Elements (SABRE) computer code, developed by the United Kingdom Atomic Energy Authority, is currently the only practical tool available for performing detailed analyses of velocity and temperature fields in the recirculating flow regions downstream of blockages in liquid-metal fast breeder reactor (LMFBR) pin bundles. SABRE is a subchannel analysis code; that is, it accurately represents the complex geometry of nuclear fuel pins arranged on a triangular lattice. The results of SABRE computational models are compared here with temperature data from two out-of-pile 19-pin test bundles from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility atmore » Oak Ridge National Laboratory. One of these bundles has a small central flow blockage (bundle 3A), while the other has a large edge blockage (bundle 5A). Values that give best agreement with experiment for the empirical thermal mixing correlation factor, FMIX, in SABRE are suggested. These values of FMIX are Reynolds-number dependent, however, indicating that the coded turbulent mixing correlation is not appropriate for wire-wrap pin bundles.« less

Performance and Mechanisms of Ultrafiltration Membrane Fouling Mitigation by Coupling Coagulation and Applied Electric Field in a Novel Electrocoagulation Membrane Reactor.

PubMed

Sun, Jingqiu; Hu, Chengzhi; Tong, Tiezheng; Zhao, Kai; Qu, Jiuhui; Liu, Huijuan; Elimelech, Menachem

2017-08-01

A novel electrocoagulation membrane reactor (ECMR) was developed, in which ultrafiltration (UF) membrane modules are placed between electrodes to improve effluent water quality and reduce membrane fouling. Experiments with feedwater containing clays (kaolinite) and natural organic matter (humic acid) revealed that the combined effect of coagulation and electric field mitigated membrane fouling in the ECMR, resulting in higher water flux than the conventional combination of electrocoagulation and UF in separate units (EC-UF). Higher current densities and weakly acidic pH in the EMCR favored faster generation of large flocs and effectively reduced membrane pore blocking. The hydraulic resistance of the formed cake layers on the membrane surface in ECMR was reduced due to an increase in cake layer porosity and polarity, induced by both coagulation and the applied electric field. The formation of a polarized cake layer was controlled by the applied current density and voltage, with cake layers formed under higher electric field strengths showing higher porosity and hydrophilicity. Compared to EC-UF, ECMR has a smaller footprint and could achieve significant energy savings due to improved fouling resistance and a more compact reactor design.

A model for the release, dispersion and environmental impact of a postulated reactor accident from a submerged commercial nuclear power plant

NASA Astrophysics Data System (ADS)

Bertch, Timothy Creston

1998-12-01

Nuclear power plants are inherently suitable for submerged applications and could provide power to the shore power grid or support future underwater applications. The technology exists today and the construction of a submerged commercial nuclear power plant may become desirable. A submerged reactor is safer to humans because the infinite supply of water for heat removal, particulate retention in the water column, sedimentation to the ocean floor and inherent shielding of the aquatic environment would significantly mitigate the effects of a reactor accident. A better understanding of reactor operation in this new environment is required to quantify the radioecological impact and to determine the suitability of this concept. The impact of release to the environment from a severe reactor accident is a new aspect of the field of marine radioecology. Current efforts have been centered on radioecological impacts of nuclear waste disposal, nuclear weapons testing fallout and shore nuclear plant discharges. This dissertation examines the environmental impact of a severe reactor accident in a submerged commercial nuclear power plant, modeling a postulated site on the Atlantic continental shelf adjacent to the United States. This effort models the effects of geography, decay, particle transport/dispersion, bioaccumulation and elimination with associated dose commitment. The use of a source term equivalent to the release from Chernobyl allows comparison between the impacts of that accident and the postulated submerged commercial reactor plant accident. All input parameters are evaluated using sensitivity analysis. The effect of the release on marine biota is determined. Study of the pathways to humans from gaseous radionuclides, consumption of contaminated marine biota and direct exposure as contaminated water reaches the shoreline is conducted. The model developed by this effort predicts a significant mitigation of the radioecological impact of the reactor accident release with a submerged commercial nuclear power plant. The two box models predict the most of the radio-ecological impact occurs during the first eight days after release. The most significant risk to humans is from consumption of biota. The reduction in impact to humans from a large radioactive release makes the concept worthy of further study.

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

Bhattacharyya, D.; Turton, R.; Zitney, S.

In this presentation, development of a plant-wide dynamic model of an advanced Integrated Gasification Combined Cycle (IGCC) plant with CO2 capture will be discussed. The IGCC reference plant generates 640 MWe of net power using Illinois No.6 coal as the feed. The plant includes an entrained, downflow, General Electric Energy (GEE) gasifier with a radiant syngas cooler (RSC), a two-stage water gas shift (WGS) conversion process, and two advanced 'F' class combustion turbines partially integrated with an elevated-pressure air separation unit (ASU). A subcritical steam cycle is considered for heat recovery steam generation. Syngas is selectively cleaned by a SELEXOLmore » acid gas removal (AGR) process. Sulfur is recovered using a two-train Claus unit with tail gas recycle to the AGR. A multistage intercooled compressor is used for compressing CO2 to the pressure required for sequestration. Using Illinois No.6 coal, the reference plant generates 640 MWe of net power. The plant-wide steady-state and dynamic IGCC simulations have been generated using the Aspen Plus{reg_sign} and Aspen Plus Dynamics{reg_sign} process simulators, respectively. The model is generated based on the Case 2 IGCC configuration detailed in the study available in the NETL website1. The GEE gasifier is represented with a restricted equilibrium reactor model where the temperature approach to equilibrium for individual reactions can be modified based on the experimental data. In this radiant-only configuration, the syngas from the Radiant Syngas Cooler (RSC) is quenched in a scrubber. The blackwater from the scrubber bottom is further cleaned in the blackwater treatment plant. The cleaned water is returned back to the scrubber and also used for slurry preparation. The acid gas from the sour water stripper (SWS) is sent to the Claus plant. The syngas from the scrubber passes through a sour shift process. The WGS reactors are modeled as adiabatic plug flow reactors with rigorous kinetics based on the mid-life activity of the shift-catalyst. The SELEXOL unit consists of the H2S and CO2 absorbers that are designed to meet the stringent environmental limits and requirements of other associated units. The model also considers the stripper for recovering H2S that is sent as a feed to a split-flow Claus unit. The tail gas from the Claus unit is recycled to the SELEXOL unit. The cleaned syngas is sent to the GE 7FB gas turbine. This turbine is modeled as per published data in the literature. Diluent N2 is used from the elevated-pressure ASU for reducing the NOx formation. The heat recovery steam generator (HRSG) is modeled by considering generation of high-pressure, intermediate-pressure, and low-pressure steam. All of the vessels, reactors, heat exchangers, and the columns have been sized. The basic IGCC process control structure has been synthesized by standard guidelines and existing practices. The steady-state simulation is solved in sequential-modular mode in Aspen Plus{reg_sign} and consists of more than 300 unit operations, 33 design specs, and 16 calculator blocks. The equation-oriented dynamic simulation consists of more than 100,000 equations solved using a multi-step Gear's integrator in Aspen Plus Dynamics{reg_sign}. The challenges faced in solving the dynamic model and key transient results from this dynamic model will also be discussed.« less

Consumption of the electric power inside silent discharge reactors

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

Yehia, Ashraf, E-mail: yehia30161@yahoo.com

An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodesmore » in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.« 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.

40 CFR 61.67 - Emission tests.

Code of Federal Regulations, 2011 CFR

2011-07-01

... = Conversion factor from ppmw to units of emission standard, 0.001 (metric units) = 0.002 (English units) PPVC...(a), or § 61.64(a)(1), (b), (c), or (d), or from any control system to which reactor emissions are... conversion factor, 1,000 g/kg (1 lb/lb). 10−6 = Conversion factor for ppm. Z = Production rate, kg/hr (lb/hr...

40 CFR 61.67 - Emission tests.

Code of Federal Regulations, 2012 CFR

2012-07-01

... = Conversion factor from ppmw to units of emission standard, 0.001 (metric units) = 0.002 (English units) PPVC...(a), or § 61.64(a)(1), (b), (c), or (d), or from any control system to which reactor emissions are... conversion factor, 1,000 g/kg (1 lb/lb). 10−6 = Conversion factor for ppm. Z = Production rate, kg/hr (lb/hr...

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

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

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

1983-06-01

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

Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

NASA Astrophysics Data System (ADS)

Tereshonok, V. A.; Kryakvin, L. V.; Pitilimov, V. A.; Karpov, S. A.; Kulikov, V. I.; Zhylmaganbetov, N. M.; Kavun, O. Yu.; Popykin, A. I.; Shevchenko, R. A.; Shevchenko, S. A.; Semenova, T. V.

2017-12-01

The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

Development of a Korean reference HLW disposal system under the Korean representative geologic conditions

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

Choi, Heui-Joo; Lee, Jong Youl; Choi, Jongwon

2007-07-01

The development of a Korean Reference disposal System for the spent fuels from PWR and CANDU reactors is outlined in this paper. Around 36,000 tU of spent fuels are being projected based on the lifetimes of 28 nuclear power reactors in Korea. Since the site for the geological disposal has not yet been decided, a hypothetical site with representative Korean geologic conditions is proposed for the conceptual design of the repository. The disposal rates of the spent fuels are determined according to the total operation time of 55 years. The canisters are optimized by considering natural Korean conditions, and themore » buffer is designed with domestic Ca-bentonite. The depth of the repository is determined to be 500 m below the ground's surface. The canister separation distances are determined through a thermal analysis. The main features of the repository are presented from the layout to the closure. A computer program has been developed to calculate and analyze the volume and the area of the disposal system to help in the cost analysis. The final output of the design is presented as a unit disposal cost, US $315 /kgU. (authors)« less

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.

Flow dynamics in bioreactors containing tissue engineering scaffolds.

PubMed

Lawrence, Benjamin J; Devarapalli, Mamatha; Madihally, Sundararajan V

2009-02-15

Bioreactors are widely used in tissue engineering as a way to distribute nutrients within porous materials and provide physical stimulus required by many tissues. However, the fluid dynamics within the large porous structure are not well understood. In this study, we explored the effect of reactor geometry by using rectangular and circular reactors with three different inlet and outlet patterns. Geometries were simulated with and without the porous structure using the computational fluid dynamics software Comsol Multiphysics 3.4 and/or ANSYS CFX 11 respectively. Residence time distribution analysis using a step change of a tracer within the reactor revealed non-ideal fluid distribution characteristics within the reactors. The Brinkman equation was used to model the permeability characteristics with in the chitosan porous structure. Pore size was varied from 10 to 200 microm and the number of pores per unit area was varied from 15 to 1,500 pores/mm(2). Effect of cellular growth and tissue remodeling on flow distribution was also assessed by changing the pore size (85-10 microm) while keeping the number of pores per unit area constant. These results showed significant increase in pressure with reduction in pore size, which could limit the fluid flow and nutrient transport. However, measured pressure drop was marginally higher than the simulation results. Maximum shear stress was similar in both reactors and ranged approximately 0.2-0.3 dynes/cm(2). The simulations were validated experimentally using both a rectangular and circular bioreactor, constructed in-house. Porous structures for the experiments were formed using 0.5% chitosan solution freeze-dried at -80 degrees C, and the pressure drop across the reactor was monitored.

DEFE0023863 Final Report, Technology for GHG Emission Reduction and CostCompetitive MilSpec Jet Fuel Production using CTL

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

Hartvigsen, Joseph J; Dimick, Paul; Laumb, Jason D

Ceramatec Inc, in collaboration with IntraMicron (IM), the Energy & Environmental Research Center (EERC) and Sustainable Energy Solutions, LLC (SES), have completed a three-year research project integrating their respective proprietary technologies in key areas to demonstrate production of a jet fuel from coal and biomass sources. The project goals and objectives were to demonstrate technology capable of producing a “commercially-viable quantity” of jet fuel and make significant progress toward compliance with Section 526 of the Energy Independence and Security Act of 2007 (EISA 2007 §526) lifecycle greenhouse gas (GHG) emissions requirements. The Ceramatec led team completed the demonstration of nominalmore » 2 bbl/day Fischer-Tropsch (FT) synthesis pilot plant design, capable of producing a nominal 1 bbl/day in the Jet-A/JP-8 fraction. This production rate would have a capacity of 1,000 gallons of jet fuel per month and provide the design basis of a 100 bbl/day module producing over 2,000 gallons of jet fuel per day. Co-gasification of coal-biomass blends enables a reduction of lifecycle greenhouse gas emissions from equivalent conventional petroleum derived fuel basis. Due to limits of biomass availability within an economic transportation range, implementation of a significant biomass feed fraction will require smaller plants than current world scale CTL and GTL FT plants. Hence a down-scaleable design is essential. The pilot plant design leverages Intramicron’s MicroFiber Entrapped Catalyst (MFEC) support which increases the catalyst bed thermal conductivity two orders of magnitude, allowing thermal management of the FT reaction exotherm in much larger reactor tubes. In this project, single tube reactors having 4.5 inch outer diameter and multi-tube reactors having 4 inch outer diameters were operated, with productivities as high as 1.5 gallons per day per linear foot of reactor tube. A significant reduction in tube count results from the use of large diameter reactor tubes, with an associated reduction in reactor cost. The pilot plant was designed with provisions for product collection capable of operating with conventional wax producing FT catalysts but was operated with a Chevron hybrid wax-free FT catalyst. Process simplification enabled by elimination of the wax hydrocracking process unit provides economic advantages in scaling to biomass capable plant sizes. Intramicron also provided a sulfur capture system based on their Oxidative Sulfur Removal (OSR) catalyst process. The integrated sulfur removal and FT systems were operated with syngas produced by the Transport Reactor Development Unit (TRDU) gasifier at the University of North Dakota EERC. SES performed modeling of their cryogenic carbon capture process on the energy, cost and CO2 emissions impact of the Coal-biomass synthetic fuel process.« less

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

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

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

Nuclear Technology: Making Informed Decisions.

ERIC Educational Resources Information Center

Altshuler, Kenneth

1989-01-01

Discusses a unit on nuclear technology which is taught in a physics class. Explains the unit design, implementation process, demonstrations used, and topics of discussion that include light and optics, naturally and artificially produced sources of radioactivity, nuclear equations, isotopes and half-lives, and power-generating nuclear reactors.…

Heat stress control in the TMI-2 (Three Mile Island Unit 2) defueling and decontamination activities

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

Schork, J.S.; Parfitt, B.A.

During the initial stages of the Three Mile Island Unit 2 (TMI-2) defueling and decontamination activities for the reactor building, it was realized that the high levels of loose radioactive contamination would require the use of extensive protective clothing by entry personnel. While there was no doubt that layered protective clothing protects workers from becoming contaminated, it was recognized that these same layers of clothing would impose a very significant heat stress burden. To prevent the potentially serious consequences of a severe reaction to heat stress by workers in the hostile environment of the TMI-2 reactor building and yet maintainmore » the reasonable work productivity necessary to perform the recovery adequately, an effective program of controlling worker exposure to heat stress had to be developed. Body-cooling devices produce a flow of cool air, which is introduced close to the skin to remove body heat through convection and increased sweat evaporation. The cooling effect produced by the Vortex tube successfully protected the workers from heat stress, however, there were several logistical and operational problems that hindered extensive use of these devices. The last type of cooling garment examined was the frozen water garment (FWG) developed by Elizier Kamon at the Pennsylvania State University as part of an Electric Power Research Institute research grant. Personal protection, i.e., body cooling, engineering controls, and administrative controls, have been implemented successfully.« less

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.).

78 FR 76328 - Application for Renewal of Special Nuclear Material License SNM-2014 From Tennessee Valley...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-17

... SNM in the form of fully-assembled fuel assemblies that would later form the initial reactor core of WBN2. The SNM in the fuel assemblies is enriched up to 5% in the isotope U-235. The fresh fuel... received the initial core for WBN2. The NRC has not yet issued the OL for the Unit 2 reactor. The...

Integrated head package for top mounted nuclear instrumentation

DOEpatents

Malandra, Louis J.; Hornak, Leonard P.; Meuschke, Robert E.

1993-01-01

A nuclear reactor such as a pressurized water reactor has an integrated head package providing structural support and increasing shielding leading toward the vessel head. A reactor vessel head engages the reactor vessel, and a control rod guide mechanism over the vessel head raises and lowers control rods in certain of the thimble tubes, traversing penetrations in the reactor vessel head, and being coupled to the control rods. An instrumentation tube structure includes instrumentation tubes with sensors movable into certain thimble tubes disposed in the fuel assemblies. Couplings for the sensors also traverse penetrations in the reactor vessel head. A shroud is attached over the reactor vessel head and encloses the control rod guide mechanism and at least a portion of the instrumentation tubes when retracted. The shroud forms a structural element of sufficient strength to support the vessel head, the control rod guide mechanism and the instrumentation tube structure, and includes radiation shielding material for limiting passage of radiation from retracted instrumentation tubes. The shroud is thicker at the bottom adjacent the vessel head, where the more irradiated lower ends of retracted sensors reside. The vessel head, shroud and contents thus can be removed from the reactor as a unit and rested safely and securely on a support.

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

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

Gougar, Hans David

2015-10-01

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

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

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

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

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.

Peach Bottom Atomic Power Station recirc pipe dose rates with zinc injection and condenser replacement

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

DiCello, D.C.; Odell, A.D.; Jackson, T.J.

1995-03-01

Peach Bottom Atomic Power Station (PBAPS) is located near the town of Delta, Pennsylvania, on the west bank of the Susquehanna River. It is situated approximately 20 miles south of Lancaster, Pennsylvania. The site contains two boiling water reactors of General Electric design and each rated at 3,293 megawatts thermal. The units are BWR 4s and went commercial in 1977. There is also a decommissioned high temperature gas-cooled reactor on site, Unit 1. PBAPS Unit 2 recirc pipe was replaced in 1985 and Unit 3 recirc pipes replaced in 1988 with 326 NGSS. The Unit 2 replacement pipe was electropolished,more » and the Unit 3 pipe was electropolished and passivated. The Unit 2 brass condenser was replaced with a Titanium condenser in the first quarter of 1991, and the Unit 3 condenser was replaced in the fourth quarter of 1991. The admiralty brass condensers were the source of natural zinc in both units. Zinc injection was initiated in Unit 2 in May 1991, and in Unit 3 in May 1992. Contact dose rate measurements were made in standard locations on the 28-inch recirc suction and discharge lines to determine the effectiveness of zinc injection and to monitor radiation build-up in the pipe. Additionally, HPGe gamma scans were performed to determine the isotopic composition of the oxide layer inside the pipe. In particular, the specific ({mu}Ci/cm{sup 2}) of Co-60 and Zn-65 were analyzed.« less

Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

PubMed

Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

2012-08-01

The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

Coal conversion processes and analysis methodologies for synthetic fuels production. [technology assessment and economic analysis of reactor design for coal gasification

NASA Technical Reports Server (NTRS)

1979-01-01

Information to identify viable coal gasification and utilization technologies is presented. Analysis capabilities required to support design and implementation of coal based synthetic fuels complexes are identified. The potential market in the Southeast United States for coal based synthetic fuels is investigated. A requirements analysis to identify the types of modeling and analysis capabilities required to conduct and monitor coal gasification project designs is discussed. Models and methodologies to satisfy these requirements are identified and evaluated, and recommendations are developed. Requirements for development of technology and data needed to improve gasification feasibility and economies are examined.

Photoelastic analysis in respect to failure mechanics problems of power plant articles and units

NASA Astrophysics Data System (ADS)

Korikhin, N. V.; Eigenson, S. N.

2009-02-01

The results of strength tests of some critical articles and units of power plants, i.e., a reactor vessel, threaded connection of vessel split, pressure header with straight nipple, turbomachine shaft, and T-weld joint of stator and rotor parts, of turbomachines are presented.

40 CFR 63.12005 - What definitions apply to this subpart?

Code of Federal Regulations, 2012 CFR

2012-07-01

... venting episode that is associated with a single unit operation. A unit operation may have more than one... characterized by a two-step anhydrous polymerization process: the formation of small resin particles in a pre... of the resin particles in a post-polymerization reactor using additional vinyl chloride monomer...

40 CFR 63.1567 - What are my requirements for inorganic HAP emissions from catalytic reforming units?

Code of Federal Regulations, 2013 CFR

2013-07-01

... different reactors in the catalytic reforming unit are regenerated in separate regeneration systems, then these emission limitations apply to each separate regeneration system. These emission limitations apply... catalyst rejuvenation operations during coke burn-off and catalyst regeneration. You can choose from the...

40 CFR 63.1567 - What are my requirements for inorganic HAP emissions from catalytic reforming units?

Code of Federal Regulations, 2012 CFR

2012-07-01

... different reactors in the catalytic reforming unit are regenerated in separate regeneration systems, then these emission limitations apply to each separate regeneration system. These emission limitations apply... catalyst rejuvenation operations during coke burn-off and catalyst regeneration. You can choose from the...

40 CFR 63.1567 - What are my requirements for inorganic HAP emissions from catalytic reforming units?

Code of Federal Regulations, 2014 CFR

2014-07-01

... different reactors in the catalytic reforming unit are regenerated in separate regeneration systems, then these emission limitations apply to each separate regeneration system. These emission limitations apply... catalyst rejuvenation operations during coke burn-off and catalyst regeneration. You can choose from the...

40 CFR 63.107 - Identification of process vents subject to this subpart.

Code of Federal Regulations, 2011 CFR

2011-07-01

... CATEGORIES National Emission Standards for Organic Hazardous Air Pollutants From the Synthetic Organic Chemical Manufacturing Industry § 63.107 Identification of process vents subject to this subpart. (a) The..., distillation unit, or reactor during operation of the chemical manufacturing process unit. (c) The discharge to...

40 CFR 63.107 - Identification of process vents subject to this subpart.

Code of Federal Regulations, 2013 CFR

2013-07-01

... CATEGORIES National Emission Standards for Organic Hazardous Air Pollutants From the Synthetic Organic Chemical Manufacturing Industry § 63.107 Identification of process vents subject to this subpart. (a) The..., distillation unit, or reactor during operation of the chemical manufacturing process unit. (c) The discharge to...

40 CFR 63.107 - Identification of process vents subject to this subpart.

Code of Federal Regulations, 2014 CFR

2014-07-01

... CATEGORIES National Emission Standards for Organic Hazardous Air Pollutants From the Synthetic Organic Chemical Manufacturing Industry § 63.107 Identification of process vents subject to this subpart. (a) The..., distillation unit, or reactor during operation of the chemical manufacturing process unit. (c) The discharge to...

76 FR 388 - Southern Nuclear Operating Company; Vogtle Electric Generating Plant, Unit Nos. 1 and 2; Notice...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-04

... Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant Pressure Boundary,'' and GDC 32, ``Inspection... Operating Company; Vogtle Electric Generating Plant, Unit Nos. 1 and 2; Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration...

78 FR 26662 - Entergy Nuclear Operations, Inc., Indian Point Nuclear Generating Unit No. 3 Extension of Public...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-07

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-286; NRC-2013-0063] Entergy Nuclear Operations, Inc., Indian Point Nuclear Generating Unit No. 3 Extension of Public Comment Period AGENCY: Nuclear Regulatory... FURTHER INFORMATION CONTACT: Douglas V. Pickett, Senior Project Manager, Office of Nuclear Reactor...

40 CFR 63.1579 - What definitions apply to this subpart?

Code of Federal Regulations, 2011 CFR

2011-07-01

... process characterized by continual batch regeneration of catalyst in situ in any one of several reactors... device that treats (in-situ) the catalytic reforming unit recirculating coke burn exhaust gases for acid... or operator's convenience for in situ catalyst regeneration. Sulfur recovery unit means a process...

40 CFR 63.1579 - What definitions apply to this subpart?

Code of Federal Regulations, 2010 CFR

2010-07-01

... process characterized by continual batch regeneration of catalyst in situ in any one of several reactors... device that treats (in-situ) the catalytic reforming unit recirculating coke burn exhaust gases for acid... or operator's convenience for in situ catalyst regeneration. Sulfur recovery unit means a process...

CO{sub 2} Reuse in Petrochemical Facilities

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

Jason Trembly; Brian Turk; Maruthi Pavani

2010-12-31

To address public concerns regarding the consequences of climate change from anthropogenic carbon dioxide (CO{sub 2}) emissions, the U.S. Department of Energy's National Energy Technology Laboratory (DOE/NETL) is actively funding a CO{sub 2} management program to develop technologies capable of mitigating CO{sub 2} emissions from power plant and industrial facilities. Over the past decade, this program has focused on reducing the costs of carbon capture and storage technologies. Recently, DOE/NETL launched an alternative CO{sub 2} mitigation program focused on beneficial CO{sub 2} reuse to support the development of technologies that mitigate emissions by converting CO{sub 2} into valuable chemicals andmore » fuels. RTI, with DOE/NETL support, has been developing an innovative beneficial CO{sub 2} reuse process for converting CO{sub 2} into substitute natural gas (SNG) by using by-product hydrogen (H{sub 2)-containing fuel gas from petrochemical facilities. This process leveraged commercial reactor technology currently used in fluid catalytic crackers in petroleum refining and a novel nickel (Ni)-based catalyst developed by RTI. The goal was to generate an SNG product that meets the pipeline specifications for natural gas, making the SNG product completely compatible with the existing natural gas infrastructure. RTI's technology development efforts focused on demonstrating the technical feasibility of this novel CO{sub 2} reuse process and obtaining the necessary engineering information to design a pilot demonstration unit for converting about 4 tons per day (tons/day) of CO{sub 2} into SNG at a suitable host site. This final report describes the results of the Phase I catalyst and process development efforts. The methanation activity of several commercial fixed-bed catalysts was evaluated under fluidized-bed conditions in a bench-scale reactor to identify catalyst performance targets. RTI developed two fluidizable Ni-based catalyst formulations (Cat-1 and Cat-3) that demonstrated equal or better performance than that of commercial methanation catalysts. The Cat-1 and Cat-3 formulations were successfully scaled up using commercial manufacturing equipment at the Sud-Chemie Inc. pilot-plant facility in Louisville, KY. Pilot transport reactor testing with RTI's Cat-1 formulation at Kellog Brown & Root's Technology Center demonstrated the ability of the process to achieve high single-pass CO{sub 2} conversion. Using information acquired from bench- and pilot-scale testing, a basic engineering design package was prepared for a 4-ton/day CO{sub 2} pilot demonstration unit, including process and instrumentation diagrams, equipment list, control philosophy, and preliminary cost estimate.« less

Influence of COD/sulfate ratios on the integrated reactor system for simultaneous removal of carbon, sulfur and nitrogen.

PubMed

Yuan, Ye; Chen, Chuan; Zhao, Youkang; Wang, Aijie; Sun, Dezhi; Huang, Cong; Liang, Bin; Tan, Wenbo; Xu, Xijun; Zhou, Xu; Lee, Duu-Jung; Ren, Nanqi

2015-01-01

An integrated reactor system was developed for the simultaneous removal of carbon, sulfur and nitrogen from sulfate-laden wastewater and for elemental sulfur (S°) reclamation. The system mainly consisted of an expanded granular sludge bed (EGSB) for sulfate reduction and organic carbon removal (SR-CR), an EGSB for denitrifying sulfide removal (DSR), a biological aerated filter for nitrification and a sedimentation tank for sulfur reclamation. This work investigated the influence of chemical oxygen demand (COD)/sulfate ratios on the performance of the system. Influent sulfate and ammonium were fixed to the level of 600 mg SO(4)(2-) L⁻¹ and 120 mg NH(4)(+) L⁻¹, respectively. Lactate was introduced to generate COD/SO(4)(2-) = 0.5:1, 1:1, 1.5:1, 2:1, 3:1, 3.5:1 and 4:1. The experimental results indicated that sulfate could be efficiently reduced in the SR-CR unit when the COD/SO(4)(2-) ratio was between 1:1 and 3:1, and sulfate reduction was inhibited by the growth of methanogenic bacteria when the COD/SO(4)(2-) ratio was between 3.5:1 and 4:1. Meanwhile, the Org-C/S²⁻/NO(3)(-) ratios affected the S(0) reclamation efficiency in the DSR unit. When the influent COD/SO(4)(2-) ratio was between 1:1 and 3:1, appropriate Org-C/S²⁻/NO(3)(-) ratios could be achieved to obtain a maximum S° recovery in the DSR unit. For the microbial community of the SR-CR unit at different COD/SO(4)(2-) ratios, 16S rRNA gene-based high throughput Illumina MiSeq sequencing was used to analyze the diversity and potential function of the dominant species.

The rate of decay of fresh fission products from a nuclear reactor

NASA Astrophysics Data System (ADS)

Dolan, David J.

Determining the rate of decay of fresh fission products from a nuclear reactor is complex because of the number of isotopes involved, different types of decay, half-lives of the isotopes, and some isotopes decay into other radioactive isotopes. Traditionally, a simplified rule of 7s and 10s is used to determine the dose rate from nuclear weapons and can be to estimate the dose rate from fresh fission products of a nuclear reactor. An experiment was designed to determine the dose rate with respect to time from fresh fission products of a nuclear reactor. The experiment exposed 0.5 grams of unenriched Uranium to a fast and thermal neutron flux from a TRIGA Research Reactor (Lakewood, CO) for ten minutes. The dose rate from the fission products was measured by four Mirion DMC 2000XB electronic personal dosimeters over a period of six days. The resulting dose rate following a rule of 10s: the dose rate of fresh fission products from a nuclear reactor decreases by a factor of 10 for every 10 units of time.

High-Temperature Fluid-Wall Reactor Technology Research, Test and Evaluation Performed at Naval Construction Battalion Center, Gulfport, Mississippi, for the United States Air Force Installation/Restoration Program

DTIC Science & Technology

1988-01-01

the reactor Duties: The Process Engineers rotate with the Lead Operator to monitor the process at the top of the reactor through the site glass...pant cuffs and coverhoods of coveralls, will be attached to gloves, boots and coveralls, using duct tape. * IF AMBIENT WORK STATIONS TEMPERATURE IS...L of the sample fortification solution (Section ýý8) containing 1C 12-2,3,7,8-TCDD at a concentration of 0.5 ng/1,Land C14-2,3,7,8-TCDD at a

Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

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

Sanchez, Rene Gerardo; Hutchinson, Jesson D.; Mcclure, Patrick Ray

2015-08-20

The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

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

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

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

Spent nuclear fuel discharges from US reactors 1992

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

Not Available

1994-05-05

This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactivemore » Waste Management.« less

Sustainability Efficiency Factor: Measuring Sustainability in Advanced Energy Systems through Exergy, Exergoeconomic, Life Cycle, and Economic Analyses

NASA Astrophysics Data System (ADS)

Boldon, Lauren

The Encyclopedia of Life Support Systems defines sustainability or industrial ecology as "the wise use of resources through critical attention to policy, social, economic, technological, and ecological management of natural and human engineered capital so as to promote innovations that assure a higher degree of human needs fulfilment, or life support, across all regions of the world, while at the same time ensuring intergenerational equity" (Encyclopedia of Life Support Systems 1998). Developing and integrating sustainable energy systems to meet growing energy demands is a daunting task. Although the technology to utilize renewable energies is well understood, there are limited locations which are ideally suited for renewable energy development. Even in areas with significant wind or solar availability, backup or redundant energy supplies are still required during periods of low renewable generation. This is precisely why it would be difficult to make the switch directly from fossil fuel to renewable energy generation. A transition period in which a base-load generation supports renewables is required, and nuclear energy suits this need well with its limited life cycle emissions and fuel price stability. Sustainability is achieved by balancing environmental, economic, and social considerations, such that energy is produced without detriment to future generations through loss of resources, harm to the environment, etcetera. In essence, the goal is to provide future generations with the same opportunities to produce energy that the current generation has. This research explores sustainability metrics as they apply to a small modular reactor (SMR)-hydrogen production plant coupled with wind energy and storage technologies to develop a new quantitative sustainability metric, the Sustainability Efficiency Factor (SEF), for comparison of energy systems. The SEF incorporates the three fundamental aspects of sustainability and provides SMR or nuclear hybrid energy system (NHES) reference case studies to (1) introduce sustainability metrics, such as life cycle assessment, (2) demonstrate the methods behind exergy and exergoeconomic analyses, (3) provide an economic analysis of the potential for SMR development from first-of-a-kind (FOAK) to nth-of-a-kind (NOAK), thereby illustrating possible cost reductions and deployment flexibility for SMRs over large conventional nuclear reactors, (4) assess the competitive potential for incorporation of storage and hydrogen production in NHES and in regulated and deregulated electricity markets, (5) compare an SMR-hydrogen production plant to a natural gas steam methane reforming plant using the SEF, and (6) identify and review the social considerations which would support future nuclear development domestically and abroad, such as public and political/regulatory needs and challenges. The Global Warming Potential (GWP) for the SMR (300 MWth)-wind (60 MWe)-high temperature steam electrolysis (200 tons Hydrogen per day) system was calculated as approximately 874 g CO2-equivalent as part of the life cycle assessment. This is 92.6% less than the GWP estimated for steam methane reforming production of hydrogen by Spath and Mann. The unit exergetic and exergoeconomic costs were determined for each flow within the NHES system as part of the exergy/exergoeconomic cost analyses. The unit exergetic cost is lower for components yielding more meaningful work like the one exiting the SMR with a unit exergetic cost of 1.075 MW/MW. In comparison, the flow exiting the turbine has a very high unit exergetic cost of 15.31, as most of the useful work was already removed through the turning of the generator/compressor shaft. In a similar manner, the high unit exergoeconomic cost of 12.45/MW*sec is observed for the return flow to the reactors, because there is very little exergy present. The first and second law efficiencies and the exergoeconomic factors were also determined over several cases. For the first or base SMR case, first and second law efficiencies of 81.5% and 93.3% were observed respectively. With an increase in reactor outlet temperature of only 20°C, both the SMR efficiencies increased, while the exergoeconomic factor decreased by 0.2%. As part of the SMR economic analysis, specific capital and total capital investment costs (TCIC) were determined in addition to conditional effects on the net present value (NPV), levelized cost of electricity (LCOE), and payback periods. For a 1260 MWe FOAK multi-module SMR site with 7 modules, the specific capital costs were 27-38% higher than that of a 1260 MWe single large reactor site. A NOAK site, on the other hand, may be 19% lower to 18% higher than the large reactor site, demonstrating that it may break even or be even more economical in average or favorable market conditions. The NOAK TCIC for single and multi-module SMR sites were determined to be 914-1,230 million and 660-967 million per module, respectively, reflecting the substantial savings incurred with sites designed for and deployed with multiple modules. For the same NOAK 7-unit multi-module site, the LCOE was calculated as 67-84/MWh, which is slightly less than that of the conventional large reactor LCOE of 89/MWh with a weighted average cost of capital of 10%, a 50%-50% share of debt and equity, and a corporate tax rate of 35%. The payback period for the SMR site, however, is 4 years longer. Construction delays were also analyzed to compare the SMR and large reactor sites, demonstrating the SMR NPV and LCOE are less sensitive to delays. For a 3 year delay, the SMR NPV decreased by 22%, while the large reactor NPV decreased by 34.1%. Similarly the SMR and large reactor LCOEs increased by 7.8% and 8.1%, respectively. An NHES case with hydrogen production and storage was performed, illustrating how the profit share of revenue is improved with the addition of hydrogen production. Although the costs are increased with the addition, 78% of the hydrogen revenue is profit, while only 50% of the electricity generation revenue is profit. A second NHES case study was analyzed to assess the NPV, LCOE, and payback differences in deregulated and regulated electricity markets. For a 60 year lifetime, Case C (with nuclear, wind, and hydrogen production) is economical in the deregulated market with an NPV of 66.3 million and a payback period of 10 years, but not in the regulated one with an NPV of approximately -115.3 million and a payback period of 11 years. With either market type, the plants levelized costs remain $82.82/MWh, which is still reasonable with respect to prior LCOE values determined for SMR and large reactor sites. Utilizing all the methodology and results obtained and presented in this thesis, the SEF may be calculated. The NHES SEF was determined to be 18.3% higher than that of natural gas steam methane reforming, illustrating a higher level of sustainability. The SEF quantitatively uses the exergoeconomic cost and irreversibilities obtained from the exergy analysis, the GWP obtained from the life cycle assessment and costs/fees associated with emissions and pollutants, and relevant economic data obtained from an economic analysis. This reflects the environmental, socio-political, and economic pillars of sustainability.

An experimental study of ammonia borane based hydrogen storage systems

NASA Astrophysics Data System (ADS)

Deshpande, Kedaresh A.

2011-12-01

Hydrogen is a promising fuel for the future, capable of meeting the demands of energy storage and low pollutant emission. Chemical hydrides are potential candidates for chemical hydrogen storage, especially for automobile applications. Ammonia borane (AB) is a chemical hydride being investigated widely for its potential to realize the hydrogen economy. In this work, the yield of hydrogen obtained during neat AB thermolysis was quantified using two reactor systems. First, an oil bath heated glass reactor system was used with AB batches of 0.13 gram (+/- 0.001 gram). The rates of hydrogen generation were measured. Based on these experimental data, an electrically heated steel reactor system was designed and constructed to handle up to 2 grams of AB per batch. A majority of components were made of stainless-steel. The system consisted of an AB reservoir and feeder, a heated reactor, a gas processing unit and a system control and monitoring unit. An electronic data acquisition system was used to record experimental data. The performance of the steel reactor system was evaluated experimentally through batch reactions of 30 minutes each, for reaction temperatures in the range from 373 K to 430 K. The experimental data showed exothermic decomposition of AB accompanied by rapid generation of hydrogen during the initial period of the reaction. 90% of the hydrogen was generated during the initial 120 seconds after addition of AB to the reactor. At 430 K, the reaction produced 12 wt.% of hydrogen. The heat diffusion in the reactor system and the process of exothermic decomposition of AB were coupled in a two-dimensional model. Neat AB thermolysis was modeled as a global first order reactions based on Arrhenius theory. The values of equation constants were derived from curve fit of experimental data. The pre-exponential constant and the activation energy were estimated to be 4 s-1 (+/- 0.4 s-1) and 13000 J mol -1 s-1 (+/- 1050 J mol-1 s -1) respectively. The model was solved in COMSOL Multiphysics. The model was capable of simulating the transient response of the system and captured the observed trends such as the decrease in reactor temperature upon addition of AB and exothermic decomposition.

A document review to characterize Atomic International SNAP fuels shipped to INEL 1966--1973

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

Wahnschaffe, S.D.; Lords, R.E.; Kneff, D.W.

1995-09-01

This report provides the results of a document search and review study to obtain information on the spent fuels for the following six Nuclear Auxiliary Power (SNAP) reactor cores now stored at the Idaho National Engineering Laboratory (INEL): SNAP-2 Experimental Reactor, SNAP-2 Development Reactor, SNAP-10A Ground Test Reactor, SNAP-8 Experimental Reactor, SNAP-8 Development Reactor, and Shield Test Reactor. The report also covers documentation on SNAP fuel materials from four in-pile materials tests: NAA-82-1, NAA-115-2, NAA-117-1, and NAA-121. Pieces of these fuel materials are also stored at INEL as part of the SNAP fuel shipments.

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.

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

Design of a heatpipe-cooled Mars-surface fission reactor

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.; Reid, Robert S.; Lipinski, Ronald J.; Wright, Steven A.; Talandis, Regina A.

2002-01-01

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars-surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heatpipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed-which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heatpipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heatpipes. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to the Stirling engine. This paper describes the attributes, specifications, and performance of a 15-kWt HOMER reactor. .

SHARP pre-release v1.0 - Current Status and Documentation

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

Mahadevan, Vijay S.; Rahaman, Ronald O.

The NEAMS Reactor Product Line effort aims to develop an integrated multiphysics simulation capability for the design and analysis of future generations of nuclear power plants. The Reactor Product Line code suite’s multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. In this report, building on a several previous report issued in September 2014, we describe our continued efforts to integrate thermal/hydraulics, neutronics, and structural mechanics modeling codes to perform coupled analysis of a representativemore » fast sodium-cooled reactor core in preparation for a unified release of the toolkit. The work reported in the current document covers the software engineering aspects of managing the entire stack of components in the SHARP toolkit and the continuous integration efforts ongoing to prepare a release candidate for interested reactor analysis users. Here we report on the continued integration effort of PROTEUS/Nek5000 and Diablo into the NEAMS framework and the software processes that enable users to utilize the capabilities without losing scientific productivity. Due to the complexity of the individual modules and their necessary/optional dependency library chain, we focus on the configuration and build aspects for the SHARP toolkit, which includes capability to autodownload dependencies and configure/install with optimal flags in an architecture-aware fashion. Such complexity is untenable without strong software engineering processes such as source management, source control, change reviews, unit tests, integration tests and continuous test suites. Details on these processes are provided in the report as a building step for a SHARP user guide that will accompany the first release, expected by Mar 2016.« less

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

Development of Nuclear Renewable Oil Shale Systems for Flexible Electricity and Reduced Fossil Fuel Emissions

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

Daniel Curtis; Charles Forsberg; Humberto Garcia

2015-05-01

We propose the development of Nuclear Renewable Oil Shale Systems (NROSS) in northern Europe, China, and the western United States to provide large supplies of flexible, dispatchable, very-low-carbon electricity and fossil fuel production with reduced CO2 emissions. NROSS are a class of large hybrid energy systems in which base-load nuclear reactors provide the primary energy used to produce shale oil from kerogen deposits and simultaneously provide flexible, dispatchable, very-low-carbon electricity to the grid. Kerogen is solid organic matter trapped in sedimentary shale, and large reserves of this resource, called oil shale, are found in northern Europe, China, and the westernmore » United States. NROSS couples electricity generation and transportation fuel production in a single operation, reduces lifecycle carbon emissions from the fuel produced, improves revenue for the nuclear plant, and enables a major shift toward a very-low-carbon electricity grid. NROSS will require a significant development effort in the United States, where kerogen resources have never been developed on a large scale. In Europe, however, nuclear plants have been used for process heat delivery (district heating), and kerogen use is familiar in certain countries. Europe, China, and the United States all have the opportunity to use large scale NROSS development to enable major growth in renewable generation and either substantially reduce or eliminate their dependence on foreign fossil fuel supplies, accelerating their transitions to cleaner, more efficient, and more reliable energy systems.« less

Performance and microbial community composition dynamics of aerobic granular sludge from sequencing batch bubble column reactors operated at 20 degrees C, 30 degrees C, and 35 degrees C.

PubMed

Ebrahimi, Sirous; Gabus, Sébastien; Rohrbach-Brandt, Emmanuelle; Hosseini, Maryam; Rossi, Pierre; Maillard, Julien; Holliger, Christof

2010-07-01

Two bubble column sequencing batch reactors fed with an artificial wastewater were operated at 20 degrees C, 30 degrees C, and 35 degrees C. In a first stage, stable granules were obtained at 20 degrees C, whereas fluffy structures were observed at 30 degrees C. Molecular analysis revealed high abundance of the operational taxonomic unit 208 (OTU 208) affiliating with filamentous bacteria Leptothrix spp. at 30 degrees C, an OTU much less abundant at 20 degrees C. The granular sludge obtained at 20 degrees C was used for the second stage during which one reactor was maintained at 20 degrees C and the second operated at 30 degrees C and 35 degrees C after prior gradual increase of temperature. Aerobic granular sludge with similar physical properties developed in both reactors but it had different nutrient elimination performances and microbial communities. At 20 degrees C, acetate was consumed during anaerobic feeding, and biological phosphorous removal was observed when Rhodocyclaceae-affiliating OTU 214 was present. At 30 degrees C and 35 degrees C, acetate was mainly consumed during aeration and phosphorous removal was insignificant. OTU 214 was almost absent but the Gammaproteobacteria-affiliating OTU 239 was more abundant than at 20 degrees C. Aerobic granular sludge at all temperatures contained abundantly the OTUs 224 and 289 affiliating with Sphingomonadaceae indicating that this bacterial family played an important role in maintaining stable granular structures.

Enhancement of operating flux in a membrane bio-reactor coupled with a mechanical sieve unit.

PubMed

Park, Seongjun; Yeon, Kyung-Min; Moon, Seheum; Kim, Jong-Oh

2018-01-01

Filtration flux is one of the key factors in regulating the performance of membrane bio-reactors (MBRs) for wastewater treatment. In this study, we explore the effectiveness of a mechanical sieve unit for effective flux enhancement through retardation of the fouling effect in a modified MBR system (SiMBR). In brief, the coarse sieve unit having 100 μm and 50 μm permits small size microorganism flocs to adjust the biomass concentration from the suspended basin to the membrane basin. As a result, the reduced biofouling effect due to the lowered biomass concentration from 7800 mg/L to 2400 mg/L, enables higher flux through the membrane. Biomass rejection rate of the sieve is identified to be the crucial design parameter for the flux enhancement through the incorporation of numerical simulations and operating critical-flux measurement in a batch reactor. Then, the sieve unit is prepared for 10 L lab-scale continuous SiMBR based on the correlation between sieve pore size and biomass rejection characteristics. During continuous operation of lab-scale SiMBR, biomass concentration is maintained with a higher biomass concentration in the aerobic basin (7400 mg/L) than that in the membrane basin (2400 mg/L). In addition, the SiMBR operations are conducted using three different commercial hollow fiber membranes to compare the permeability to that of conventional MBR operations. For all cases, the modified MBR having a sieve unit clearly results in enhanced permeability. These results successfully validate that SiMBR can effectively improve flux through direct reduction of biomass concentration. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

Overview of the U.S. DOE Accident Tolerant Fuel Development Program

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

Jon Carmack; Frank Goldner; Shannon M. Bragg-Sitton

2013-09-01

The United States Fuel Cycle Research and Development Advanced Fuels Campaign has been given the responsibility to conduct research and development on enhanced accident tolerant fuels with the goal of performing a lead test assembly or lead test rod irradiation in a commercial reactor by 2022. The Advanced Fuels Campaign has defined fuels with enhanced accident tolerance as those that, in comparison with the standard UO2-Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining ormore » improving the fuel performance during normal operations and operational transients, as well as design-basis and beyond design-basis events. This paper provides an overview of the FCRD Accident Tolerant Fuel program. The ATF attributes will be presented and discussed. Attributes identified as potentially important to enhance accident tolerance include reduced hydrogen generation (resulting from cladding oxidation), enhanced fission product retention under severe accident conditions, reduced cladding reaction with high-temperature steam, and improved fuel-cladding interaction for enhanced performance under extreme conditions. To demonstrate the enhanced accident tolerance of candidate fuel designs, metrics must be developed and evaluated using a combination of design features for a given LWR design, potential improvements to that design, and the design of an advanced fuel/cladding system. The aforementioned attributes provide qualitative guidance for parameters that will be considered for fuels with enhanced accident tolerance. It may be unnecessary to improve in all attributes and it is likely that some attributes or combination of attributes provide meaningful gains in accident tolerance, while others may provide only marginal benefits. Thus, an initial step in program implementation will be the development of quantitative metrics. A companion paper in these proceedings provides an update on the status of establishing these quantitative metrics for accident tolerant LWR fuel.1 The United States FCRD Advanced Fuels Campaign has embarked on an aggressive schedule for development of enhanced accident tolerant LWR fuels. The goal of developing such a fuel system that can be deployed in the U.S. LWR fleet in the next 10 to 20 years supports the sustainability of clean nuclear power generation in the United States.« less

Anaerobic Treatment of Palm Oil Mill Effluent in Pilot-Scale Anaerobic EGSB Reactor

PubMed Central

Mahmood, Qaisar; Qiu, Jiang-Ping; Li, Yin-Sheng; Chang, Yoon-Seong; Li, Xu-Dong

2015-01-01

Large volumes of untreated palm oil mill effluent (POME) pose threat to aquatic environment due to the presence of very high organic content. The present investigation involved two pilot-scale anaerobic expanded granular sludge bed (EGSB) reactors, continuously operated for 1 year to treat POME. Setting HRT at 9.8 d, the anaerobic EGSB reactors reduced COD from 71179 mg/L to 12341 mg/L and recycled half of sludge by a dissolved air flotation (DAF). The average effluent COD was 3587 mg/L with the consistent COD removal efficiency of 94.89%. Adding cationic polymer (PAM) dose of 30 mg/L to DAF unit and recycling its half of sludge caused granulation of anaerobic sludge. Bacilli and small coccid bacteria were the dominant microbial species of the reactor. The reactor produced 27.65 m3 of biogas per m3 of POME which was utilized for electricity generation. PMID:26167485

Information basis for developing comprehensive waste management system-US-Japan joint nuclear energy action plan waste management working group phase I report.

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

Nutt, M.; Nuclear Engineering Division

2010-05-25

The activity of Phase I of the Waste Management Working Group under the United States - Japan Joint Nuclear Energy Action Plan started in 2007. The US-Japan JNEAP is a bilateral collaborative framework to support the global implementation of safe, secure, and sustainable, nuclear fuel cycles (referred to in this document as fuel cycles). The Waste Management Working Group was established by strong interest of both parties, which arise from the recognition that development and optimization of waste management and disposal system(s) are central issues of the present and future nuclear fuel cycles. This report summarizes the activity of themore » Waste Management Working Group that focused on consolidation of the existing technical basis between the U.S. and Japan and the joint development of a plan for future collaborative activities. Firstly, the political/regulatory frameworks related to nuclear fuel cycles in both countries were reviewed. The various advanced fuel cycle scenarios that have been considered in both countries were then surveyed and summarized. The working group established the working reference scenario for the future cooperative activity that corresponds to a fuel cycle scenario being considered both in Japan and the U.S. This working scenario involves transitioning from a once-through fuel cycle utilizing light water reactors to a one-pass uranium-plutonium fuel recycle in light water reactors to a combination of light water reactors and fast reactors with plutonium, uranium, and minor actinide recycle, ultimately concluding with multiple recycle passes primarily using fast reactors. Considering the scenario, current and future expected waste streams, treatment and inventory were discussed, and the relevant information was summarized. Second, the waste management/disposal system optimization was discussed. Repository system concepts were reviewed, repository design concepts for the various classifications of nuclear waste were summarized, and the factors to consider in repository design and optimization were then discussed. Japan is considering various alternatives and options for the geologic disposal facility and the framework for future analysis of repository concepts was discussed. Regarding the advanced waste and storage form development, waste form technologies developed in both countries were surveyed and compared. Potential collaboration areas and activities were next identified. Disposal system optimization processes and techniques were reviewed, and factors to consider in future repository design optimization activities were also discussed. Then the potential collaboration areas and activities related to the optimization problem were extracted.« less

The Best Defense: Making Maximum Sense of Minimum Deterrence

DTIC Science & Technology

2011-06-01

uranium fuel cycles and has unmatched experience in the thorium fuel cycle.25 Published sources claim India produces between 20 and 40kg of plutonium...nuclear energy was moderate at best. Pakistan‘s first reactor , which it received from the United States, did not become operational until 1965.4...In 1974 Pakistan signed an agreement with France to supply a reprocessing plant for extracting plutonium from spent fuel from power reactors

Calculation of the temperature in the container unit with a modified design for the production of {sup 99}Mo at the VVR-Ts research reactor facility (IVV.10M)

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

Kazantsev, A. A., E-mail: kazantsevanatoly@gmail.com; Sergeev, V. V.; Kochnov, O. Yu.

The temperature regime is calculated for two different designs of containers with uranium-bearing material for the upgraded VVR-Ts research reactor facility (IVV.10M). The containers are to be used in the production of {sup 99}Mo. It is demonstrated that the modification of the container design leads to a considerable temperature reduction and an increase in the near-wall boiling margin and allows one to raise the amount of material loaded into the container. The calculations were conducted using the international thermohydraulic contour code TRAC intended to analyze the technical safety of water-cooled nuclear power units.

Nuclear thermionic converter. [tungsten-thorium oxide rods

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Mondt, J. F. (Inventor)

1977-01-01

Efficient nuclear reactor thermionic converter units are described which can be constructed at low cost and assembled in a reactor which requires a minimum of fuel. Each converter unit utilizes an emitter rod with a fluted exterior, several fuel passages located in the bulges that are formed in the rod between the flutes, and a collector receiving passage formed through the center of the rod. An array of rods is closely packed in an interfitting arrangement, with the bulges of the rods received in the recesses formed between the bulges of other rods, thereby closely packing the nuclear fuel. The rods are constructed of a mixture of tungsten and thorium oxide to provide high power output, high efficiency, high strength, and good machinability.

Regeneration of oxygen from carbon dioxide and water.

NASA Technical Reports Server (NTRS)

Weissbart, J.; Smart, W. H.; Wydeven, T.

1972-01-01

In a closed ecological system it is necessary to reclaim most of the oxygen required for breathing from respired carbon dioxide and the remainder from waste water. One of the advanced physicochemical systems being developed for generating oxygen in manned spacecraft is the solid electrolyte-electrolysis system. The solid electrolyte system consists of two basic units, an electrolyzer and a carbon monoxide disproportionator. The electrolyzer can reclaim oxygen from both carbon dioxide and water. Electrolyzer preparation and assembly are discussed together with questions of reactor design and electrolyzer performance data.

A method for predicting gamma-radiation dose rates in the premises of the multiple forced circulation circuit of an RBMK-1000 reactor from the data of chemical and radiospectrometric monitoring of coolant

NASA Astrophysics Data System (ADS)

Chernikov, O. G.; Kovalev, S. M.; Epikhin, A. I.; Kozlov, E. P.; Petrov, S. I.; Rodionov, Yu. A.; Kritskii, V. G.; Styazhkin, P. S.

2009-05-01

A mathematical model for predicting gamma-radiation dose rate in the premises of the multiple forced circulation circuit is developed, which is based on the data of water chemistry in the circuit, radionuclide composition of coolant, and hydraulic characteristics of equipment. Data on approbation of the model are presented that were obtained during the shutdown of power units at the Leningrad and Smolensk nuclear power stations.

Expert system for maintenance management of a boiling water reactor power plant

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

Hong Shen; Liou, L.W.; Levine, S.

1992-01-01

An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, whichmore » must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components.« less

Development of NSSS Thermal-Hydraulic Model for KNPEC-2 Simulator Using the Best-Estimate Code RETRAN-3D

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

Kim, Kyung-Doo; Jeong, Jae-Jun; Lee, Seung-Wook

The Nuclear Steam Supply System (NSSS) thermal-hydraulic model adopted in the Korea Nuclear Plant Education Center (KNPEC)-2 simulator was provided in the early 1980s. The reference plant for KNPEC-2 is the Yong Gwang Nuclear Unit 1, which is a Westinghouse-type 3-loop, 950 MW(electric) pressurized water reactor. Because of the limited computational capability at that time, it uses overly simplified physical models and assumptions for a real-time simulation of NSSS thermal-hydraulic transients. This may entail inaccurate results and thus, the possibility of so-called ''negative training,'' especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developedmore » a realistic NSSS thermal-hydraulic program (named ARTS code) based on the best-estimate code RETRAN-3D. The systematic assessment of ARTS has been conducted by both a stand-alone test and an integrated test in the simulator environment. The non-integrated stand-alone test (NIST) results were reasonable in terms of accuracy, real-time simulation capability, and robustness. After successful completion of the NIST, ARTS was integrated with a 3-D reactor kinetics model and other system models. The site acceptance test (SAT) has been completed successively and confirmed to comply with the ANSI/ANS-3.5-1998 simulator software performance criteria. This paper presents our efforts for the ARTS development and some test results of the NIST and SAT.« less

US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations

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

Amway, P.; Andrews, N.; Bixby, Willis

Although it is clear that the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limitedmore » full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings (TEPCO) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document reports recent results from the US Forensics Effort to use information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. This Forensics Effort, which is sponsored by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of US experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO information from Daiichi that address these needs. Examples presented in this report demonstrate that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities.« less

Status and progress of the RERTR program in the year 2003.

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

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to define the feasibility of converting each Russian-designed research and test reactor to either fuel type. The plan for the Accelerated RERTR Program is structured to achieve LEU conversion of all HEU research reactors supplied by the United States and Russia during the next nine years. This effort will address, in addition to the fuel development and qualification, the analyses and performance/economic/safety evaluations needed to implement the conversions. In combination with this over-arching goal, the RERTR program plans to achieve at the earliest possible date qualification of LEU U-Mo dispersion fuels with uranium densities of 6 g/cm{sup 3} and 7 g/cm{sup 3}. Reactors currently using or planning to use LEU silicide fuel will rely on this fuel after termination of the FRRSNFA program, because it is acceptable to COGEMA for reprocessing. Qualification of LEU U-Mo dispersion fuels has suffered some unavoidable delays but, to accelerate it as much as possible, the RERTR program, the French CEA, and the Australian ANSTO have agreed to jointly pursue a two-element qualification test of LEU U-Mo dispersion fuel with uranium density of 7.0 g/cm{sup 3} to be performed in the Osiris reactor during 2004. The RERTR program also intends to eliminate all obstacles to the utilization of LEU in targets for isotope production, so that this important function can be performed without the need for weapons-grade materials. All of us, working together as we have for many years, can ensure that all these goals will be achieved. By promoting the efficiency and safety of research reactors while eliminating the traffic in weapons-grade uranium, we can prevent the possibility that some of this material might fall in the wrong hands. Few causes can be more deserving of our joint efforts.« less

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. .

Markov Model of Accident Progression at Fukushima Daiichi

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

Cuadra A.; Bari R.; Cheng, L-Y

2012-11-11

On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unitmore » 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.« less

Nuclear power: levels of safety.

PubMed

Lidsky, L M

1988-02-01

The rise and fall of the nuclear power industry in the United States is a well-documented story with enough socio-technological conflict to fill dozens of scholarly, and not so scholarly, books. Whatever the reasons for the situation we are now in, and no matter how we apportion the blame, the ultimate choice of whether to use nuclear power in this country is made by the utilities and by the public. Their choices are, finally, based on some form of risk-benefit analysis. Such analysis is done in well-documented and apparently logical form by the utilities and in a rather more inchoate but not necessarily less accurate form by the public. Nuclear power has failed in the United States because both the real and perceived risks outweigh the potential benefits. The national decision not to rely upon nuclear power in its present form is not an irrational one. A wide ranging public balancing of risk and benefit requires a classification of risk which is clear and believable for the public to be able to assess the risks associated with given technological structures. The qualitative four-level safety ladder provides such a framework. Nuclear reactors have been designed which fit clearly and demonstrably into each of the possible qualitative safety levels. Surprisingly, it appears that safer may also mean cheaper. The intellectual and technical prerequisites are in hand for an important national decision. Deployment of a qualitatively different second generation of nuclear reactors can have important benefits for the United States. Surprisingly, it may well be the "nuclear establishment" itself, with enormous investments of money and pride in the existing nuclear systems, that rejects second generation reactors. It may be that we will not have a second generation of reactors until the first generation of nuclear engineers and nuclear power advocates has retired.

Production of aluminum-silicon alloy and ferrosilicon and commercial purity aluminum by the direct reduction process. Second interim technical report, Phase B, December 1, 1978-February 28, 1979

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

Bruno, M.J.

1979-03-01

Experimental runs were made to determine the effect of a cooler product reservoir on metal alloy yield and recovery. The reservoir temperature had no significant effect. Difficulties were experienced with operation of an oxygen injected bench scale reactor. Many tests were terminated by burden bridging or flooding of the oxygen tuyeres with metal and slag. Runs were made in which refluxing vapors were condensed in a liquid slag. The addition of CaO decreased the tendency for formation of thick, strong burden bridges but did not completely eliminate bridging. Reduction of flame temperatures did not affect the volatilization rate in themore » bench reactor. Operation of VSR-1 pilot reactor with O injection was achieved after resolving reactor shell leakage problems, by replacing the permeable ceramic shell with impermeable fused silica. Various combustion parameters were investigated, including coke size, burden height and oxygen flow rate. Steady state operation of the oxygen-coke system was attained with smooth burden movement and a 2000/sup 0/C bed temperature in the raceway vicinity. To further reduce heat losses from the raceway area. VSR-1 was redesigned to facilitate locating an induction coil below the oxygen inlets. Further evaluation of effects of impurities on alloy purification in the bench scale unit indicated a 50% decrease in product yield for starting charges containing Fe greater than 5%. Site installation for the entire alloy purification complex was completed. Operations were continued in the bench scale units to obtain design information for the pilot commercial grade Al purification unit. Procurement of construction material was established.« less

A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

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

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2015-03-01

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoidmore » overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety system components such as the safety relief valve (SRV), the RCIC system, the wet well, and the dry well. The results show reasonable system behaviors while exhibiting rich dynamics such as variable flow rates through RCIC turbine and pump during the SBO transient. The model has the potential to resolve the Fukushima RCIC mystery after adding the off-design two-phase turbine operation model and other additional improvements.« less

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.

Nitrate removal with lateral flow sulphur autotrophic denitrification reactor.

PubMed

Lv, Xiaomei; Shao, Mingfei; Li, Ji; Xie, Chuanbo

2014-01-01

An innovative lateral flow sulphur autotrophic denitrification (LFSAD) reactor was developed in this study; the treatment performance was evaluated and compared with traditional sulphur/limestone autotrophic denitrification (SLAD) reactor. Results showed that nitrite accumulation in the LFSAD reactor was less than 1.0 mg/L during the whole operation. Denitrification rate increased with the increased initial alkalinity and was approaching saturation when initial alkalinity exceeded 2.5 times the theoretical value. Higher influent nitrate concentration could facilitate nitrate removal capacity. In addition, denitrification efficiency could be promoted under an appropriate reflux ratio, and the highest nitrate removal percentage was achieved under reflux ratio of 200%, increased by 23.8% than that without reflux. Running resistance was only about 1/9 of that in SLAD reactor with equal amount of nitrate removed, which was the prominent excellence of the new reactor. In short, this study indicated that the developed reactor was feasible for nitrate removal from waters with lower concentrations, including contaminated surface water, groundwater or secondary effluent of municipal wastewater treatment with fairly low running resistance. The innovation in reactor design in this study may bring forth new ideas of reactor development of sulphur autotrophic denitrification for nitrate-contaminated water treatment.

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.

40 CFR 63.1567 - What are my requirements for inorganic HAP emissions from catalytic reforming units?

Code of Federal Regulations, 2011 CFR

2011-07-01

... reactors in the catalytic reforming unit are regenerated in separate regeneration systems, then these emission limitations apply to each separate regeneration system. These emission limitations apply to... rejuvenation operations during coke burn-off and catalyst regeneration. You can choose from the two options in...

76 FR 30204 - Exelon Nuclear, Dresden Nuclear Power Station, Unit 1; Exemption From Certain Security Requirements

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-24

... Power Station, Unit 1; Exemption From Certain Security Requirements 1.0 Background Exelon Nuclear is the licensee and holder of Facility Operating License No. DPR-2 issued for Dresden Nuclear Power Station (DNPS... protection of licensed activities in nuclear power reactors against radiological sabotage,'' paragraph (b)(1...

75 FR 71152 - Southern California Edison; San Onofre Nuclear Generating Station, Unit 2 and Unit 3; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-22

... nuclear power plants, but noted that the Commission's regulations provide mechanisms for individual.... Borchardt (NRC) to M. S. Fertel (Nuclear Energy Institute) dated June 4, 2009. The licensee's request for an... effect. The facility consists of two pressurized-water reactors located in San Diego County, California...

Assessment for advanced fuel cycle options in CANDU

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

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Kulesza, Joel A.; Roudén, Jenny; Green, Eva-Lena

2016-02-01

This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ˜ 25 effective full power years) of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV) fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M) and calculated (C) results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE)/C ratios of 1.10 for both neutron (E >1.0 MeV) flux and iron atom displacement rate.

Etching Rate of Silicon Dioxide Using Chlorine Trifluoride Gas

NASA Astrophysics Data System (ADS)

Miura, Yutaka; Kasahara, Yu; Habuka, Hitoshi; Takechi, Naoto; Fukae, Katsuya

2009-02-01

The etching rate behavior of silicon dioxide (SiO2, fused silica) using chlorine trifluoride (ClF3) gas is studied at substrate temperatures between 573 and 1273 K at atmospheric pressure in a horizontal cold-wall reactor. The etching rate increases with the ClF3 gas concentration, and the overall reaction is recognized to be of the first order. The change of the etching rate with increasing substrate temperature is nonlinear, and the etching rate tends to approach a constant value at temperatures exceeding 1173 K. The overall rate constant is estimated by numerical calculation, taking into account the transport phenomena in the reactor, including the chemical reaction at the substrate surface. The activation energy obtained in this study is 45.8 kJ mol-1, and the rate constant is consistent with the measured etching rate behavior. A reactor system in which there is minimum etching of the fused silica chamber by ClF3 gas can be achieved using an IR lamp heating unit and a chamber cooling unit to maintain a sufficiently low temperature of the chamber wall.

Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems

NASA Technical Reports Server (NTRS)

Harty, Richard B.; Durand, Richard E.

1993-01-01

An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage.