The RERTR Program status and progress

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

Travelli, A.

1995-12-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1995 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1994. The revelation that Iraq was on the verge of developing a nuclear weapon at the time of the Gulf War, and that it was planning to do so by extracting HEU from the fuel of its research reactors, has given new impetus and urgency to the RERTR commitment of eliminating HEU use in research and test reactors worldwide.more » Development of advanced LEU research reactor fuels is scheduled to begin in October 1995. The Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels, is now in operation. A Statement of Intent was signed by high US and Chinese officials, endorsing cooperative activities between the RERTR program and Chinese laboratories involved in similar activities. Joint studies of LEU technical feasibility were completed for the SAFARI-I reactor in South Africa and for the ANS reactor in the US. A new study has been initiated for the FRM-II reactor in Germany. Significant progress was made on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU. A cooperation agreements is in place with the Indonesian BATAN. The first prototypical irradiation of an LEU metal-foil target for {sup 99}Mo production was accomplished in Indonesia. The TR-2 reactor, in Turkey, began conversion. SAPHIR, in Switzerland, was shut down. LEU fuel fabrication has begun for the conversion of two more US reactors. Twelve foreign reactors and nine domestic reactors have been fully converted. Approximately 60 % of the work required to eliminate the use of HEU in US-supplied research reactors has been accomplished.« less

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

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

Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)« less

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Clifton, W. B.; Hickman, Robert R.; Wang, Ten-See; Dobson, Christopher C.

2011-01-01

An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

Development of the reactor antineutrino detection technology within the iDream project

NASA Astrophysics Data System (ADS)

Gromov, M.; Kuznetsov, D.; Murchenko, A.; Novikova, G.; Obinyakov, B.; Oralbaev, A.; Plakitina, K.; Skorokhvatov, M.; Sukhotin, S.; Chepurnov, A.; Etenko, A.

2017-12-01

The iDREAM (industrial Detector for reactor antineutrino monitoring) project is aimed at remote monitoring of the operating modes of the atomic reactor on nuclear power plant to ensure a technical support of IAEA non-proliferation safeguards. The detector is a scintillator spectrometer. The sensitive volume (target) is filled with a liquid organic scintillator based on linear alkylbenzene where reactor antineutrinos will be detected via inverse beta-decay reaction. We present first results of laboratory tests after physical launch. The detector was deployed at sea level without background shielding. The number of calibrations with radioactive sources was conducted. All data were obtained by means of a slow control system which was put into operation.

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

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

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

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.

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

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

T. R. Allen; J. B. Benson; J. A. Foster

2009-05-01

To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities ismore » granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team projects and faculty/staff exchanges. In June of 2008, the first week-long ATR NSUF Summer Session was attended by 68 students, university faculty and industry representatives. The Summer Session featured presentations by 19 technical experts from across the country and covered topics including irradiation damage mechanisms, degradation of reactor materials, LWR and gas reactor fuels, and non-destructive evaluation. High impact research results from leveraging the entire research infrastructure, including universities, industry, small business, and the national laboratories. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. Current partner facilities include the MIT Reactor, the University of Michigan Irradiated Materials Testing Laboratory, the University of Wisconsin Characterization Laboratory, and the University of Nevada, Las Vegas transmission Electron Microscope User Facility. Needs for irradiation of material specimens at tightly controlled temperatures are being met by dedication of a large in-pile pressurized water loop facility for use by ATR NSUF users. Several environmental mechanical testing systems are under construction to determine crack growth rates and fracture toughness on irradiated test systems.« less

REACTOR PHYSICS QUARTERLY REPORT JANUARY, FEBRUARY, MARCH 1970

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

Schmid, L. C.; Clayton, E. D.; Heineman, R. E.

1970-05-01

The objective of the Reactor Physics Quarterly Report is to inform the scientific community in a timely manner of the technical progress made on the many phases of reactor physics work within the laboratory. The report contains brief technical discussions of accomplishments in all areas where significant progress has been made during the quarter.

Technology development for cobalt F-T catalysts. Quarterly technical progress report number 10, January 1--March 31, 1995

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

Singleton, A.H.

1995-06-28

The goal of this project is the development of a commercially-viable, cobalt-based Fischer-Tropsch (F-T) catalyst for use in a slurry bubble column reactor. The major objectives of this work are (1) to develop a cobalt-based F-T catalyst with low (< 5%) methane selectivity, (2) to develop a cobalt-based F-T catalyst with water-gas shift activity, and (3) to combine both these improvements into one catalyst. The project consists of five major tasks: catalyst development; catalyst testing; catalyst reproducibility tests; catalyst aging tests; and preliminary design and cost estimate for a demonstrate scale catalyst production facility. Technical accomplishments during this reporting periodmore » include the following. It appears that the higher activity obtained for the catalysts prepared using an organic solution and reduced directly without prior calcination was the result of higher dispersions obtained under such pretreatment. A Ru-promoted Co catalyst on alumina with 30% Co loading exhibited a 4-fold increase in dispersion and a 2-fold increase in activity in the fixed-bed reactor from that obtained with the non-promoted catalyst. Several reactor runs have again focused on pushing conversion to higher levels. The maximum conversion obtained has been 49.7% with 26g catalyst. Further investigations of the effect of reaction temperature on the performance of Co catalysts during F-T synthesis were started using a low activity catalyst and one of the most active catalysts. The three 1 kg catalyst batches prepared by Calsicat for the reproducibility and aging studies were tested in both the fixed-bed and slurry bubble column reactors under the standard reaction conditions. The effects of adding various promoters to some cobalt catalysts have also been addressed. Results are presented and discussed.« less

Integrated intelligent systems in advanced reactor control rooms

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

Beckmeyer, R.R.

1989-01-01

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs.,more » 5 figs.« less

Technical note on results of production test IP-628-AI: Crosstie flow test and flushing calibration, and the earlier production tests IP-573-AI and IP-499-AI

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

Jones, S.S.

1964-06-15

A series of three experimental tests was conducted on the operation and adequacy of the K-Reactors` third, or last-ditch, cooling system. The first test showed considerable line corrosion the second test was performed directly after line cleaning, and the third test showed a significant amount of additional line corrosion after only nine months of service. The present cooling adequacy of this last-ditch system at the KE and KW reactors is summarized these show the power levels for which we have adequate last-ditch cooling as a function of the crosstie coolant temperature. These figures include the effects of increasing the numbermore » of pumps that remain in operation at the other K-Reactor, and various other operating or emergency conditions. These curves are for diesel pump speeds of 750 rpm which are planned for June of this year. In these figures the crosstie temperature is assumed to be a conservative 5{degrees}C above the process water inlet temperature.« less

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

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

None

1960-10-31

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

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

A Roadmap of Innovative Nuclear Energy System

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2017-01-01

Nuclear is a dense energy without CO2 emission. It can be used for more than 100,000 years using fast breeder reactors with uranium from the sea. However, it raises difficult problems associated with severe accidents, spent fuel waste and nuclear threats, which should be solved with acceptable costs. Some innovative reactors have attracted interest, and many designs have been proposed for small reactors. These reactors are considered much safer than conventional large reactors and have fewer technical obstructions. Breed-and-burn reactors have high potential to solve all inherent problems for peaceful use of nuclear energy. However, they have some technical problems with materials. A roadmap for innovative reactors is presented herein.

First flush reactor for stormwater treatment for elevated linear transportation projects : technical summary.

DOT National Transportation Integrated Search

2009-06-01

The primary objective of this study was to design and test a first flush-based stormwater treatment device for elevated linear transportation projects/roadways that is capable of complying with MS4 regulations. The innovative idea behind the device i...

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

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

ALLIANCE LIMITED, NGNP INDUSTRY

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

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

Bernander, O.; Haga, I.; Segerberg, F.

BS>From international nuclear industries fair; Basel, Switzerland (16 Oct 1972). Although the present status of the boiling water reactor is one of proven technology, design refinements and technical innovations are still being made to further improve reliability, economy and safety. The new standard ASEA- ATOM BWR features a number of such refinements and design improvements involving main circulation punips, containment design, refuelling system and off-gas treatment plant. In some respects the nuclear and hydraulic design of the ASEA- ATOM BWR differs from that adopted by other BWR manufacturers. Since the Oskarshamn I plant was the first nuclear power station havingmore » these features an extensive physics and hydraulics test program was made during the reactor start- up. The results of these tests have fully confirmed the ability of calculation methods to predict the behavior of the reactor. (auth)« less

JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

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

G. Borges

2006-01-31

The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long termmore » microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.« less

SP-100 ground engineering system test site description and progress update

NASA Astrophysics Data System (ADS)

Baxter, William F.; Burchell, Gail P.; Fitzgibbon, Davis G.; Swita, Walter R.

1991-01-01

The SP-100 Ground Engineering System Test Site will provide the facilities for the testing of an SP-100 reactor, which is technically prototypic of the generic design for producing 100 kilowatts of electricity. This effort is part of the program to develop a compact, space-based power system capable of producing several hundred kilowatts of electrical power. The test site is located on the U.S. Department of Energy's Hanford Site near Richland, Washington. The site is minimizing capital equipment costs by utilizing existing facilities and equipment to the maximum extent possible. The test cell is located in a decommissioned reactor containment building, and the secondary sodium cooling loop will use equipment from the Fast Flux Test Facility plant which has never been put into service. Modifications to the facility and special equipment are needed to accommodate the testing of the SP-100 reactor. Definitive design of the Ground Engineering System Test Site facility modifications and systems is in progress. The design of the test facility and the testing equipment will comply with the regulations and specifications of the U.S. Department of Energy and the State of Washington.

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

Safety philosophy of gas turbine high temperature reactor (GTHTR300)

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

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

2002-07-01

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

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

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

Jacobsen, George; Back, Christina

2015-10-30

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called themore » endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.« less

Liquid Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.

2007-01-01

Multiple liquid metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to rest prototypical space nuclear surface power system components. Conduction, induction and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated with incorporation into a reactor system. A thermoelectric electromagnetic pump is selected as the best option for use in NASA-MSFC's Fission Surface Power-Primary Test Circuit reactor simulator based on its relative simplicity, low power supply mass penalty, flight heritage, and the promise of increased pump efficiency over those earlier pump designs through the use of skutterudite thermoelectric elements.

Status of the US RERTR Program

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

Travelli, A.

1995-02-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1994 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1993 in collaboration with its many international partners. The RERTR Program has moved aggressively to support President Clinton`s nonproliferation policy and his goal {open_quotes}to minimize the use of highly-enriched uranium in civil nuclear programs{close_quotes}. An Environmental Assessment which addresses the urgent-relief acceptance of 409 spent fuel elements was completed, and the first shipment of spent fuel elements is scheduledmore » for this month. An Environmental Impact Statement addressing the acceptance of spent research reactor fuel containing enriched uranium of U.S. origin is scheduled for completion by the end of June 1995. The U.S. administration has decided to resume development of high-density LEU research reactor fuels. DOE funding and guidance are expected to begin soon. A preliminary plan for the resumption of fuel development has been prepared and is ready for implementation. The scope and main technical activities of a plan to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels was agreed upon by the RERTR Program and four Russian institutes lead by RDIPE. Both Secretary O`Leary and Minister Michailov have expressed strong support for this initiative. Joint studies have made significant progress, especially in assessing the technical and economic feasibility of using reduced enrichment fuels in the SAFARI-I reactor in South Africa and in the Advanced Neutron Source reactor under design at ORNL. Significant progress was achieved on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU to the achievement of the common goal.« less

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

NASA Technical Reports Server (NTRS)

Knight, J. C.

1996-01-01

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

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

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

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

Assessment of Sensor Technologies for Advanced Reactors

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

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

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

ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

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

Geringer, J. W.; Katoh, Yutai; Howard, Richard H.

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterizationmore » of irradiated materials and the shipment of irradiated materials to Japan. This report discusses the conceptual design, the development and irradiation of the test vehicles.« less

Double Retort System for Materials Compatibility Testing

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

V. Munne; EV Carelli

2006-02-23

With Naval Reactors (NR) approval of the Naval Reactors Prime Contractor Team (NRPCT) recommendation to develop a gas cooled reactor directly coupled to a Brayton power conversion system as the Space Nuclear Power Plant (SNPP) for Project Prometheus (References a and b) there was a need to investigate compatibility between the various materials to be used throughout the SNPP. Of particular interest was the transport of interstitial impurities from the nickel-base superalloys, which were leading candidates for most of the piping and turbine components to the refractory metal alloys planned for use in the reactor core. This kind of contaminationmore » has the potential to affect the lifetime of the core materials. This letter provides technical information regarding the assembly and operation of a double retort materials compatibility testing system and initial experimental results. The use of a double retort system to test materials compatibility through the transfer of impurities from a source to a sink material is described here. The system has independent temperature control for both materials and is far less complex than closed loops. The system is described in detail and the results of three experiments are presented.« less

The TRIGA Reactor Facility at the Armed Forces Radiobiology Research Institute: A Simplified Technical Description.

DTIC Science & Technology

1986-05-01

COUNT Technical FROM_ TO May 1986 20 16. SUPPLEMENTARY NOTATION 17. COSATI CODES 18. SUBJECT TERMS iConitinue on reverse if neceasary and identify by...Reactor, Modes of Operation, The AFRRI Reactor, Exposure Facilities, and Cerenkov Radiation. I- 20 DISTRISUTIONIAVAILABILITY OF ABSTRACT 21. ABSTRACT...6 Exposure Facilities 12 Cerenkov Radiation 17 Acoessiofl For NTIS GRA&I DT.C TABUnamnnounced [] UusnriOfltond -. By IZ Distribution/ Availability

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

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

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

Diamond, D. J.; Varuttamaseni, A.

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

A reload and startup plan for conversion of the NIST research reactor

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

D. J. Diamond

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

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

In-pile Hydrothermal Corrosion Evaluation of Coated SiC Ceramics and Composites

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

Carpenter, David; Ang, Caen; Katoh, Yutai

2017-09-01

Hydrothermal corrosion accelerated by water radiolysis during normal operation is among the most critical technical feasibility issues remaining for silicon carbide (SiC) composite-based cladding that could provide enhanced accident-tolerance fuel technology for light water reactors. An integrated in-pile test was developed and performed to determine the synergistic effects of neutron irradiation, radiolysis, and pressurized water flow, all of which are relevant to a typical pressurized water reactor (PWR). The test specimens were chosen to cover a range of SiC materials and a variety of potential options for environmental barrier coatings. This document provides a summary of the irradiation vehicle design,more » operations of the experiment, and the specimen loading into the irradiation vehicle.« less

Liquid-Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, K. A.

2007-01-01

Multiple liquid-metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to test prototypical space nuclear system components. Conduction, induction, and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated with incorporation into a reactor system. The thermoelectric pump is recommended for inclusion in the planned system at NASA MSFC based on its relative simplicity, low power supply mass penalty, flight heritage, and the promise of increased pump efficiency over earlier flight pump designs through the use of skutterudite thermoelectric elements.

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

NASA Astrophysics Data System (ADS)

Poston, David I.

2003-01-01

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

Alternate Tritium Production Methods Using A Liquid Lithium Target

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

Wilson, J.

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

Analysis of UF6 breeder reactor power plants

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1976-01-01

Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

77 FR 38742 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-29

...-0087] RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory Commission. ACTION... reactors. This contemplated rulemaking would also make conforming changes to address technical issues in existing non-power reactor regulations. The NRC is seeking input from the public, licensees, certificate...

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

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

E.T. Robinson; John Sirman; Prasad Apte

2005-05-01

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

An Overview of Facilities and Capabilities to Support the Development of Nuclear Thermal Propulsion

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

James Werner; Sam Bhattacharyya; Mike Houts

Abstract. The future of American space exploration depends on the ability to rapidly and economically access locations of interest throughout the solar system. There is a large body of work (both in the US and the Former Soviet Union) that show that Nuclear Thermal Propulsion (NTP) is the most technically mature, advanced propulsion system that can enable this rapid and economical access by its ability to provide a step increase above what is a feasible using a traditional chemical rocket system. For an NTP system to be deployed, the earlier measurements and recent predictions of the performance of the fuelmore » and the reactor system need to be confirmed experimentally prior to launch. Major fuel and reactor system issues to be addressed include fuel performance at temperature, hydrogen compatibility, fission product retention, and restart capability. The prime issue to be addressed for reactor system performance testing involves finding an affordable and environmentally acceptable method to test a range of engine sizes using a combination of nuclear and non-nuclear test facilities. This paper provides an assessment of some of the capabilities and facilities that are available or will be needed to develop and test the nuclear fuel, and reactor components. It will also address briefly options to take advantage of the greatly improvement in computation/simulation and materials processing capabilities that would contribute to making the development of an NTP system more affordable. Keywords: Nuclear Thermal Propulsion (NTP), Fuel fabrication, nuclear testing, test facilities.« less

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2013 CFR

2013-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2014 CFR

2014-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2012 CFR

2012-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

Implementation of k0-INAA standardisation at ITU TRIGA Mark II research reactor, Turkey based on k0-IAEA software

NASA Astrophysics Data System (ADS)

Esen, Ayse Nur; Haciyakupoglu, Sevilay

2016-02-01

The purpose of this study is to test the applicability of k0-INAA method at the Istanbul Technical University TRIGA Mark II research reactor. The neutron spectrum parameters such as epithermal neutron flux distribution parameter (α), thermal to epithermal neutron flux ratio (f) and thermal neutron flux (φth) were determined at the central irradiation channel of the ITU TRIGA Mark II research reactor using bare triple-monitor method. HPGe detector calibrations and calculations were carried out by k0-IAEA software. The α, f and φth values were calculated to be -0.009, 15.4 and 7.92·1012 cm-2 s-1, respectively. NIST SRM 1633b coal fly ash and intercomparison samples consisting of clay and sandy soil samples were used to evaluate the validity of the method. For selected elements, the statistical evaluation of the analysis results was carried out by z-score test. A good agreement between certified/reported and experimental values was obtained.

Fluidized bed gasification of industrial solid recovered fuels.

PubMed

Arena, Umberto; Di Gregorio, Fabrizio

2016-04-01

The study evaluates the technical feasibility of the fluidized bed gasification of three solid recovered fuels (SRFs), obtained as co-products of a recycling process. The SRFs were pelletized and fed to a pilot scale bubbling fluidized bed reactor, operated in gasification and co-gasification mode. The tests were carried out under conditions of thermal and chemical steady state, with a bed of olivine particles and at different values of equivalence ratio. The results provide a complete syngas characterization, in terms of its heating value and composition (including tars, particulates, and acid/basic pollutants) and of the chemical and physical characterization of bed material and entrained fines collected at the cyclone outlet. The feasibility of the fluidized bed gasification process of the different SRFs was evaluated with the support of a material and substance flow analysis, and a feedstock energy analysis. The results confirm the flexibility of fluidized bed reactor, which makes it one of the preferable technologies for the gasification of different kind of wastes, even in co-gasification mode. The fluidized bed gasification process of the tested SRFs appears technically feasible, yielding a syngas of valuable quality for energy applications in an appropriate plant configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

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.

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

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

Yamada, K.; Aksan, S. N.

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

Key findings and remaining questions in the areas of core-concrete interaction and debris coolability

DOE PAGES

Farmer, M. T.; Gerardi, C.; Bremer, N.; ...

2016-10-31

The reactor accidents at Fukushima-Dai-ichi have rekindled interest in late phase severe accident behavior involving reactor pressure vessel breach and discharge of molten core melt into the containment. Two technical issues of interest in this area include core-concrete interaction and the extent to which the core debris may be quenched and rendered coolable by top flooding. The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) programs at Argonne National Laboratory included the conduct of large scale reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensionalmore » molten core-concrete interactions under both wet and dry cavity conditions. These tests provided a broad database to support accident management planning, as well as the development and validation of models and codes that can be used to extrapolate the experiment results to plant conditions. This paper provides a high level overview of the key experiment results obtained during the program. Finally, a discussion is also provided that describes technical gaps that remain in this area, several of which have arisen based on the sequence of events and operator actions during Fukushima.« less

Key findings and remaining questions in the areas of core-concrete interaction and debris coolability

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

Farmer, M. T.; Gerardi, C.; Bremer, N.

The reactor accidents at Fukushima-Dai-ichi have rekindled interest in late phase severe accident behavior involving reactor pressure vessel breach and discharge of molten core melt into the containment. Two technical issues of interest in this area include core-concrete interaction and the extent to which the core debris may be quenched and rendered coolable by top flooding. The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) programs at Argonne National Laboratory included the conduct of large scale reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensionalmore » molten core-concrete interactions under both wet and dry cavity conditions. These tests provided a broad database to support accident management planning, as well as the development and validation of models and codes that can be used to extrapolate the experiment results to plant conditions. This paper provides a high level overview of the key experiment results obtained during the program. Finally, a discussion is also provided that describes technical gaps that remain in this area, several of which have arisen based on the sequence of events and operator actions during Fukushima.« less

Reducing The Nuclear Danger

DTIC Science & Technology

1995-10-01

off convention • Eliminate the civil use of HEU (includes RERTR ) • Reduce stockpiles of civil HEU and plutonium • Promote alternatives to the...these countries. ANL supports the Department’s Reduced Enrichment for Research and Test Reactor ( RERTR ) Program by providing the technical means to...scientists and engineers at 60 institutes in Russia, Ukraine, Kazakhstan and Belarus. The United States and Russia have agreed to pursue a joint RERTR

The Light Ion Pulsed Power Induction Accelerator for ETF

DTIC Science & Technology

1995-07-01

the technical development necessary to demonstrate scientific and engineering feasibility for fusion energy production with a reprated driver. In...order for ETF to be cost effective, the accelerator system must be able to drive several target chambers which will test various Inertial Fusion ... Energy (IFE) reactor technologies. We envision an elevator system positioning and removing multiple target chambers from the center area of the ion beam

10 CFR 54.21 - Contents of application-technical information.

Code of Federal Regulations, 2011 CFR

2011-01-01

...), motors, diesel generators, air compressors, snubbers, the control rod drive, ventilation dampers..., the reactor vessel, the reactor coolant system pressure boundary, steam generators, the pressurizer...

10 CFR 54.21 - Contents of application-technical information.

Code of Federal Regulations, 2010 CFR

2010-01-01

...), motors, diesel generators, air compressors, snubbers, the control rod drive, ventilation dampers..., the reactor vessel, the reactor coolant system pressure boundary, steam generators, the pressurizer...

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

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

Ingersoll, Daniel T

2007-01-01

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

Heater Development, Fabrication, and Testing: Analysis of Fabricated Heaters

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Dickens, R. E.; Farmer, J. T.; Davis, J. D.; Adams, M. R.; Martin, J. J.; Webster, K. L.

2008-01-01

Thermal simulators (highly designed heater elements) developed at the Early Flight Fission Test Facility (EFF-TF) are used to simulate the heat from nuclear fission in a variety of reactor concepts. When inserted into the reactor geometry, the purpose of the thermal simulators is to deliver thermal power to the test article in the same fashion as if nuclear fuel were present. Considerable effort has been expended to mimic heat from fission as closely as possible. To accurately represent the fuel, the simulators should be capable of matching the overall properties of the nuclear fuel rather than simply matching the fuel temperatures. This includes matching thermal stresses in the pin, pin conductivities, total core power, and core power profile (axial and radial). This Technical Memorandum discusses the historical development of the thermal simulators used in nonnuclear testing at the EFF-TF and provides a basis for the development of the current series of thermal simulators. The status of current heater fabrication and testing is assessed, providing data and analyses for both successes and failures experienced in the heater development and testing program.

Engineering analysis activities in support of susquehanna unit 1 startup testing and cycle 1 operations

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

Miller, G.D.; Kukielka, C.A.; Olson, L.M.

The engineering analysis group is responsible for all nuclear plant systems analysis and reactor analysis activities, excluding fuel management analysis, at Pennsylvania Power and Light Company. These activities include making pretest and posttest predictions of startup tests; analyzing unplanned or unexpected transient events; providing technical training to plant personnel; assisting in the development of emergency drill scenarios; providing engineering evaluations to support design and technical specification changes, and evaluating, assessing, and resolving a number of license conditions. Many of these activities have required the direct use of RETRAN models. Two RETRAN analyses that were completed to support plant operations -more » a pretest analysis of the turbine trip startup test, and a posttest analysis of the loss of startup transformer event - are investigated. For each case, RETRAN results are compared with available plant data and comparisons are drawn on the acceptability of the performance of the plant systems.« less

Some methods for achieving more efficient performance of fuel assemblies

NASA Astrophysics Data System (ADS)

Boltenko, E. A.

2014-07-01

More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2011 CFR

2011-01-01

... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C... radioactive waste, and/or reactor-related GTCC waste as appropriate, including how the ISFSI or MRS will be... of spent fuel, high-level radioactive waste, and/or reactor-related GTCC waste as appropriate for...

78 FR 79020 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Reliability...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-27

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Reliability & PRA; Notice of Meeting The ACRS Subcommittee on Reliability & PRA will hold a..., 2013. Cayetano Santos, Chief, Technical Support Branch, Advisory Committee on Reactor Safeguards. [FR...

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

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

Woods, Brian; Gutowska, Izabela; Chiger, Howard

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

Overview of Non-nuclear Testing of the Safe, Affordable 30-kW Fission Engine, Including End-to-End Demonstrator Testing

NASA Technical Reports Server (NTRS)

VanDyke, M. K.; Martin, J. J.; Houts, M. G.

2003-01-01

Successful development of space fission systems will require an extensive program of affordable and realistic testing. In addition to tests related to design/development of the fission system, realistic testing of the actual flight unit must also be performed. At the power levels under consideration (3-300 kW electric power), almost all technical issues are thermal or stress related and will not be strongly affected by the radiation environment. These issues can be resolved more thoroughly, less expensively, and in a more timely fashing with nonnuclear testing, provided it is prototypic of the system in question. This approach was used for the safe, affordable fission engine test article development program and accomplished viz cooperative efforts with Department of Energy labs, industry, universiites, and other NASA centers. This Technical Memorandum covers the analysis, testing, and data reduction of a 30-kW simulated reactor as well as an end-to-end demonstrator, including a power conversion system and an electric propulsion engine, the first of its kind in the United States.

Non-nuclear Testing of Reactor Systems in the Early Flight Fission Test Facilities (EFF-TF)

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Martin, James

2004-01-01

The Early Flight Fission-Test Facility (EFF-TF) can assist in the &sign and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are "non-nuclear" in nature (e.g. system s nuclear operations are understood). For many systems. thermal simulators can he used to closely mimic fission heat deposition. Axial power profile, radial power profile. and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other NASA centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004.

Nuclear Physics Made Very, Very Easy

NASA Technical Reports Server (NTRS)

Hanlen, D. F.; Morse, W. J.

1968-01-01

The fundamental approach to nuclear physics was prepared to introduce basic reactor principles to various groups of non-nuclear technical personnel associated with NERVA Test Operations. NERVA Test Operations functions as the field test group for the Nuclear Rocket Engine Program. Nuclear Engine for Rocket Vehicle Application (NERVA) program is the combined efforts of Aerojet-General Corporation as prime contractor, and Westinghouse Astronuclear Laboratory as the major subcontractor, for the assembly and testing of nuclear rocket engines. Development of the NERVA Program is under the direction of the Space Nuclear Propulsion Office, a joint agency of the U.S. Atomic Energy Commission and the National Aeronautics and Space Administration.

76 FR 69296 - Proposed Models for Plant-Specific Adoption of Technical Specifications Task Force Traveler TSTF...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-08

... Adoption of Technical Specifications Task Force Traveler TSTF-500, Revision 2, ``DC Electrical Rewrite... Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, ``DC Electrical Rewrite--Update to... Reactor Systems Engineer, Technical Specifications Branch, Mail Stop: O-7 C2A, Division of Inspection and...

Nuclear powerplants for mobile applications.

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. This paper examines the technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

Nuclear power plants for mobile applications

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. The technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants are examined. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

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...

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.

TECHNICAL SCOPE OF GAS-COOLED REACTOR FUEL ELEMENT IRRADIATION PROGRAM

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

None

A set of 55 experiments hss been outiined to provide a minimum irradiation program for selection of UO/sub 2/, pellet geometry and fabricntion techniques, and canning technology. These experiments fall into three catagories: prototype: untts in which radial dimension and heat fluxes sre close to proposed design values, but irradiation times are long; reduced-size prototype for accelerated tests in which most variables will be studied; and miniaurized pellet irradiation to obtain high burnup for fission gas release studies. Reactor space has been found generally available and several installations are now examining their capabilities to participate in the program. A tentativemore » schedule has been drawn to illustrate the feasibility of the program. (auth)« 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.

Acceptability of reactors in space

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

Buden, D.

1981-04-01

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

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

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

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

1992-12-31

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

Long-term storage facility for reactor compartments in Sayda Bay - German support for utilization of nuclear submarines in Russia

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

Wolff, Dietmar; Voelzke, Holger; Weber, Wolfgang

2007-07-01

The German-Russian project that is part of the G8 initiative on Global Partnership Against the Spread of Weapons and Materials of Mass Destruction focuses on the speedy construction of a land-based interim storage facility for nuclear submarine reactor compartments at Sayda Bay near Murmansk. This project includes the required infrastructure facilities for long-term storage of about 150 reactor compartments for a period of about 70 years. The interim storage facility is a precondition for effective activities of decommissioning and dismantlement of almost all nuclear-powered submarines of the Russian Northern Fleet. The project also includes the establishment of a computer-assisted wastemore » monitoring system. In addition, the project involves clearing Sayda Bay of other shipwrecks of the Russian navy. On the German side the project is carried out by the Energiewerke Nord GmbH (EWN) on behalf of the Federal Ministry of Economics and Labour (BMWi). On the Russian side the Kurchatov Institute holds the project management of the long-term interim storage facility in Sayda Bay, whilst the Nerpa Shipyard, which is about 25 km away from the storage facility, is dismantling the submarines and preparing the reactor compartments for long-term interim storage. The technical monitoring of the German part of this project, being implemented by BMWi, is the responsibility of the Federal Institute for Materials Research and Testing (BAM). This paper gives an overview of the German-Russian project and a brief description of solutions for nuclear submarine disposal in other countries. At Nerpa shipyard, being refurbished with logistic and technical support from Germany, the reactor compartments are sealed by welding, provided with biological shielding, subjected to surface treatment and conservation measures. Using floating docks, a tugboat tows the reactor compartments from Nerpa shipyard to the interim storage facility at Sayda Bay where they will be left on the on-shore concrete storage space to allow the radioactivity to decay. For transport of reactor compartments at the shipyard, at the dock and at the storage facility, hydraulic keel blocks, developed and supplied by German subcontractors, are used. In July 2006 the first stage of the reactor compartment storage facility was commissioned and the first seven reactor compartments have been delivered from Nerpa shipyard. Following transports of reactor compartments to the storage facility are expected in 2007. (authors)« less

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

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

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

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« less

Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

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

Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.

2009-10-09

The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected tomore » come from increasingly diverse educational and experiential backgrounds.« less

Reference Reactor Module for the Affordable Fission Surface Power System

NASA Astrophysics Data System (ADS)

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

2008-01-01

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

SNAP (Space Nuclear Auxiliary Power) reactor overview. Final report, June 1982-December 1983

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

Voss, S.S.

1984-08-01

The SNAP reactor programs are outlined in this report. A summary of the program is included along with a technical outline of the SER, S2DR, SNAP 10A/SNAPSHOT, S8ER, and S8DR reactor systems. Specifications of the designs, the design logic and a conclusion outlining some of the program weaknesses are given.

Implementing Stakeholders' Access to Expertise: Experimenting on Nuclear Installations' Safety Cases - 12160

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

Gilli, Ludivine; Charron, Sylvie

2012-07-01

In 2009 and 2010, the Institute for Nuclear Safety and Radiation Protection (IRSN) led two pilot actions dealing with nuclear installations' safety cases. One concerned the periodical review of the French 900 MWe nuclear reactors, the other concerned the decommissioning of a workshop located on the site of Areva's La Hague fuel-reprocessing plant site in Northwestern France. The purpose of both these programs was to test ways for IRSN and a small number of stakeholders (Non-Governmental Organizations (NGOs) members, local elected officials, etc.) to engage in technical discussions. The discussions were intended to enable the stakeholders to review future applicationsmore » and provide valuable input. The test cases confirmed there is a definite challenge in successfully opening a meaningful dialogue to discuss technical issues, in particular the fact that most expertise reports were not public and the conflict that exists between the contrary demands of transparency and confidentiality of information. The test case also confirmed there are ways to further improvement of stakeholders' involvement. (authors)« less

Municipal waste stabilization in a reactor with an integrated active and passive aeration system.

PubMed

Kasinski, Slawomir; Slota, Monika; Markowski, Michal; Kaminska, Anna

2016-04-01

To test whether an integrated passive and active aeration system could be an effective solution for aerobic decomposition of municipal waste in technical conditions, a full-scale composting reactor was designed. The waste was actively aerated for 5d, passively aerated for 35 d, and then actively aerated for 5d, and the entire composting process was monitored. During the 45-day observation period, changes in the fractional, morphological and physico-chemical characteristics of the waste at the top of the reactor differed from those in the center of the reactor. The fractional and morphological analysis made during the entire process of stabilization, showed the total reduction of organic matter measured of 82 wt% and 86 wt% at the respective depths. The reduction of organic matter calculated using the results of Lost of Ignition (LOI) and Total Organic Carbon (TOC) showed, respectively, 40.51-46.62% organic matter loss at the top and 45.33-53.39% in the center of the reactor. At the end of the process, moisture content, LOI and TOC at the top were 3.29%, 6.10% and 4.13% higher, respectively, than in the center. The results showed that application of passive aeration in larger scale simultaneously allows the thermophilic levels to be maintained during municipal solid waste composting process while not inhibiting microbial activity in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

Status and progress of the RERTR program in the year 2002.

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

Travelli, A.; Technology Development

2003-01-01

Following the cancellation of the 2001 International RERTR Meeting, which had been planned to occur in Bali, Indonesia, this paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the years 2001 and 2002, and discusses the main activities planned for the year 2003. The past two years have been characterized by very important achievements of the RERTR program, but these technical achievements have been overshadowed by the terrible events of September 11, 2001. Those events have caused the U.S. Government to reevaluate the importance andmore » urgency of the RERTR program goals. A recommendation made at the highest levels of the government calls for an immediate acceleration of the program activities, 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.« less

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

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.

Nuclear Fuel Reprocessing

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

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

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

Tiberi, V.

2012-07-01

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity ofmore » the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)« less

Self-teaching neural network learns difficult reactor control problem

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

Jouse, W.C.

1989-01-01

A self-teaching neural network used as an adaptive controller quickly learns to control an unstable reactor configuration. The network models the behavior of a human operator. It is trained by allowing it to operate the reactivity control impulsively. It is punished whenever either the power or fuel temperature stray outside technical limits. Using a simple paradigm, the network constructs an internal representation of the punishment and of the reactor system. The reactor is constrained to small power orbits.

Summary of the Workshop on Molten Salt Reactor Technologies Commemorating the 50th Anniversary of the Startup of the Molten Salt Reactor Experiment

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

Betzler, Benjamin R; Mays, Gary T

2016-01-01

A workshop on Molten Salt Reactor (MSR) technologies commemorating the 50th anniversary of the Molten Salt Reactor Experiment (MSRE) was held at Oak Ridge National Laboratory on October 15 16, 2015. The MSRE represented a pioneering experiment that demonstrated an advanced reactor technology: the molten salt eutectic-fueled reactor. A multinational group of more than 130 individuals representing a diverse set of stakeholders gathered to discuss the historical, current, and future technical challenges and paths to deployment of MSR technology. This paper provides a summary of the key messages from this workshop.

Computer simulation of magnetization-controlled shunt reactors for calculating electromagnetic transients in power systems

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

Karpov, A. S.

2013-01-15

A computer procedure for simulating magnetization-controlled dc shunt reactors is described, which enables the electromagnetic transients in electric power systems to be calculated. It is shown that, by taking technically simple measures in the control system, one can obtain high-speed reactors sufficient for many purposes, and dispense with the use of high-power devices for compensating higher harmonic components.

77 FR 30335 - Proposed Revision 3 to Standard Review Plan, Section 19.1 on Determining the Technical Adequacy...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-22

...). The Office of New Reactors and Office of Nuclear Reactor Regulation are revising SRP Section 19.1... of the Code of Federal Regulations (10 CFR), 50.71(h)(1), (h)(2), and (h)(3) for new reactors, (2... searching on http://www.regulations.gov under Docket ID NRC-2012-0113. You may submit comments by the...

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

James K. Neathery; Gary Jacobs; Burtron H. Davis

In this reporting period, a fundamental filtration study was started to investigate the separation of Fischer-Tropsch Synthesis (FTS) liquids from iron-based catalyst particles. Slurry-phase FTS in slurry bubble column reactor systems is the preferred mode of production since the reaction is highly exothermic. Consequently, heavy wax products must be separated from catalyst particles before being removed from the reactor system. Achieving an efficient wax product separation from iron-based catalysts is one of the most challenging technical problems associated with slurry-phase FTS. The separation problem is further compounded by catalyst particle attrition and the formation of ultra-fine iron carbide and/or carbonmore » particles. Existing pilot-scale equipment was modified to include a filtration test apparatus. After undergoing an extensive plant shakedown period, filtration tests with cross-flow filter modules using simulant FTS wax slurry were conducted. The focus of these early tests was to find adequate mixtures of polyethylene wax to simulate FTS wax. Catalyst particle size analysis techniques were also developed. Initial analyses of the slurry and filter permeate particles will be used by the research team to design improved filter media and cleaning strategies.« less

PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

NASA Astrophysics Data System (ADS)

Leclercq, Sylvain; Lidbury, David; Van Dyck, Steven; Moinereau, Dominique; Alamo, Ana; Mazouzi, Abdou Al

2010-11-01

In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations… from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach young researchers in the field of materials' degradation. PERFORM 60 has officially started on March 1st, 2009 with 20 European organizations and Universities involved in the nuclear field.

U.S.-Russian Civilian Nuclear Cooperation Agreement: Issues for Congress

DTIC Science & Technology

2010-07-09

for nuclear cooperation in 1973 to allow for cooperation in controlled thermonuclear fusion, fast breeder reactors , and fundamental research. The...that a 123 agreement is needed to implement this action plan—for example, full scale technical cooperation on fast reactors and demonstration of...superpowers convened a Joint Coordinating Committee for Civilian Reactor Safety starting in 1988.10 After the fall of the Soviet Union and prior to July

Reactor-pumped laser facility at DOE's Nevada Test Site

NASA Astrophysics Data System (ADS)

Lipinski, Ronald J.

1994-05-01

The Nevada Test Site (NTS) is one excellent possibility for a laser power beaming site. It is in the low latitudes of the U.S., is in an exceptionally cloud-free area of the southwest, is already an area of restricted access (which enhances safety considerations), and possesses a highly skilled technical team with extensive engineering and research capabilities from underground testing of our nation's nuclear deterrence. The average availability of cloud-free clear line of site to a given point in space is about 84%. With a beaming angle of +/- 60 degree(s) from the zenith, about 52 geostationary-orbit (GEO) satellites could be accessed continuously from NTS. In addition, the site would provide an average view factor of about 10% for orbital transfer from low earth orbit to GEO. One of the major candidates for a long-duration, high- power laser is a reactor-pumped laser being developed by DOE. The extensive nuclear expertise at NTS makes this site a prime candidate for utilizing the capabilities of a rector pumped laser for power beaming. The site then could be used for many dual-use roles such as industrial material processing research, defense testing, and removing space debris.

Reactor Simulator Integration and Testing

NASA Technical Reports Server (NTRS)

Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

2013-01-01

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

76 FR 73727 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-29

... Detection Instrumentation,'' to define a new time limit for restoring inoperable RCS leakage detection... Specifications Section 3.4.15 regarding reactor coolant leakage detection instrumentation to be consistent with... amendments revised the Technical Specifications (TSs) ``RCS [reactor coolant system] Leakage Detection...

Heat production in depth up to 2500m via in situ combustion of methane using a counter-current heat-exchange reactor

NASA Astrophysics Data System (ADS)

Schicks, Judith Maria; Spangenberg, Erik; Giese, Ronny; Heeschen, Katja; Priegnitz, Mike; Luzi-Helbing, Manja; Thaler, Jan; Abendroth, Sven; Klump, Jens

2014-05-01

In situ combustion is a well-known method used for exploitation of unconventional oil deposits such as heavy oil/bitumen reservoirs where the required heat is produced directly within the oil reservoir by combustion of a small percentage of the oil. A new application of in situ combustion for the production of methane from hydrate-bearing sediments was tested at pilot plant scale within the first phase of the German national gas hydrate project SUGAR. The applied method of in situ combustion was a flameless, catalytic oxidation of CH4 in a counter-current heat-exchange reactor with no direct contact between the catalytic reaction zone and the reservoir. The catalyst permitted a flameless combustion of CH4 with air to CO2 and H2O below the auto-ignition temperature of CH4 in air (868 K) and outside the flammability limits. This led to a double secured application of the reactor. The relatively low reaction temperature allowed the use of cost-effective standard materials for the reactor and prevented NOx formation. Preliminary results were promising and showed that only 15% of the produced CH4 was needed to be catalytically burned to provide enough heat to dissociate the hydrates in the environment and release CH4. The location of the heat source right within the hydrate-bearing sediment is a major advantage for the gas production from natural gas hydrates as the heat is generated where it is needed without loss of energy due to transportation. As part of the second period of the SUGAR project the reactor prototype of the first project phase was developed further to a borehole tool. The dimensions of this counter-current heat-exchange reactor are about 540 cm in length and 9 cm in diameter. It is designed for applications up to depths of 2500 m. A functionality test and a pressure test of the reactor were successfully carried out in October 2013 at the continental deep drilling site (KTB) in Windischeschenbach, Germany, in 600 m depth and 2000 m depth, respectively. In this study we present technical details of the reactor, the catalyst and potential fields of application beside the production of natural gas from hydrate bearing sediments.

History of Nuclear India

NASA Astrophysics Data System (ADS)

Chaturvedi, Ram

2000-04-01

India emerged as a free and democratic country in 1947, and entered into the nuclear age in 1948 by establishing the Atomic Energy Commission (AEC), with Homi Bhabha as the chairman. Later on the Department of Atomic Energy (DAE) was created under the Office of the Prime Minister Jawahar Lal Nehru. Initially the AEC and DAE received international cooperation, and by 1963 India had two research reactors and four nuclear power reactors. In spite of the humiliating defeat in the border war by China in 1962 and China's nuclear testing in 1964, India continued to adhere to the peaceful uses of nuclear energy. On May 18, 1974 India performed a 15 kt Peaceful Nuclear Explosion (PNE). The western powers considered it nuclear weapons proliferation and cut off all financial and technical help, even for the production of nuclear power. However, India used existing infrastructure to build nuclear power reactors and exploded both fission and fusion devices on May 11 and 13, 1998. The international community viewed the later activity as a serious road block for the Non-Proliferation Treaty and the Comprehensive Test Ban Treaty; both deemed essential to stop the spread of nuclear weapons. India considers these treaties favoring nuclear states and is prepared to sign if genuine nuclear disarmament is included as an integral part of these treaties.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

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

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

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

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

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

2016-01-01

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

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2011 CFR

2011-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2010 CFR

2010-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2013 CFR

2013-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2014 CFR

2014-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2012 CFR

2012-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

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

Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

1992-03-01

This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

10 CFR 52.137 - Contents of applications; technical information.

Code of Federal Regulations, 2010 CFR

2010-01-01

... limits on its operation, and presents a safety analysis of the structures, systems, and components and of... products. The description shall be sufficient to permit understanding of the system designs and their relationship to the safety evaluations. Items such as the reactor core, reactor coolant system, instrumentation...

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

Not Available

The SPS Concept Development and Evaluation Program includes a comparative assessment. An early first step in the assessment process is the selection and characterization of alternative technologies. This document describes the cost and performance (i.e., technical and environmental) characteristics of six central station energy alternatives: (1) conventional coal-fired powerplant; (2) conventional light water reactor (LWR); (3) combined cycle powerplant with low-Btu gasifiers; (4) liquid metal fast breeder reactor (LMFBR); (5) photovoltaic system without storage; and (6) fusion reactor.

10 CFR 72.28 - Contents of application: Applicant's technical qualifications.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Contents of application: Applicant's technical qualifications. 72.28 Section 72.28 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED...

Beta-glucan production by Botryosphaeria rhodina in different bench-top bioreactors.

PubMed

Selbmann, L; Crognale, S; Petruccioli, M

2004-01-01

Evaluation of the technical feasibility of transferring beta-glucan production by Botryosphaeria rhodina DABAC-P82 from shaken flasks to bench-top bioreactors. Three different bioreactors were used: 3 l stirred tank reactor (STR-1) equipped with two different six-blade turbines; STR as above but equipped with a three-blade marine propeller plus draft-tube (STR-2); 2 l air-lift column reactor (ALR) equipped with an external loop. STR-1, tested at three different stirrer speeds (300, 500 and 700 rev min(-1)) appeared to be less suitable for beta-glucan production by the fungus, being maximum production (19.4 g l(-1)), productivity (0.42 g l(-1) h(-1)) and yield (0.48 g g(-1) of glucose consumed) markedly lower than those obtained in shaken culture (29.7 g l(-1), 1.23 g l(-1) h(-1) and 0.61 g g(-1), respectively). Better performances were obtained with both STR-2 and ALR. With the latter, in particular, the increase of production was accompanied by reduced fermentation time (25.7 g l(-1) after only 22 h); productivity and yield were highest (1.17 g l(-1) h(-1) and 0.62 g g(-1) of glucose consumed, respectively). Using an air-lift reactor with external loop, the scaling up from shaken flasks to bench-top bioreactor of the beta-glucan production by B. rhodina DABAC-P82 is technically feasible. Although culture conditions are still to be optimized, the results obtained using the ARL are highly promising.

Neutron fluxes in test reactors

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

Youinou, Gilles Jean-Michel

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

Safety Issues at the Defense Production Reactors. A Report to the U.S. Department of Energy.

ERIC Educational Resources Information Center

National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

This report provides an assessment of safety management, safety review, and safety methodology employed by the Department of Energy (DOE) and private contractors. Chapter 1, "The DOE Safety Framework," examines safety objectives for production reactors and processes to implement the objectives. Chapter 2, "Technical Issues,"…

Documentation of Stainless Steel Lithium Circuit Test Section Design. Suppl

NASA Technical Reports Server (NTRS)

Godfroy, Thomas J. (Compiler); Martin, James J.

2010-01-01

The Early Flight Fission-Test Facilities (EFF-TF) team was tasked by Naval Reactors Prime Contract Team (NRPCT) to design, fabricate, and test an actively pumped lithium (Li) flow circuit. This Li circuit takes advantage of work in progress at the EFF TF on a stainless steel sodium/potassium (NaK) circuit. The effort involved modifying the original stainless steel NaK circuit such that it could be operated with Li in place of NaK. This new design considered freeze/thaw issues and required the addition of an expansion tank and expansion/extrusion volumes in the circuit plumbing. Instrumentation has been specified for Li and circuit heaters have been placed throughout the design to ensure adequate operational temperatures and no uncontrolled freezing of the Li. All major components have been designed and fabricated prior to circuit redesign for Li and were not modified. Basic circuit components include: reactor segment, Li to gas heat exchanger, electromagnetic liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. The reactor segment, based on a Los Alamos National Laboratory 100-kW design study with 120 fuel pins, is the only prototypic component in the circuit. However, due to earlier funding constraints, a 37-pin partial-array of the core, including the central three rings of fuel pins (pin and flow path dimensions are the same as those in the full design), was selected for fabrication and test. This Technical Publication summarizes the design and integration of the pumped liquid metal Li flow circuit as of May 1, 2005. This supplement contains drawings, analysis, and calculations

Documentation of Stainless Steel Lithium Circuit Test Section Design

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Martin, J. J.; Stewart, E. T.; Rhys, N. O.

2010-01-01

The Early Flight Fission-Test Facilities (EFF-TF) team was tasked by Naval Reactors Prime Contract Team (NRPCT) to design, fabricate, and test an actively pumped lithium (Li) flow circuit. This Li circuit takes advantage of work in progress at the EFF TF on a stainless steel sodium/potassium (NaK) circuit. The effort involved modifying the original stainless steel NaK circuit such that it could be operated with Li in place of NaK. This new design considered freeze/thaw issues and required the addition of an expansion tank and expansion/extrusion volumes in the circuit plumbing. Instrumentation has been specified for Li and circuit heaters have been placed throughout the design to ensure adequate operational temperatures and no uncontrolled freezing of the Li. All major components have been designed and fabricated prior to circuit redesign for Li and were not modified. Basic circuit components include: reactor segment, Li to gas heat exchanger, electromagnetic liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. The reactor segment, based on a Los Alamos National Laboratory 100-kW design study with 120 fuel pins, is the only prototypic component in the circuit. However, due to earlier funding constraints, a 37-pin partial-array of the core, including the central three rings of fuel pins (pin and flow path dimensions are the same as those in the full design), was selected for fabrication and test. This Technical Publication summarizes the design and integration of the pumped liquid metal Li flow circuit as of May 1, 2005.

10 CFR 72.26 - Contents of application: Technical specifications.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Contents of application: Technical specifications. 72.26 Section 72.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.26 - Contents of application: Technical specifications.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Contents of application: Technical specifications. 72.26 Section 72.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

A Figure of Merit: Quantifying the Probability of a Nuclear Reactor Accident.

PubMed

Wellock, Thomas R

In recent decades, probabilistic risk assessment (PRA) has become an essential tool in risk analysis and management in many industries and government agencies. The origins of PRA date to the 1975 publication of the U.S. Nuclear Regulatory Commission's (NRC) Reactor Safety Study led by MIT professor Norman Rasmussen. The "Rasmussen Report" inspired considerable political and scholarly disputes over the motives behind it and the value of its methods and numerical estimates of risk. The Report's controversies have overshadowed the deeper technical origins of risk assessment. Nuclear experts had long sought to express risk in a "figure of merit" to verify the safety of weapons and, later, civilian reactors. By the 1970s, technical advances in PRA gave the methodology the potential to serve political ends, too. The Report, it was hoped, would prove nuclear power's safety to a growing chorus of critics. Subsequent attacks on the Report's methods and numerical estimates damaged the NRC's credibility. PRA's fortunes revived when the 1979 Three Mile Island accident demonstrated PRA's potential for improving the safety of nuclear power and other technical systems. Nevertheless, the Report's controversies endure in mistrust of PRA and its experts.

Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson J. Boise; Reid, Robert S.

2007-01-01

As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

Experimental Evaluation of a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson, J. Boise; Reid, Robert S.

2006-01-01

As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

FAST FLUX TEST FACILITY CONCEPTUAL FACILTY DESIGN DESCRIPTION FOR THE INERT GAS CELL EXAMINATION FACILITY NO. 71

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

None

1968-12-12

The purpose of this Conceptual Facility Design Description (CFDD) is to provide a technical description of the Inert Gas Cell Examination Facility such that agreement with RDT on a Conceptual Design can be reached . The CFDD also serves to establish a common understanding of the facility concept among all responsible FFTF Project parties including the Architect Engineer and Reactor Designer. Included are functions and design requirements, a physical description of the facility, safety considerations, principles of operation, and maintenance principles.

Nuclear thermal propulsion transportation systems for lunar/Mars exploration

NASA Technical Reports Server (NTRS)

Clark, John S.; Borowski, Stanley K.; Mcilwain, Melvin C.; Pellaccio, Dennis G.

1992-01-01

Nuclear thermal propulsion technology development is underway at NASA and DoE for Space Exploration Initiative (SEI) missions to Mars, with initial near-earth flights to validate flight readiness. Several reactor concepts are being considered for these missions, and important selection criteria will be evaluated before final selection of a system. These criteria include: safety and reliability, technical risk, cost, and performance, in that order. Of the concepts evaluated to date, the Nuclear Engine for Rocket Vehicle Applications (NERVA) derivative (NDR) is the only concept that has demonstrated full power, life, and performance in actual reactor tests. Other concepts will require significant design work and must demonstrate proof-of-concept. Technical risk, and hence, development cost should therefore be lowest for the concept, and the NDR concept is currently being considered for the initial SEI missions. As lighter weight, higher performance systems are developed and validated, including appropriate safety and astronaut-rating requirements, they will be considered to support future SEI application. A space transportation system using a modular nuclear thermal rocket (NTR) system for lunar and Mars missions is expected to result in significant life cycle cost savings. Finally, several key issues remain for NTR's, including public acceptance and operational issues. Nonetheless, NTR's are believed to be the 'next generation' of space propulsion systems - the key to space exploration.

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

Characterization of alternative electric generation technologies for the SPS comparative assessment. Volume 1: Summary of central station technologies

NASA Astrophysics Data System (ADS)

1980-08-01

The technologies selected for the detailed characterization were: solar technology; terrestrial photovoltaic (200 MWe); coal technologies; conventional high sulfur coal combustion with advanced fine gas desulfurization (1250 MWe), and open cycle gas turbine combined cycle plant with low Btu gasifier (1250 MWe); and nuclear technologies: conventional light water reactor (1250 MWe), liquid metal fast breeder reactor (1250 MWe), and magnetic fusion reactor (1320 MWe). A brief technical summary of each power plant design is given.

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

Edgue, E.

The point kinetics approach is a classical useful method for a reactor transient analysis. It is helpful to known, however, when a more elaborate transient analysis, involving the space-dependence change of the flux through a given transient, should be considered. In this paper, the authors present a rather elegant and quick method to check the need for a space-dependent flux analysis through a control rod transient in a given nuclear reactor. The method is applied to a series of rod ejection experiments in the TRIGA MARK-II reactor of Istanbul Technical University (ITU).

Integrated Ceramic Membrane System for Hydrogen Production

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

Schwartz, Joseph; Lim, Hankwon; Drnevich, Raymond

2010-08-05

Phase I was a technoeconomic feasibility study that defined the process scheme for the integrated ceramic membrane system for hydrogen production and determined the plan for Phase II. The hydrogen production system is comprised of an oxygen transport membrane (OTM) and a hydrogen transport membrane (HTM). Two process options were evaluated: 1) Integrated OTM-HTM reactor – in this configuration, the HTM was a ceramic proton conductor operating at temperatures up to 900°C, and 2) Sequential OTM and HTM reactors – in this configuration, the HTM was assumed to be a Pd alloy operating at less than 600°C. The analysis suggestedmore » that there are no technical issues related to either system that cannot be managed. The process with the sequential reactors was found to be more efficient, less expensive, and more likely to be commercialized in a shorter time than the single reactor. Therefore, Phase II focused on the sequential reactor system, specifically, the second stage, or the HTM portion. Work on the OTM portion was conducted in a separate program. Phase IIA began in February 2003. Candidate substrate materials and alloys were identified and porous ceramic tubes were produced and coated with Pd. Much effort was made to develop porous substrates with reasonable pore sizes suitable for Pd alloy coating. The second generation of tubes showed some improvement in pore size control, but this was not enough to get a viable membrane. Further improvements were made to the porous ceramic tube manufacturing process. When a support tube was successfully coated, the membrane was tested to determine the hydrogen flux. The results from all these tests were used to update the technoeconomic analysis from Phase I to confirm that the sequential membrane reactor system can potentially be a low-cost hydrogen supply option when using an existing membrane on a larger scale. Phase IIB began in October 2004 and focused on demonstrating an integrated HTM/water gas shift (WGS) reactor to increase CO conversion and produce more hydrogen than a standard water gas shift reactor would. Substantial improvements in substrate and membrane performance were achieved in another DOE project (DE-FC26-07NT43054). These improved membranes were used for testing in a water gas shift environment in this program. The amount of net H2 generated (defined as the difference of hydrogen produced and fed) was greater than would be produced at equilibrium using conventional water gas shift reactors up to 75 psig because of the shift in equilibrium caused by continuous hydrogen removal. However, methanation happened at higher pressures, 100 and 125 psig, and resulted in less net H2 generated than would be expected by equilibrium conversion alone. An effort to avoid methanation by testing in more oxidizing conditions (by increasing CO2/CO ratio in a feed gas) was successful and net H2 generated was higher (40-60%) than a conventional reactor at equilibrium at all pressures tested (up to 125 psig). A model was developed to predict reactor performance in both cases with and without methanation. The required membrane area depends on conditions, but the required membrane area is about 10 ft2 to produce about 2000 scfh of hydrogen. The maximum amount of hydrogen that can be produced in a membrane reactor decreased significantly due to methanation from about 2600 scfh to about 2400 scfh. Therefore, it is critical to eliminate methanation to fully benefit from the use of a membrane in the reaction. Other modeling work showed that operating a membrane reactor at higher temperature provides an opportunity to make the reactor smaller and potentially provides a significant capital cost savings compared to a shift reactor/PSA combination.« less

77 FR 58590 - Determining Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-21

... reactors or for activities associated with review of applications for early site permits and combined licenses (COL) for the Office of New Reactors (NRO). DATES: The effective date of this SRP update is... Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2010-0138. Address...

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

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

Smith, Curtis

2013-09-01

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modelingmore » and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.« less

The status of beryllium technology for fusion

NASA Astrophysics Data System (ADS)

Scaffidi-Argentina, F.; Longhurst, G. R.; Shestakov, V.; Kawamura, H.

2000-12-01

Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented.

Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry; Garcia, Susan

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

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

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

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF 2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologiesmore » include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.« 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

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Production of aluminum--silicon alloy and ferrosilicon and commercial purity aluminum by the direct reduction process. First interim technical report, Phase B, September 1--November 30, 1978

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

Bruno, M.J.

1978-12-01

The parameters of charge content, reaction temperatures and residence time were studied in a bench reactor concerning the production of Al--Si and Fe--Si alloys. Results confirmed that minimum final stage reaction temperature is 1950 to 2000/sup 0/C. Residence time varied with initial charge concentration. Fe/sub 2/O/sub 3/ additions to the charge produced a significant increase in metallic yield. A burden preparation procedure was developed for making acceptable agglomerates containing Fe/sub 2/O/sub 3/, bauxite, clay and coke. Particle size distribution of starting materials was correlated with agglomerate strength. A new bench scale reactor was designed and built to facilitate semi-continuous operation,more » using O/sub 2/ injection to burn coke supporting the burden, resulting in burden movement. In a number of runs bridging of the burden material occurred due to condensation of volatilized sub-oxides in the cooler zones of the reactor. The reactor operated smoothly as an iron blast furnace at 1500/sup 0/C, demonstrating the validity of the equipment and test procedures. Initial construction of pilot reactor VSR-1 was completed. Bench scale fractional crystallizer runs were continued to determine the impurity effects of Fe up to 6.9% and Ti up to 1.25% on alloy product concentration and yield. High levels of impurities formed intermetallic complexes with Al and reduced product yield.« less

Integrated low emissions cleanup system for direct coal-fueled turbines

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

Lippert, T.E.; Newby, R.A.; Alvin, M.A.

1992-01-01

The Westinghouse Electric Corporation, Science Technology Center (W-STC) is developing an Integrated Low Emissions Cleanup (ILEC) concept for high-temperature gas cleaning to meet environmental standards, as well as to economical gas turbine life. The ILEC concept simultaneously controls sulfur, particulate, and alkali contaminants in high-pressure fuel gases or combustion gases at temperatures up to 1850[degrees]F for advanced power generation systems (PFBC, APFBC, IGCC, DCF7). The objective of this program is to demonstrate, at a bench scale, the conceptual, technical feasibility of the REC concept. The ELEC development program has a 3 phase structure: Phase 1 - laboratory-scale testing; phase 2more » - bench-scale equipment; design and fabrication; and phase 3 - bench-scale testing. Phase 1 laboratory testing has been completed. In Phase 1, entrained sulfur and alkali sorbent kinetics were measured and evaluated, and commercial-scale performance was projected. Related cold flow model testing has shown that gas-particle contacting within the ceramic barrier filter vessel will provide a good reactor environment. The Phase 1 test results and the commercial evaluation conducted in the Phase 1 program support the bench-scale facility testing to be performed in Phase 3. Phase 2 is nearing completion with the design and assembly of a modified, bench-scale test facility to demonstrate the technical feasibility of the ILEC features. This feasibility testing will be conducted in Phase 3.« less

Integrated low emissions cleanup system for direct coal-fueled turbines

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

Lippert, T.E.; Newby, R.A.; Alvin, M.A.

1992-12-31

The Westinghouse Electric Corporation, Science & Technology Center (W-STC) is developing an Integrated Low Emissions Cleanup (ILEC) concept for high-temperature gas cleaning to meet environmental standards, as well as to economical gas turbine life. The ILEC concept simultaneously controls sulfur, particulate, and alkali contaminants in high-pressure fuel gases or combustion gases at temperatures up to 1850{degrees}F for advanced power generation systems (PFBC, APFBC, IGCC, DCF7). The objective of this program is to demonstrate, at a bench scale, the conceptual, technical feasibility of the REC concept. The ELEC development program has a 3 phase structure: Phase 1 - laboratory-scale testing; phasemore » 2 - bench-scale equipment; design and fabrication; and phase 3 - bench-scale testing. Phase 1 laboratory testing has been completed. In Phase 1, entrained sulfur and alkali sorbent kinetics were measured and evaluated, and commercial-scale performance was projected. Related cold flow model testing has shown that gas-particle contacting within the ceramic barrier filter vessel will provide a good reactor environment. The Phase 1 test results and the commercial evaluation conducted in the Phase 1 program support the bench-scale facility testing to be performed in Phase 3. Phase 2 is nearing completion with the design and assembly of a modified, bench-scale test facility to demonstrate the technical feasibility of the ILEC features. This feasibility testing will be conducted in Phase 3.« 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

Small low mass advanced PBR's for propulsion

NASA Astrophysics Data System (ADS)

Powell, J. R.; Todosow, M.; Ludewig, H.

1993-10-01

The advanced Particle Bed Reactor (PBR) to be described in this paper is characterized by relatively low power, and low cost, while still maintaining competition values for thrust/weight, specific impulse and operating times. In order to retain competitive values for the thrust/weight ratio while reducing the reactor size, it is necessary to change the basic reactor layout, by incorporating new concepts. The new reactor design concept is termed SIRIUS (Small Lightweight Reactor Integral Propulsion System). The following modifications are proposed for the reactor design to be discussed in this paper: Pre-heater (U-235 included in Moderator); Hy-C (Hydride/De-hydride for Reactor Control); Afterburner (U-235 impregnated into Hot Frit); and Hy-S (Hydride Spike Inside Hot Frit). Each of the modifications will be briefly discussed below, with benefits, technical issues, design approach, and risk levels addressed. The paper discusses conceptual assumptions, feasibility analysis, mass estimates, and information needs.

Renewing Liquid Fueled Molten Salt Reactor Research and Development

NASA Astrophysics Data System (ADS)

Towell, Rusty; NEXT Lab Team

2016-09-01

Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

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

Environmental Cracking and Irradiation Resistant Stainless Steels by Additive Manufacturing

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

Rebak, Raul B.; Lou, Xiaoyuan

Metal additive manufacturing (AM), or metal 3D printing is an emergent advanced manufacturing method that can create near net shape geometries directly from computer models. This technology can provide the capability to rapidly fabricate complex parts that may be required to enhance the integrity of reactor internals components. Such opportunities may be observed during a plant refueling outage and AM parts can be rapidly custom designed, manufactured and deployed within the outage interval. Additive manufacturing of stainless steel (SS) components can add business benefits on fast delivery on repair hardware, installation tooling, new design prototypes tests, etc. For the nuclearmore » industry, the supply chain is always an issue for reactor service. AM can provide through-life supply chain (40-60 years) for high-value low-volume components. In the meantime, the capability of generating complex geometries and functional gradient materials will improve the performance, reduce the overall component cost, plant asset management cost and increase the plant reliability by the improvement in materials performance in nuclear environments. While extensive work has been conducted regarding additively manufacturing of austenitic SS parts, most efforts focused only on basic attributes such as porosity, residual stress, basic tensile properties, along with components yield and process monitoring. Little work has been done to define and evaluate the material requirements for nuclear applications. Technical gaps exist, which limit this technology adoption in the nuclear industry, which includes high manufacturing cost, unknown risks, limited nuclear related data, lack of specification and qualification methods, and no prior business experience. The main objective of this program was to generate research data to address all these technical gaps and establish a commercial practice to use AM technology in the nuclear power industry. The detailed objectives are listed as follows: (1) Evaluate nuclear related properties of AM 316L SS, including microstructure, tensile properties, impact toughness, stress corrosion cracking (SCC), corrosion fatigue (CF), irradiation effects, and irradiation assisted stress corrosion cracking (IASCC). (2) Understand the correlations among laser processing, heat treatment, microstructure and SCC/irradiation properties; (3) Optimize and improve the manufacturing process to achieve enhanced nuclear application properties; (4) Fabricate, evaluate, qualify and test a prototype reactor component to demonstrate the commercial viability and cost benefit; (5) Create regulatory approval path and commercialization plans for the production of a commercial reactor component.« less

76 FR 9614 - Notice of Availability of the Proposed Models for Plant-Specific Adoption of Technical...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-18

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0039] Notice of Availability of the Proposed Models for..., ``Technical Specifications End States, NEDC-32988-A,'' for Boiling Water Reactor Plants Using the Consolidated... Availability. SUMMARY: The NRC is announcing the availability of the model application (with model no...

126. ARAII Plot plan showing location of SL1 power plant ...

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

126. ARA-II Plot plan showing location of SL-1 power plant (reactor) building, and planned location of administrative and technical support building. C.A. Sundberg and Associates 866-area/ALPR-606-U-1. Date: May 1958. Ineel index code no. 070-0100-00-822-102834. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Magnetic Flux Compression Reactor Concepts for Spacecraft Propulsion and Power (MSFC Center Director's Discretionary Fund; Project No. 99-24). Part 1

NASA Technical Reports Server (NTRS)

Litchford, R. J.; Robertson, G. A.; Hawk, C. W.; Turner, M. W.; Koelfgen, S.; Litchford, Ron J. (Technical Monitor)

2001-01-01

This technical publication (TP) examines performance and design issues associated with magnetic flux compression reactor concepts for nuclear/chemical pulse propulsion and power. Assuming that low-yield microfusion detonations or chemical detonations using high-energy density matter can eventually be realized in practice, various magnetic flux compression concepts are conceivable. In particular, reactors in which a magnetic field would be compressed between an expanding detonation-driven plasma cloud and a stationary structure formed from a high-temperature superconductor are envisioned. Primary interest is accomplishing two important functions: (1) Collimation and reflection of a hot diamagnetic plasma for direct thrust production, and (2) electric power generation for fusion standoff drivers and/or dense plasma formation. In this TP, performance potential is examined, major technical uncertainties related to this concept accessed, and a simple performance model for a radial-mode reactor developed. Flux trapping effectiveness is analyzed using a skin layer methodology, which accounts for magnetic diffusion losses into the plasma armature and the stationary stator. The results of laboratory-scale experiments on magnetic diffusion in bulk-processed type II superconductors are also presented.

The U.S. RERTR program status and progress.

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

Travelli, A.

1998-01-21

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement ofmore » equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU Fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be satisfied by any existing fuel. These new advanced fuels will enable conversion of the reactors which cannot be converted today, ensure better efficiency and performance for all research reactors, and allow the design of more powerful new advanced LEU reactors. As in the past, the success of the RERTR program will depend on free exchange of ideas and information, and on the international friendship and cooperation that have been a trademark of the RERTR program since its inception.« less

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

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

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)« less

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

None, None

This report is concerned with the nature and scope of the technical services to be rendered and the general plan proposed for operation of Building 3525, High Radiation Level Examination Laboratory (HRLEL). The role of postirradiation examination in implementing the over- all task of irradiation testing for various programs under way at the Oak Ridge National Laboratory (ORNL) and the importance of this effort to the United Stat es reactor development program are stressed . The shielded-cell complex with provisions for remote decontamination, hot-equipment storage, and maintenance is described, as well as other supporting activities which are incorporated into themore » facility. The proposed technical functions include general observation, mensuration, nondestructive testing, burnup and induced-activity measurements, fission-gas sampling and analysis, corrosion evaluation and related measurements, disassembly and cutup, metallographic examination, mechanical-property determinations , and x -ray diffraction analyses. Equipment design and operational features to improve detection and measurement of selected properties in radioactive material s are described, also. The current status on design, procurement, construction, and preoperational testing of in- cell equipment in the mockup is presented along with a forecast of future needs. The mode of operation, manpower requirements, and management of the facility are discussed.« less

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

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

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

2008-07-15

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

Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

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

Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.

2013-05-17

This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components thatmore » may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and potential synergies with other national laboratory and university partners.« less

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

Davis, B.E.; Ahner, P.F.; Singelton, A.H.

In situ gasification of steeply dipping coal beds (UCG-SDB) has significant advantages over the more conventional horizontal UCG. In fact, the UCG-SDB process appears to be both technically and operationally competitive with surface gasifiers. The results of the Rawlins UCG-SDB field test program suggest that the process can compete with more conventional sources of synthesis gas on an economic basis. The SDB process mechanism has several advantages over the horizontal process and performs in a fashion similar to surface packedbed reactors. The oxygen requirements for the process are quite low and the degree of process control observed at Rawlins ismore » very attractive.« less

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

Davis, B.E.; Ahner, P.F.

In-situ gasification of steeply dipping coal beds (UCG-SDB) has significant advantages over the more conventional horizontal UCG. In fact, the UCG-SDB process appears to be both technically and operationally competitive with surface gasifiers. The results of the Rawlins UCG-SDB field test program suggest that the process can compete with more conventional sources of synthesis gas on an economic basis. The SDB process mechanism has several advantages over the horizontal process and performs in a fashion similar to surface packed bed reactors. The oxygen requirements for the process are quite low and the degree of process control observed at Rawlins ismore » very attractive.« less

Thermionic fuel element for the S-prime reactor

NASA Astrophysics Data System (ADS)

Van Hagan, Thomas H.; Drees, Elizabeth A.

1993-01-01

Technical aspects of the thermionic fuel element (TFE) design proposed for the S-PRIME space nuclear power system are discussed. Topics covered include the rational for selecting a multicell TFE approach, a technical description of the S-PRIME TFE and its estimated performance, and the technology readiness of the design, which emphasizes techology maturity and low risk.

129. ARAII Administrative and technical support building (ARA606) sections showing ...

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

129. ARA-II Administrative and technical support building (ARA-606) sections showing roof and wall details and longitudinal section. C.A. Sundberg and Associates 866-area/ALPR-606-A-5. Date: May 1958. Ineel index code no. 070-0606-00-822-102828. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

128. ARAII Administrative and technical support building (ARA606) elevations for ...

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

128. ARA-II Administrative and technical support building (ARA-606) elevations for northwest, southwest, northeast, and southeast sides. C.A. Sundberg and Associates 866-area/ALPR-606-A-3. Date: May 1958. Ineel index code no. 070-0606-00-822-102826. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

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

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

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

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

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

DOE PAGES

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

2017-02-26

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

Chemical vapor deposition growth

NASA Technical Reports Server (NTRS)

Ruth, R. P.; Manasevit, H. M.; Campbell, A. G.; Johnson, R. E.; Kenty, J. L.; Moudy, L. A.; Shaw, G. L.; Simpson, W. I.; Yang, J. J.

1978-01-01

The objective was to investigate and develop chemical vapor deposition (CVD) techniques for the growth of large areas of Si sheet on inexpensive substrate materials, with resulting sheet properties suitable for fabricating solar cells that would meet the technical goals of the Low Cost Silicon Solar Array Project. The program involved six main technical tasks: (1) modification and test of an existing vertical-chamber CVD reactor system; (2) identification and/or development of suitable inexpensive substrate materials; (3) experimental investigation of CVD process parameters using various candidate substrate materials; (4) preparation of Si sheet samples for various special studies, including solar cell fabrication; (5) evaluation of the properties of the Si sheet material produced by the CVD process; and (6) fabrication and evaluation of experimental solar cell structures, using impurity diffusion and other standard and near-standard processing techniques supplemented late in the program by the in situ CVD growth of n(+)/p/p(+) sheet structures subsequently processed into experimental cells.

Savannah River Plant engineering and design history. Volume 4: 300/700 Areas & general services and facilities

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

Not Available

1957-01-01

The primary function of the 300 Area is the production and preparation of the fuel and target elements required for the 100 Area production reactors. Uranium slugs and lithium-aluminium alloy control and blanket rods are prepared in separate structures. Other facilities include a test pile, a physics assembly laboratory, an office and change house, an electrical substation, and various service facilities such as rail lines, roads, sewers, steam and water distribution lines, etc. The 700 Area contains housing and facilities for plant management, general plant services, and certain technical activities. The technical buildings include the Main Technical Laboratory, the Wastemore » Concentration Building, the Health Physics Headquarters, and the Health Physics Calibration building. Sections of this report describe the following: development of the 300-M Area; selection and description of process; design of main facilities of the 300 Area; development of the 700-A Area; design of the main facilities of the 700 Area; and general services and facilities, including transportation, plant protection, waste disposal and drainage, site work, pilot plants, storage, and furniture and fixtures.« less

Kinetics of Supercritical Water Oxidation

DTIC Science & Technology

1995-12-31

milestone and Sandia Technical Report. A much-needed report describing in detail the operation of the Supercritical Fluids Reactor (SFR) was also...years. In addition, the literature research required to arrive at this optimal design will be used to improve the performance of the Supercritical Fluids ...the Supercritical Fluids Reactor (SFR)" (Sandia National Laboratories Report SAND-8203, Livermore, CA, 1995). R. R. Steeper, "Methane and Methanol

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

NASA Astrophysics Data System (ADS)

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

2014-08-01

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

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

Introduction to the special issue on the technical status of materials for a fusion reactor

NASA Astrophysics Data System (ADS)

Stork, D.; Zinkle, S. J.

2017-09-01

Materials determine in a fundamental way the performance and environmental attractiveness of a fusion reactor: through the size (power fluxes to the divertor, neutron fluxes to the first wall); economics (replacement lifetime of critical in-vessel components, thermodynamic efficiency through operating temperature etc); plasma performance (erosion by plasma fluxes to the divertor surfaces); robustness against off-normal accidents (safety); and the effects of post-operation radioactivity on waste disposal and maintenance. The major philosophies and methodologies used to formulate programmes for the development of fusion materials are outlined, as the basis for other articles in this special issue, which deal with the fundamental understanding of the issues regarding these materials and their technical status and prospects for development.

Simulation of sodium pumps for nuclear power plants. Technical report 1 Oct 80-1 May 81

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

Boadu, H.O.

1981-05-01

A single-phase pump model for analysis of transients in sodium cooled fast breeder nuclear power plants has been presented, where homologous characteristic curves are used to predict the behavior of the pump during operating transients. The pump model has been incorporated into BRENDA and FFTF; two system cases to simulate Clinch River Breeder Reactor Plant (CRBRP) and the Fast Flux Test Facility (FFTF) respectively. Two simulation test results for BRENDA which is one loop representation of a three loop plant have been presented. They are: (1) Primary pump coastdown to natural circulation coupled with scram failure, and (2) 10 percentmore » deviation of primary speed with plant controllers incorporated.« less

KINETICS OF TREAT USED AS A TEST REACTOR

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

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

The RERTR Program : a status report.

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

Travelli, A.

1998-10-19

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

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

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

Ligeti, G.

1962-04-01

Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« 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

Goals of thermionic program for space power

NASA Technical Reports Server (NTRS)

English, R. E.

1981-01-01

The thermionic and Brayton reactor concepts were compared for application to space power. For a turbine inlet temperature of 15000 K the Brayton powerplant weighted 5 to 40% less than the thermionic concept. The out of core concept separates the thermionic converters from their reactor. Technical risks are diminished by: (1) moving the insolator out of the reactor; (2) allowing a higher thermal flux for the thermionic converters than is required of the reactor fuel; and (3) eliminating fuel swelling's threat against lifetime of the thermionic converters. Overall performance can be improved by including power processing in system optimization for design and technology on more efficient, higher temperature power processors. The thermionic reactors will be larger than those for competitive systems with higher conversion efficiency and lower reactor operating temperatures. It is concluded that although the effect of reactor size on shield weight will be modest for unmanned spacecraft, the penalty in shield weight will be large for manned or man-tended spacecraft.

Impacts Analyses Supporting the National Environmental Policy Act Environmental Assessment for the Resumption of Transient Testing Program

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

Schafer, Annette L.; Brown, LLoyd C.; Carathers, David C.

2014-02-01

This document contains the analysis details and summary of analyses conducted to evaluate the environmental impacts for the Resumption of Transient Fuel and Materials Testing Program. It provides an assessment of the impacts for the two action alternatives being evaluated in the environmental assessment. These alternatives are (1) resumption of transient testing using the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) and (2) conducting transient testing using the Annular Core Research Reactor (ACRR) at Sandia National Laboratory in New Mexico (SNL/NM). Analyses are provided for radiologic emissions, other air emissions, soil contamination, and groundwater contamination that couldmore » occur (1) during normal operations, (2) as a result of accidents in one of the facilities, and (3) during transport. It does not include an assessment of the biotic, cultural resources, waste generation, or other impacts that could result from the resumption of transient testing. Analyses were conducted by technical professionals at INL and SNL/NM as noted throughout this report. The analyses are based on bounding radionuclide inventories, with the same inventories used for test materials by both alternatives and different inventories for the TREAT Reactor and ACRR. An upper value on the number of tests was assumed, with a test frequency determined by the realistic turn-around times required between experiments. The estimates provided for impacts during normal operations are based on historical emission rates and projected usage rates; therefore, they are bounding. Estimated doses for members of the public, collocated workers, and facility workers that could be incurred as a result of an accident are very conservative. They do not credit safety systems or administrative procedures (such as evacuation plans or use of personal protective equipment) that could be used to limit worker doses. Doses estimated for transportation are conservative and are based on transport of the bounding radiologic inventory that will be contained in any given test. The transportation analysis assumes all transports will contain the bounding inventory.« 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

127. ARAII Administrative and technical support building (ARA606) ground floor ...

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

127. ARA-II Administrative and technical support building (ARA-606) ground floor plan. Indicates use of rooms for classrooms, offices, and lunch room. C.A. Sundberg and Associates 866-area-ALPR-606-A-2. Date: June 1958. Ineel index code no. 070-0606-00-822-102825. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Advanced Coal Liquefaction Research and Development Facility, Wilsonville, Alabama. Run 260 with Black Thunder Mine subbituminous coal: Technical progress report

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

Not Available

This report presents the results of Run 260 performed at the Advanced Coal Liquefaction R&D Facility in Wilsonville. The run was started on July 17, 1990 and continued until November 14, 1990, operating in the Close-Coupled Integrated Two-Stage Liquefaction mode processing Black Thunder mine subbituminous coal (Wyodak-Anderson seam from Wyoming Powder River Basin). Both thermal/catalytic and catalytic/thermal tests were performed to determine the methods for reducing solids buildup in a subbituminous coal operation, and to improve product yields. A new, smaller interstage separator was tested to reduce solids buildup by increasing the slurry space velocity in the separator. In ordermore » to obtain improved coal and resid conversions (compared to Run 258) full-volume thermal reactor and 3/4-volume catalytic reactor were used. Shell 324 catalyst, 1/16 in. cylindrical extrudate, at a replacement rate of 3 lb/ton of MF coal was used in the catalytic stage. Iron oxide was used as slurry catalyst at a rate of 2 wt % MF coal throughout the run. (TNPS was the sulfiding agent.)« less

Edwin I. Hatch nuclear plant implementation of improved technical specifications

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

Mahler, S.R.; Pendry, D.

1994-12-31

Edwin I. Hatch nuclear plant consists of two General Electric boiling water reactor/4 units, with a common control room and a common refueling floor. In March 1993, Hatch began conversion of both units` technical specifications utilizing NUREG 1433. The technical specifications amendment request was submitted February 25, 1994. Issuance is scheduled for October 21, 1994, with implementation on March 15, 1994. The current unit-1 technical specifications are in the {open_quotes}custom{close_quotes} format, and the unit-2 technical specifications are in the old standard format. Hatch previously relocated the fire protection and radiological technical specifications requirements. The Hatch conversion will provide consistency betweenmore » the two units, to the extent practicable.« less

Research gaps and technology needs in development of PHM for passive AdvSMR components

NASA Astrophysics Data System (ADS)

Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Hirt, Evelyn H.; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henagar, Chuck H., Jr.

2014-02-01

Advanced small modular reactors (AdvSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near-term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. SMRs are challenged economically because of losses in economy of scale; thus, there is increased motivation to reduce the controllable operations and maintenance costs through automation technologies including prognostics health management (PHM) systems. In this regard, PHM systems have the potential to play a vital role in supporting the deployment of AdvSMRs and face several unique challenges with respect to implementation for passive AdvSMR components. This paper presents a summary of a research gaps and technical needs assessment performed for implementation of PHM for passive AdvSMR components.

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

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

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

Alloy catalysts with monolith supports for methanation of coal-derived gases. Quarterly technical progress report, September 21-December 20, 1979

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

Bartholomew, C.H.

1980-01-05

Tests for catalyst deactivation by carbon deposition in a Berty reactor showed that CO partial pressure exerted a greater influence on deactivation than H/sub 2/ partial pressure. In kinetic studies in which H/sub 2/O vapor was added to the reactant gases, H/sub 2/O was found to inhibit the methanation reaction. H/sub 2/O inhibition was found to increase with temperature and loss of activity was observed at H/sub 2/O/CO ratios greater than one. The order of methanation with respect to H/sub 2/ and CO varies over the range of temperature from 498 to 598/sup 0/K. Rate data indicate a change inmore » mechanism or rate determining step at higher temperatures. Our experience with a quartz CFSTR has shown that this reactor is very delicate and needs near constant attention to maintain proper working order.« less

MOCVD process technology for affordable, high-yield, high-performance MESFET structures. Phase 3: MIMIC

NASA Astrophysics Data System (ADS)

1993-01-01

Under the MIMIC Program, Spire has pursued improvements in the manufacturing of low cost, high quality gallium arsenide MOCVD wafers for advanced MIMIC FET applications. As a demonstration of such improvements, Spire was tasked to supply MOCVD wafers for comparison to MBE wafers in the fabrication of millimeter and microwave integrated circuits. In this, the final technical report for Spire's two-year MIMIC contract, we report the results of our work. The main objectives of Spire's MIMIC Phase 3 Program, as outlined in the Statement of Work, were as follows: Optimize the MOCVD growth conditions for the best possible electrical and morphological gallium arsenide. Optimization should include substrate and source qualification as well as determination of the optimum reactor growth conditions; Perform all work on 75 millimeter diameter wafers, using a reactor capable of at least three wafers per run; and Evaluate epitaxial layers using electrical, optical, and morphological tests to obtain thickness, carrier concentration, and mobility data across wafers.

Biodegradation of tech-hexachlorocyclohexane in a upflow anaerobic sludge blanket (UASB) reactor.

PubMed

Bhat, Praveena; Kumar, M Suresh; Mudliar, Sandeep N; Chakrabarti, T

2006-04-01

Biodegradability of technical grade hexachlorocyclohexane (tech-HCH) was studied in an upflow anaerobic sludge blanket reactor (UASB) under continuous mode of operation in concentration range of 100-200 mg/l and constant HRT of 48 h. At steady state operation more than 85% removal of tech-HCH (upto 175 mg/l concentration) and complete disappearance of beta-HCH was observed. Kinetic constants in terms of maximum specific tech-HCH utilization rate (k) and half saturation velocity constant (K(L)) were found to be 11.88 mg/g/day and 8.11 mg/g/day, respectively. The tech-HCH degrading seed preparation, UASB reactor startup and degradation in continuous mode of operation of the reactor is presented in this paper.

Measurements and calculations of water velocity, momentum flux, and related flow parameters obtaned from single-phase water integral acceptance tests of the PKL instrumented spool pieces

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

Stein, W.

The operation of the emergency core cooling system and its related steam-binding problems in pressurized water reactors is the subject of a cooperative study by the United States, Germany, and Japan. Lawrence Livermore Laboratory and EG and G, Inc., San Ramon Operations, are responsible for the design, hardware, and software of the 80.8-mm and 113-mm spool piece measurement systems for the German Primarkreislauf (PKL) Test Facility at Kraftwerk Union in Erlangen, West Germany. This work was done for the US Nuclear Regulatory Commission, Division of Reactor Safety Research, under its 3-D Technical Support and Instrumentation Program. Four instrumented spools capablemore » of measuring individual phase parameters in two-phase flows were constructed. Each spool contains a flow turbine, drag screen, three-beam densitometer, and pressure and temperature probes. A computerized data acquisition system is also provided to store and analyze data from the four spools. The four spools were shipped to the PKL Test Facility in West Germany for acceptance testing in a water-flow loop. Spool measurements of velocity and momentum flux were compared to the values obtained from an orifice meter installed in the loop piping system. The turbine flowmeter velocity data for all tests were within allowable tolerances. Drag screen momentum flux measurements were also within tolerance with the exception of a few points.« less

ENHANCED PRACTICAL PHOTOSYNTHETIC CO2 MITIGATION

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

Dr. Gregory Kremer; Dr. David J. Bayless; Dr. Morgan Vis

2001-07-25

This quarterly report documents significant achievements in the Enhanced Practical Photosynthetic CO{sub 2} Mitigation project during the period from 4/03/2001 through 7/02/2001. Most of the achievements are milestones in our efforts to complete the tasks and subtasks that constitute the project objectives. Note that this version of the quarterly technical report is a revision to add the reports from subcontractors Montana State and Oak Ridge National Laboratories The significant accomplishments for this quarter include: Development of an experimental plan and initiation of experiments to create a calibration curve that correlates algal chlorophyll levels with carbon levels (to simplify future experimentalmore » procedures); Completion of debugging of the slug flow reactor system, and development of a plan for testing the pressure drop of the slug flow reactor; Design and development of a new bioreactor screen design which integrates the nutrient delivery drip system and the harvesting system; Development of an experimental setup for testing the new integrated drip system/harvesting system; Completion of model-scale bioreactor tests examining the effects of CO{sub 2} concentration levels and lighting levels on Nostoc 86-3 growth rates; Completion of the construction of a larger model-scale bioreactor to improve and expand testing capabilities and initiation of tests; Substantial progress on construction of a pilot-scale bioreactor; and Preliminary economic analysis of photobioreactor deployment. Plans for next quarter's work are included in the conclusions. A preliminary economic analysis is included as an appendix.« less

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

Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

NASA Technical Reports Server (NTRS)

Silverman, S. W.; Willenberg, H. J.; Robertson, C.

1985-01-01

An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

Thermo-Physics Technical Note No. 60: thermal analysis of SNAP 10A reactor core during atmospheric reentry and resulting core disintegration and fuel element separation

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

Mouradian, E.M.

1966-02-16

A thermal analysis is carried out to determine the temperature distribution throughout a SNAP 10A reactor core, particularly in the vicinity of the grid plates, during atmospheric reentry. The transient temperatue distribution of the grid plate indicates when sufficient melting occurs so that fuel elements are free to be released and continue their descent individually.

Photocatalytic Degradation Property of NANO-TiO2/DIATOMITE for Rodamine B Dye Wastewater

NASA Astrophysics Data System (ADS)

Liu, Yue; Zheng, Shuilin; Du, Gaoxiang; Shu, Feng; Chen, Juntao

The Nano-TiO2/Diatomite compound photocatalyst is used to degrade rhodamine B dye wastewater in photochemical reactor. The test result indicates that the rate of photodegradation of rhodamine B is influenced by reactive conditions. The best technical conditions are concentration of rhodamine B solution 10mg/L, ultraviolet light 300W, the compound photocatalyst amount used 1g/L, the pH 5.8, reaction time 20min. Under these conditions the rate of photodegradation of rhodamine B may reach as high as 97.80%. And the efficiency of photodegradation of catalyst only has a little changed in recycling.

Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

NASA Astrophysics Data System (ADS)

Ritz, G.; Hirai, T.; Norajitra, P.; Reiser, J.; Giniyatulin, R.; Makhankov, A.; Mazul, I.; Pintsuk, G.; Linke, J.

2009-12-01

Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ~14 MW m-2, the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

Small and medium power reactors 1987

NASA Astrophysics Data System (ADS)

1987-12-01

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

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

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

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

Index to Nuclear Safety. A technical progress review by chronology, permuted title, and author. Vol. 11, No. 1--Vol. 17, No. 6

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

Cottrell, W.B.; Klein, A.

1977-02-23

This index to Nuclear Safety covers articles in Nuclear Safety Vol. 11, No. 1 (Jan.-Feb. 1970), through Vol. 17, No. 6 (Nov.-Dec. 1976). The index includes a chronological list of articles (including abstract) followed by KWIC and Author Indexes. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. The index lists over 350 technical articles in the last six years of publication.

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

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

Block, R.C.; Feiner, F.

1995-09-01

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

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

Enhanced biodegradation of hexachlorocyclohexane in upflow anaerobic sludge blanket reactor using methanol as an electron donor.

PubMed

Bhatt, Praveena; Kumar, M Suresh; Mudliar, Sandeep; Chakrabarti, Tapan

2008-05-01

Anaerobic dechlorination of technical grade hexachlorocyclohexane (THCH) was studied in a continuous upflow anaerobic sludge blanket (UASB) reactor with methanol as a supplementary substrate and electron donor. A reactor without methanol served as the experimental control. The inlet feed concentration of THCH in both the experimental and the control UASB reactor was 100 mg l(-1). After 60 days of continuous operation, the removal of THCH was >99% in the methanol-supplemented reactor as compared to 20-35% in the control reactor. THCH was completely dechlorinated in the methanol fed reactor at 48 h HRT after 2 months of continuous operation. This period was also accompanied by increase in biomass in the reactor, which was not observed in the experimental control. Batch studies using other supplementary substrates as well as electron donors namely acetate, butyrate, formate and ethanol showed lower % dechlorination (<85%) and dechlorination rates (<3 mg g(-1)d(-1)) as compared to methanol (98%, 5 mg g(-1)d(-1)). The optimum concentration of methanol required, for stable dechlorination of THCH (100 mg l(-1)) in the UASB reactor, was found to be 500 mg l(-1). Results indicate that addition of methanol as electron donor enhances dechlorination of THCH at high inlet concentration, and is also required for stable UASB reactor performance.

Combustion Chemistry of Fuels: Quantitative Speciation Data Obtained from an Atmospheric High-temperature Flow Reactor with Coupled Molecular-beam Mass Spectrometer.

PubMed

Köhler, Markus; Oßwald, Patrick; Krueger, Dominik; Whitside, Ryan

2018-02-19

This manuscript describes a high-temperature flow reactor experiment coupled to the powerful molecular beam mass spectrometry (MBMS) technique. This flexible tool offers a detailed observation of chemical gas-phase kinetics in reacting flows under well-controlled conditions. The vast range of operating conditions available in a laminar flow reactor enables access to extraordinary combustion applications that are typically not achievable by flame experiments. These include rich conditions at high temperatures relevant for gasification processes, the peroxy chemistry governing the low temperature oxidation regime or investigations of complex technical fuels. The presented setup allows measurements of quantitative speciation data for reaction model validation of combustion, gasification and pyrolysis processes, while enabling a systematic general understanding of the reaction chemistry. Validation of kinetic reaction models is generally performed by investigating combustion processes of pure compounds. The flow reactor has been enhanced to be suitable for technical fuels (e.g. multi-component mixtures like Jet A-1) to allow for phenomenological analysis of occurring combustion intermediates like soot precursors or pollutants. The controlled and comparable boundary conditions provided by the experimental design allow for predictions of pollutant formation tendencies. Cold reactants are fed premixed into the reactor that are highly diluted (in around 99 vol% in Ar) in order to suppress self-sustaining combustion reactions. The laminar flowing reactant mixture passes through a known temperature field, while the gas composition is determined at the reactors exhaust as a function of the oven temperature. The flow reactor is operated at atmospheric pressures with temperatures up to 1,800 K. The measurements themselves are performed by decreasing the temperature monotonically at a rate of -200 K/h. With the sensitive MBMS technique, detailed speciation data is acquired and quantified for almost all chemical species in the reactive process, including radical species.

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

NASA Astrophysics Data System (ADS)

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

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

Degradation of toxaphene in water during anaerobic and aerobic conditions.

PubMed

LacayoR, M; van Bavel, B; Mattiasson, B

2004-08-01

The degradation of technical toxaphene in water with two kinds of bioreactors operating in sequence was studied. One packed bed reactor was filled with Poraver (foam glass particles) running at anaerobic conditions and one suspended carrier biofilm reactor working aerobically. Chemical oxygen demand (COD), chloride, sulphate, pH, dissolved oxygen, total toxaphene and specific toxaphene isomers were measured. After 6 weeks approx. 87% of the total toxaphene was degraded reaching 98% by week 39. The majority of the conversion took place in the anaerobic reactor. The concentrations of toxaphene isomers with more chlorine substituents decreased more rapidly than for isomers with less chlorine substituents.

Preliminary assessment of high power, NERVA-class dual-mode space nuclear propulsion and power systems

NASA Astrophysics Data System (ADS)

Buksa, John J.; Kirk, William L.; Cappiello, Michael W.

A preliminary assessment of the technical feasibility and mass competitiveness of a dual-mode nuclear propulsion and power system based on the NERVA rocket engine has been completed. Results indicate that the coupling of the Rover reactor to a direct Brayton power conversion system can be accomplished through a number of design features. Furthermore, based on previously published and independently calculated component masses, the dual-mode system was found to have the potential to be mass competitive with propulsion/power systems that use separate reactors. The uncertainties of reactor design modification and shielding requirements were identified as important issues requiring future investigation.

Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

NASA Astrophysics Data System (ADS)

Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

2017-11-01

The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

Reactor technology assessment and selection utilizing systems engineering approach

NASA Astrophysics Data System (ADS)

Zolkaffly, Muhammed Zulfakar; Han, Ki-In

2014-02-01

The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

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.

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

NASA Astrophysics Data System (ADS)

Stork, D.; Agostini, P.; Boutard, J. L.; Buckthorpe, D.; Diegele, E.; Dudarev, S. L.; English, C.; Federici, G.; Gilbert, M. R.; Gonzalez, S.; Ibarra, A.; Linsmeier, Ch.; Li Puma, A.; Marbach, G.; Morris, P. F.; Packer, L. W.; Raj, B.; Rieth, M.; Tran, M. Q.; Ward, D. J.; Zinkle, S. J.

2014-12-01

The findings of the EU 'Materials Assessment Group' (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision. A DEMO phase I with a 'Starter Blanket' and 'Starter Divertor' is foreseen: the blanket being capable of withstanding ⩾2 MW yr m-2 fusion neutron fluence (∼20 dpa in the front-wall steel). A second phase ensues for DEMO with ⩾5 MW yr m-2 first wall neutron fluence. Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction with functional materials/coolants; and a project-based risk analysis, with R&D to mitigate risks from material shortcomings including development of specific risk mitigation materials. The DEMO balance of plant constrains the blanket and divertor coolants to remain unchanged between the two phases. The blanket coolant choices (He gas or pressurised water) put technical constraints on the blanket steels, either to have high strength at higher temperatures than current baseline variants (above 650 °C for high thermodynamic efficiency from He-gas coolant), or superior radiation-embrittlement properties at lower temperatures (∼290-320 °C), for construction of water-cooled blankets. Risk mitigation proposed would develop these options in parallel, and computational and modelling techniques to shorten the cycle-time of new steel development will be important to achieve tight R&D timescales. The superior power handling of a water-cooled divertor target suggests a substructure temperature operating window (∼200-350 °C) that could be realised, as a baseline-concept, using tungsten on a copper-alloy substructure. The difficulty of establishing design codes for brittle tungsten puts great urgency on the development of a range of advanced ductile or strengthened tungsten and copper compounds. Lessons learned from Fission reactor material development have been included, especially in safety and licensing, fabrication/joining techniques and designing for in-vessel inspection. The technical basis of using the ITER licensing experience to refine the issues in nuclear testing of materials is discussed. Testing with 14 MeV neutrons is essential to Fusion Materials development, and the Roadmap requires acquisition of ⩾30 dpa (steels) 14 MeV test data by 2026. The value and limits of pre-screening testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams are evaluated to help determine the minimum14 MeV testing programme requirements.

A facility for testing 10 to 100-kWe space power reactors

NASA Astrophysics Data System (ADS)

Carlson, William F.; Bitten, Ernest J.

1993-01-01

This paper describes an existing facility that could be used in a cost-effective manner to test space power reactors in the 10 to 100-kWe range before launch. The facility has been designed to conduct full power tests of 100-kWe SP-100 reactor systems and already has the structural features that would be required for lower power testing. The paper describes a reasonable scenario starting with the acceptance at the test site of the unfueled reactor assembly and the separately shipped nuclear fuel. After fueling the reactor and installing it in the facility, cold critical tests are performed, and the reactor is then shipped to the launch site. The availability of this facility represents a cost-effective means of performing the required prelaunch test program.

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

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

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

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

Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

NASA Technical Reports Server (NTRS)

Sibille, Laurent; Dominguez, Jesus A.

2012-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a Joule-heated (sometimes called 'self-heating') reactor in which the electrolytic currents generate enough Joule heat to create a molten bath. Solutions obtained by multiphysics modeling allow the identification of the critical dimensions of concept reactors.

Reactor concepts for bioelectrochemical syntheses and energy conversion.

PubMed

Krieg, Thomas; Sydow, Anne; Schröder, Uwe; Schrader, Jens; Holtmann, Dirk

2014-12-01

In bioelectrochemical systems (BESs) at least one electrode reaction is catalyzed by microorganisms or isolated enzymes. One of the existing challenges for BESs is shifting the technology towards industrial use and engineering reactor systems at adequate scales. Due to the fact that most BESs are usually deployed in the production of large-volume but low-value products (e.g., energy, fuels, and bulk chemicals), investment and operating costs must be minimized. Recent advances in reactor concepts for different BESs, in particular biofuel cells and electrosynthesis, are summarized in this review including electrode development and first applications on a technical scale. A better understanding of the impact of reactor components on the performance of the reaction system is an important step towards commercialization of BESs. Copyright © 2014 Elsevier Ltd. All rights reserved.

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.

TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR ...

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

TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR AREAS SOUTH OF PERCH AVENUE. "COLD" SERVICES NORTH OF PERCH. ADVANCED TEST REACTOR IN NEW SECTION WEST OF COLD SERVICES SECTION. NEW PERIMETER FENCE ENCLOSES BETA RAY SPECTROMETER, TRA-669, AN ATR SUPPORT FACILITY, AND ATR STACK. UTM LOCATORS HAVE BEEN DELETED. IDAHO NUCLEAR CORPORATION, FROM A BLAW-KNOX DRAWING, 3/1968. INL INDEX NO. 530-0100-00-400-011646, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Advanced thermally stable jet fuels. Technical progress report, January 1995--March 1995

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

Schobert, H.H.; Eser, S.; Song, C.

Quantitative structure-property relationships have been applied to study the thermal stability of pure hydrocarbons typical of jet fuel components. A simple method of chemical structure description in terms of Benson groups was tested in searching for structure-property relationships for the hydrocarbons tested experimentally in this program. Molecular connectivity as a structure-based approach to chemical structure-property relationship analysis was also tested. Further development of both the experimental data base and computational methods will be necessary. Thermal decomposition studies, using glass tube reactors, were extended to two additional model compounds: n-decane and n-dodecane. Efforts on refining the deposit growth measurement and characterizationmore » of suspended matter in stressed fuels have lead to improvements in the analysis of stressed fuels. Catalytic hydrogenation and dehydrogenation studies utilizing a molybdenum sulfide catalyst are also described.« less

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

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

Forsbacka, M.; Moore, M.

1989-12-01

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

NPR design basis

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

Locke, G.L.

1958-09-08

The design basis is composed of requirements and conditions for the design of the reactor plant (composed of the reactor and heat dissipation system). Its intent is to insure that the final product meets the economic, safety, and technical objectives of the project. The design basis is dependent on the ground rules, objectives, technical criteria, and practical design considerations. This document is being issued with the understanding that these items are not yet firmly established in all respects, and therefore, the numbers put down here are subject to change. Consideration of the spectrum of probable changes that might be mademore » leads to the conclusion that the numbers here are close to the final ones and are satisfactory as a basis for the initial stages of design. Some numbers are omitted because of insufficient data at this time.« less

The nuclear battery

NASA Astrophysics Data System (ADS)

Kozier, K. S.; Rosinger, H. E.

The evolution and present status of an Atomic Energy of Canada Limited program to develop a small, solid-state, passively cooled reactor power supply known as the Nuclear Battery is reviewed. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy, which may be converted into electricity in an organic Rankine cycle engine or used to produce high-pressure steam. The present reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years. Technical and safety features are described along with recent progress in component hardware development programs and market assessment work.

Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

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

Sienicki, James J.; Grandy, Christopher

A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. Themore » various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.« less

Toxoplasmosis serology: an efficient hemagglutination procedure to detect IgG and IgM antibodies.

PubMed

Camargo, M E; Moura, M E; Leser, P G

1989-01-01

In search of an efficient but simple, low cost procedure for the serodiagnosis of Toxoplasmosis, especially suited for routine laboratories facing technical and budget limitations as in less developed countries, the diagnostic capability of Hematoxo, an hemagglutination test for toxoplasmosis, was evaluated in relation to a battery of tests including IgG- and IgM-immunofluorescence tests, hemagglutination and an IgM-capture enzymatic assay. Detecting a little as 5 I.U. of IgG antitoxoplasma antibodies, Hematoxo showed a straight agreement as to reactivity and non-reactivity for the 443 non-reactive and the 387 reactive serum samples, included in this study. In 23 cases presenting a serological pattern of acute toxoplasmosis and showing IgM antibodies, Hematoxo could detect IgM antibodies in 18, indicated by negativation or a significant decrease in titers as a result of treating samples with 2-mercapto-ethanol. However, a neat increase in sensitivity for IgM specific antibodies could be achieved by previously removing IgG from the sample, as demonstrated in a series of acute toxoplasmosis sera. A simple procedure was developed for this purpose, by reconstituting a lyophilized suspension of Protein A--rich Staphylococcus with the lowest serum dilution to be tested. Of low cost and easy to perform, Hematoxo affords not only a practical qualitative procedure for screening reactors and non-reactors, as in prenatal services, but also quantitative assays that permit to titrate antibodies as well as to identify IgM antibodies.

Run 263 with Black Thunder Mine subbituminous coal and dispersed molybdenum catalysts. Technical progress report

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

Not Available

This report presents the results of Run 263 performed at the Advanced Coal Liquefaction R&D Facility in Wilsonville, Alabama. The run started on October 31, 1991 and continued until February 23, 1992. Tests were conducted by operating the reactors in the Close-Coupled Integrated Two-Stage Liquefaction mode and by processing Black Thunder Mine subbituminous coal from Wyodak-Anderson seam in Wyoming Powder River Basin. Half volume reactors were used for the entire run. In the first part of Run 263, a dispersed molybdenum catalyst was evaluated for its performance without a supported catalyst in the second stage. Molyvan L and Molyvan 822more » (commercially available as friction reducing lubricants) were used as precursors for the dispersed molybdenum catalyst. The effect of the dispersed catalyst on eliminating the solids buildup was also evaluated. For the second part of the run, the hybrid catalyst system was tested with supported Criterion 324 1/1611 catalyst in the second stage at catalyst replacement rates of 2 and 3 lb/ton of MF coal. The molybdenum concentration was 100--200 ppm based on MF coal. Iron oxide was used as a slurry catalyst precursor at a rate of 1--2 wt % MF coal throughout the run with dimethyl disulfide (DMDS) as the sulfiding agent. The close-coupled reactor unit was on-stream for 2482 hours for an on-stream factor of 91.2% and the ROSE-SR{sup sm} unit was on-feed for 2126 hours for an on-stream factor of 96.4% for the entire run.« 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.

Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

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

Chang, L.K.; Mohr, D.; Planchon, H.P.

This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« 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. .

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.

Production of Hydrogen by Superadiabatic Decomposition of Hydrogen Sulfide - Final Technical Report for the Period June 1, 1999 - September 30, 2000

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

Rachid B. Slimane; Francis S. Lau; Javad Abbasian

2000-10-01

The objective of this program is to develop an economical process for hydrogen production, with no additional carbon dioxide emission, through the thermal decomposition of hydrogen sulfide (H{sub 2}S) in H{sub 2}S-rich waste streams to high-purity hydrogen and elemental sulfur. The novel feature of the process being developed is the superadiabatic combustion (SAC) of part of the H{sub 2}S in the waste stream to provide the thermal energy required for the decomposition reaction such that no additional energy is required. The program is divided into two phases. In Phase 1, detailed thermochemical and kinetic modeling of the SAC reactor withmore » H{sub 2}S-rich fuel gas and air/enriched air feeds is undertaken to evaluate the effects of operating conditions on exit gas products and conversion efficiency, and to identify key process parameters. Preliminary modeling results are used as a basis to conduct a thorough evaluation of SAC process design options, including reactor configuration, operating conditions, and productivity-product separation schemes, with respect to potential product yields, thermal efficiency, capital and operating costs, and reliability, ultimately leading to the preparation of a design package and cost estimate for a bench-scale reactor testing system to be assembled and tested in Phase 2 of the program. A detailed parametric testing plan was also developed for process design optimization and model verification in Phase 2. During Phase 2 of this program, IGT, UIC, and industry advisors UOP and BP Amoco will validate the SAC concept through construction of the bench-scale unit and parametric testing. The computer model developed in Phase 1 will be updated with the experimental data and used in future scale-up efforts. The process design will be refined and the cost estimate updated. Market survey and assessment will continue so that a commercial demonstration project can be identified.« less

Sodium Handling Technology and Engineering Design of the Madison Dynamo Experiment.

NASA Astrophysics Data System (ADS)

Kendrick, R.; Forest, C. B.; O'Connell, R.; Wright, A.; Robinson, K.

1998-11-01

A new liquid metal MHD experiment is being constructed at the University of Wisconsin to test several key predictions of dynamo theory: magnetic instabilities driven by sheared flow, the effects of turbulence on current generation, and the back-reaction of the self-generated magnetic field on the fluid motion which brings saturation. This presentation describes the engineering design of the experiment, which is a 0.5 m radius spherical vessel, filled with liquid sodium at 150 degrees Celsius. The experiment is designed to achieve a magnetic Reynolds number in excess of 100, which requires approximately 80 Hp of mechanical drive, producing flow velocities in sodium of 15 m/s through impellers. Handling liquid sodium offers a number of technical challenges, but routine techniques have been developed over the past several decades for safely handling large quantities for the fast breeder reactor. The handling strategy is discussed, technical details concerning seals and pressurazation are presented, and safety elements are highlighted.

Initial operation with sodium in the Madison Dynamo Experiment.

NASA Astrophysics Data System (ADS)

Kendrick, R.; Spence, Ej; Forest, C. B.; O'Connell, R.; Nornberg, Md; Canary, Hw; Wright, A.; Robinson, K.

1999-11-01

A new liquid metal MHD experiment has been constructed at the University of Wisconsin to test several key predictions of dynamo theory: magnetic instabilities driven by sheared flow, the effects of turbulence on current generation, and the back-reaction of the self-generated magnetic field on the fluid motion which brings saturation. This presentation describes the engineering design of the experiment, which is a 0.5 m radius spherical vessel, filled with liquid sodium at 150 ^circC. The experiment is designed to achieve a magnetic Reynolds number in excess of 100, which requires approximately 80 Hp of mechanical drive, producing flow velocities in sodium of 15 m/s through impellers. Handling liquid sodium offers a number of technical challenges, but routine techniques have been developed over the past several decades for safely handling large quantities for the fast breeder reactor. The handling strategy is discussed, technical details concerning seals and pressurization are presented, and safety elements are highlighted.

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

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

Nuclear reactors built, being built, or planned: 1995

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

NONE

1996-08-01

This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. 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: (1) a commercial reactor locator map and tables of the characteristicmore » 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. 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

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

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

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Recirculation of reverse osmosis concentrate in lab-scale anaerobic and aerobic landfill simulation reactors.

PubMed

Morello, Luca; Cossu, Raffaello; Raga, Roberto; Pivato, Alberto; Lavagnolo, Maria Cristina

2016-10-01

Leachate treatment is a major issue in the context of landfill management, particularly in view of the consistent changes manifested over time in the quality and quantity of leachate produced, linked to both waste and landfill characteristics, which renders the procedure technically difficult and expensive. Leachate recirculation may afford a series of potential advantages, including improvement of leachate quality, enhancement of gas production, acceleration of biochemical processes, control of moisture content, as well as nutrients and microbe migration within the landfill. Recirculation of the products of leachate treatment, such as reverse osmosis (RO) concentrate, is a less common practice, with widespread controversy relating to its suitability, potential impacts on landfill management and future gaseous and leachable emissions. Scientific literature provides the results of only a few full-scale applications of concentrate recirculation. In some cases, an increase of COD and ammonium nitrogen in leachate was observed, coupled with an increase of salinity; which, additionally, might negatively affect performance of the RO plant itself. In other cases, not only did leachate production not increase significantly but the characteristics of leachate extracted from the well closest to the re-injection point also remained unchanged. This paper presents the results of lab-scale tests conducted in landfill simulation reactors, in which the effects of injection of municipal solid waste (MSW) landfill leachate RO concentrate were evaluated. Six reactors were managed with different weekly concentrate inputs, under both anaerobic and aerobic conditions, with the aim of investigating the short and long-term effects of this practice on landfill emissions. Lab-scale tests resulted in a more reliable identification of compound accumulation and kinetic changes than full-scale applications, further enhancing the development of a mass balance in which gaseous emissions and waste characteristics were also taken into consideration. Results showed that RO concentrate recirculation did not produce consistent changes in COD emissions and methane production. Simultaneously, ammonium ion showed a consistent increase in leachate (more than 25%) in anaerobic reactors, free ammonia gaseous emissions doubled with concentrate injection, while chloride resulted accumulated inside the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA603. CONTEXTUAL VIEW OF MTR ...

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

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA-603. CONTEXTUAL VIEW OF MTR BUILDING SHOWING NORTH SIDES OF THE HIGH-BAY REACTOR BUILDING, ITS SECOND/THIRD FLOOR BALCONY LEVEL, AND THE ATTACHED ONE-STORY OFFICE/LABORATORY BUILDING, TRA-604. CAMERA FACING SOUTHEAST. VERTICAL CONCRETE-SHROUDED BEAMS SUPPORT PRECAST CONCRETE PANELS. CONCRETE PROJECTION FORMED AS A BUNKER AT LEFT OF VIEW IS TRA-657, PLUG STORAGE BUILDING. INL NEGATIVE NO. HD46-42-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Functional materials for breeding blankets—status and developments

NASA Astrophysics Data System (ADS)

Konishi, S.; Enoeda, M.; Nakamichi, M.; Hoshino, T.; Ying, A.; Sharafat, S.; Smolentsev, S.

2017-09-01

The development of tritium breeder, neutron multiplier and flow channel insert materials for the breeding blanket of the DEMO reactor is reviewed. Present emphasis is on the ITER test blanket module (TBM); lithium metatitanate (Li2TiO3) and lithium orthosilicate (Li4SiO4) pebbles have been developed by leading TBM parties. Beryllium pebbles have been selected as the neutron multiplier. Good progress has been made in their fabrication; however, verification of the design by experiments is in the planning stage. Irradiation data are also limited, but the decrease in thermal conductivity of beryllium due to irradiation followed by swelling is a concern. Tests at ITER are regarded as a major milestone. For the DEMO reactor, improvement of the breeder has been attempted to obtain a higher lithium content, and Be12Ti and other beryllide intermetallic compounds that have superior chemical stability have been studied. LiPb eutectic has been considered as a DEMO blanket in the liquid breeder option and is used as a coolant to achieve a higher outlet temperature; a SiC flow channel insert is used to prevent magnetohydrodynamic pressure drop and corrosion. A significant technical gap between ITER TBM and DEMO is recognized, and the world fusion community is working on ITER TBM and DEMO blanket development in parallel.

WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. ...

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

WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. INL NEGATIVE NO. 3925. Unknown Photographer, 12/14/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Eddy Current Flow Measurements in the FFTF

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

Nielsen, Deborah L.; Polzin, David L.; Omberg, Ronald P.

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuablemore » information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.« less

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

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

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

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

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

Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

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

Hallbert, Bruce Perry; Thomas, Kenneth David

2015-10-01

Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

Chernobyl Doses. Volume 3. Habitat and Vegetation Near the Chernobyl Nuclear Reactor Station

DTIC Science & Technology

1993-01-01

AD-A260 167 A lexandria, VA 22310-3398 l,* Defense Nuclear Agency Alexandria, VA 22310-.3398 DNA-TR-92-37-V3 Chernobyl Doses, Volume 3-Habitat and...Vegetation Near the Chernobyl Nuclear Reactor Station DTIC~ ELECTF. Elizabeth L. Painter i IN•9 199EIF F. Ward Whicker JAN % 93f Pacific-Sierra...930101 Technical 870929- 920228 4. TITLE AND SUBTITLE 5. FUNDING NUMBERS Chernobyl Doses C - DNA 001-87-C-0104 Volume 3-Habitat and Vegetation Near the

A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

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

Rizwan-uddin; Nick Karancevic; Stefano Markidis

2008-04-23

many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

ENGINEERING TEST REACTOR (ETR) BUILDING, TRA642. CONTEXTUAL VIEW, CAMERA FACING ...

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

ENGINEERING TEST REACTOR (ETR) BUILDING, TRA-642. CONTEXTUAL VIEW, CAMERA FACING EAST. VERTICAL METAL SIDING. ROOF IS SLIGHTLY ELEVATED AT CENTER LINE FOR DRAINAGE. WEST SIDE OF ETR COMPRESSOR BUILDING, TRA-643, PROJECTS TOWARD LEFT AT FAR END OF ETR BUILDING. INL NEGATIVE NO. HD46-37-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

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

Rice, R.E.

Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Ramuhalli, Pradeep; Vlim, R.

2016-10-01

Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributesmore » to the design and implementation of AdvRx concepts.« less

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

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

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

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

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

1993-01-01

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

SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

ScienceCinema

None

2018-01-16

SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

Mass tracking and material accounting in the integral fast reactor (IFR)

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

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstratedmore » in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.« less

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

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

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01

This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operatingmore » experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.« less

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.

Technical Aspects Regarding the Management of Radioactive Waste from Decommissioning of Nuclear Facilities

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

Dragolici, F.; Turcanu, C. N.; Rotarescu, G.

2003-02-25

The proper application of the nuclear techniques and technologies in Romania started in 1957, once with the commissioning of the Research Reactor VVR-S from IFIN-HH-Magurele. During the last 45 years, appear thousands of nuclear application units with extremely diverse profiles (research, biology, medicine, education, agriculture, transport, all types of industry) which used different nuclear facilities containing radioactive sources and generating a great variety of radioactive waste during the decommissioning after the operation lifetime is accomplished. A new aspect appears by the planning of VVR-S Research Reactor decommissioning which will be a new source of radioactive waste generated by decontamination, disassemblingmore » and demolition activities. By construction and exploitation of the Radioactive Waste Treatment Plant (STDR)--Magurele and the National Repository for Low and Intermediate Radioactive Waste (DNDR)--Baita, Bihor county, in Romania was solved the management of radioactive wastes arising from operation and decommissioning of small nuclear facilities, being assured the protection of the people and environment. The present paper makes a review of the present technical status of the Romanian waste management facilities, especially raising on treatment capabilities of ''problem'' wastes such as Ra-266, Pu-238, Am-241 Co-60, Co-57, Sr-90, Cs-137 sealed sources from industrial, research and medical applications. Also, contain a preliminary estimation of quantities and types of wastes, which would result during the decommissioning project of the VVR-S Research Reactor from IFIN-HH giving attention to some special category of wastes like aluminum, graphite and equipment, components and structures that became radioactive through neutron activation. After analyzing the technical and scientific potential of STDR and DNDR to handle big amounts of wastes resulting from the decommissioning of VVR-S Research Reactor and small nuclear facilities, the necessity of up-gradation of these nuclear objectives before starting the decommissioning plan is revealed. A short presentation of the up-grading needs is also presented.« less

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

NASA Astrophysics Data System (ADS)

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

2008-10-01

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

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

77 FR 63897 - Notice of License Terminations for National Aeronautics and Space Administration; Plum Brook...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-17

... test reactor, constructed to perform irradiation testing of fueled and unfueled experiments for space... constructed to test ``mock-up'' irradiation components for the Plum Brook Reactor. The reactors operated from...

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.

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

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

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

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

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.

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

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

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

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

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

Donald Rhodes, left, and Clyde Greer, right, monitor the operation of the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility from the control room. The 60-megawatt test reactor, NASA’s only reactor, was the eighth largest test reactor in the world. The facility was built by the Lewis Research Center in the late 1950s to study the effects of radiation on different materials that could be used to construct nuclear propulsion systems for aircraft or rockets. The reactor went critical for the first time in 1961. For the next two years, two operators were on duty 24 hours per day working on the fission process until the reactor reached its full-power level in 1963. Reactor Operators were responsible for monitoring and controlling the reactor systems. Once the reactor was running under normal operating conditions, the work was relatively uneventful. Normally the reactor was kept at a designated power level within certain limits. Occasionally the operators had to increase the power for a certain test. The shift supervisor and several different people would get together and discuss the change before boosting the power. All operators were required to maintain a Reactor Operator License from the Atomic Energy Commission. The license included six months of training, an eight-hour written exam, a four-hour walkaround, and testing on the reactor controls.

Final Report for the “WSU Neutron Capture Therapy Facility Support”

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

Gerald E. Tripard; Keith G. Fox

2006-08-24

The objective for the cooperative research program for which this report has been written was to provide separate NCT facility user support for the students, faculty and scientists who would be doing the U.S. Department of Energy Office (DOE) of Science supported advanced radiotargeted research at the WSU 1 megawatt TRIGA reactor. The participants were the Idaho National laboratory (INL, P.I., Dave Nigg), the Veterinary Medical Research Center of Washington State University (WSU, Janean Fidel and Patrick Gavin), and the Washington State University Nuclear Radiation Center (WSU, P.I., Gerald Tripard). A significant number of DOE supported modifications were made tomore » the WSU reactor in order to create an epithermal neutron beam while at the same time maintaining the other activities of the 1 MW reactor. These modifications were: (1) Removal of the old thermal column. (2) Construction and insertion of a new epithermal filter, collimator and shield. (3) Construction of a shielded room that could accommodate the very high radiation field created by an intense neutron beam. (4) Removal of the previous reactor core fuel cluster arrangement. (5) Design and loading of the new reactor core fuel cluster arrangement in order to optimize the neutron flux entering the epithermal neutron filter. (6) The integration of the shielded rooms interlocks and radiological controls into the SCRAM chain and operating electronics of the reactor. (7) Construction of a motorized mechanism for moving and remotely controlling the position of the entire reactor bridge. (8) The integration of the reactor bridge control electronics into the SCRAM chain and operating electronics of the reactor. (9) The design, construction and attachment to the support structure of the reactor of an irradiation box that could be inserted into position next to the face of the reactor. (Necessitated by the previously mentioned core rearrangement). All of the above modifications were successfully completed and tested. The resulting epithermal beam of 1 x 10{sup 9} n/sec-cm{sup 2} was measured by Idaho National Laboratory with assistance from WSU's Neutron Activation Analysis Group. The beam is as good as our initial proposals for the project had predicted. In addition to all of the design, construction and insertion of the hardware, shielding, electronics and radiation monitoring systems there was considerable manpower and effort put into changes in the Technical Specifications of the reactor and implementing procedures for use of the new facility. This staff involvement is one of the reasons we requested special facility support from the DOE. Once the facility was competed and all of the recalibrations and measurements made to characterize the differences between this reactor core and the previous core we began to assist INL in making their beam measurements with foils and phantoms. Although we proposed support for only one additional staff position to support this new NCT facility the staff support provided by the WSU Nuclear Radiation Center was greater than had been anticipated by our initial proposal. INL was also assisted in the testing of a heavy water (deuterated water) bladder that can be inserted into the collimator in order to produce an intense, external thermal neutron beam. The external epithermal and/or thermal neutron beam capability remains available for use, if funding becomes available for future research projects.« less

Space reactor fuel element testing in upgraded TREAT

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

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

1993-01-14

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

Space reactor fuel element testing in upgraded TREAT

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

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

1993-05-01

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

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

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

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

Oxidation of aluminum alloy cladding for research and test reactor fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

SAFARI-1: Achieving conversion to LEU - A local challenge

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

Piani, C.S.B.

2008-07-15

Two years have passed since the South African Department of Minerals and Energy authorised the conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the South African Research Reactor (SAFARI-1) and the associated fuel manufacturing at Pelindaba. The scheduling, as originally proposed, allowed approximately three years for the full conversion of the reactor, anticipating simultaneous manufacturing ability from the fuel production plant. Due to technical difficulties experienced in the conversion of the local manufacturing plant from HEU (UAl alloy) to LEU (U Silicide) and the uncertainty as to costing and scheduling of such an achievement, the conversionmore » of SAFARI-1 based on local supply has been allocated a lower priority. The acquisition in mid-2006 of 2 LEU silicide elements of SA design, manufactured by AREVA- CERCA and irradiated as test elements in SAFARI-1 to burn-ups of {approx}65% each; was successfully accomplished within 9 cycles of irradiation each. Furthermore, four 'Hybrid' elements (AREVA-CERCA plates assembled locally at Pelindaba) are ready for irradiation and have received regulatory authorisation to load. This will enable the SAFARI-1 conversion program to continue systematically according to an agreed schedule. This paper will trace the developments of the above and reflect the current status and the rescheduled conversion phases of the reactor according to latest expectations. (author)« less

Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

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

Sullivan, Edmund J.; Anderson, Michael T.

2014-06-10

This technical letter report provides the status of an assessment undertaken by PNNL at the request of the NRC to verify the capability of periodic ASME-required volumetric examinations of reactor vessels to characterize the density and distribution of flaws of interest for applying §50.61a on a plant-by-plant basis. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. Recently, themore » NRC completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed and promulgated an alternate PTS rule, §50.61a, that can be implemented by PWR licensees. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants.« less

SNAP 10A FS-3 reactor performance

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

Hawley, J.P.; Johnson, R.A.

1966-08-15

SNAP 10FS-3 was the first flight-qualified SNAP reactor system to be operated in a simulated space environment. Prestart-up qualification testing, automatic start-up, endurance period performance, extended operation test and reactor shutdown are described as they affected, or were affected by, overall reactor performance. Performance of the reactor control system and the diagnostic instrumentation is critically evaluted.

Closeout Report for the Refractory Metal Accelerated Heat Pipe Life Test Activity

NASA Technical Reports Server (NTRS)

Martin, J.; Reid, R.; Stewart, E.; Hickman, R.; Mireles, O.

2013-01-01

With the selection of a gas-cooled reactor, this heat pipe accelerated life test activity was closed out and its resources redirected. The scope of this project was to establish the long-term aging effects on Mo-44.5%Re sodium heat pipes when subjected to space reactor temperature and mass fluences. To date, investigators have demonstrated heat pipe life tests of alkali metal systems up to .50,000 hours. Unfortunately, resources have not been available to examine the effect of temperature, mass fluence, or impurity level on corrosion or to conduct post-test forensic examination of heat pipes. The key objective of this effort was to establish a cost/time effective method to systematically test alkali metal heat pipes with both practical and theoretical benefits. During execution of the project, a heat pipe design was established, a majority of the laboratory test equipment systems specified, and operating and test procedures developed. Procurements for the heat pipe units and all major test components were underway at the time the stop work order was issued. An extremely important outcome was the successful fabrication of an annular wick from Mo-5%Re screen (the single, most difficult component to manufacture) using a hot isostatic pressing technique. This Technical Publication (TP) includes specifics regarding the heat pipe calorimeter water-cooling system, vendor design for the radio frequency heating system, possible alternative calorimeter designs, and progress on the vanadium equilibration technique. The methods provided in this TP and preceding project documentation would serve as a good starting point to rapidly implement an accelerated life test. Relevant test data can become available within months, not years, and destructive examination of the first life test heat pipe might begin within 6 months of test initiation. Final conclusions could be drawn in less than a quarter of the mission duration for a long-lived, fission-powered, deep space probe.

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

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

Exploratory evaluation of ceramics for automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1972-01-01

An exploratory evaluation of ceramics for automobile thermal reactors was conducted. Potential ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance lasting over 800 hours in engine dynamometer tests and over 15,000 miles (24,200 km) of vehicle road tests. Reactors containing glass-ceramic components did not perform as well as silicon carbide. But the glass-ceramics still offer good potential for reactor use. The results of this study are considered to be a reasonable demonstration of the potential use of ceramics in thermal reactors.

Realizing "2001: A Space Odyssey": Piloted Spherical Torus Nuclear Fusion Propulsion

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Dudzinski, Leonard A.; Borowski, Stanley K.; Juhasz, Albert J.

2005-01-01

A conceptual vehicle design enabling fast, piloted outer solar system travel was created predicated on a small aspect ratio spherical torus nuclear fusion reactor. The initial requirements were satisfied by the vehicle concept, which could deliver a 172 mt crew payload from Earth to Jupiter rendezvous in 118 days, with an initial mass in low Earth orbit of 1,690 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including artificial gravity payload, central truss, nuclear fusion reactor, power conversion, magnetic nozzle, fast wave plasma heating, tankage, fuel pellet injector, startup/re-start fission reactor and battery bank, refrigeration, reaction control, communications, mission design, and space operations. Detailed fusion reactor design included analysis of plasma characteristics, power balance/utilization, first wall, toroidal field coils, heat transfer, and neutron/x-ray radiation. Technical comparisons are made between the vehicle concept and the interplanetary spacecraft depicted in the motion picture 2001: A Space Odyssey.

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

PBF Reactor Building (PER620). PBF crane holds fuel test assembly ...

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

PBF Reactor Building (PER-620). PBF crane holds fuel test assembly aloft prior to lowering into reactor for test. Date: 1982. INEEL negative no. 82-4909 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Multiphysics Modeling for Dimensional Analysis of a Self-Heated Molten Regolith Electrolysis Reactor for Oxygen and Metals Production on the Moon and Mars

NASA Technical Reports Server (NTRS)

Dominguez, Jesus; Sibille, Laurent

2010-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a self-heating reactor in which the electrolytic currents generate enough Joule heat to create a molten bath.

A Comparison Framework for Reactor Anti-Neutrino Detectors in Near-Field Nuclear Safeguards Applications

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

Mendenhall, M.; Bowden, N.; Brodsky, J.

Electron anti-neutrino ( e) detectors can support nuclear safeguards, from reactor monitoring to spent fuel characterization. In recent years, the scientific community has developed multiple detector concepts, many of which have been prototyped or deployed for specific measurements by their respective collaborations. However, the diversity of technical approaches, deployment conditions, and analysis techniques complicates direct performance comparison between designs. We have begun development of a simulation framework to compare and evaluate existing and proposed detector designs for nonproliferation applications in a uniform manner. This report demonstrates the intent and capabilities of the framework by evaluating four detector design concepts, calculatingmore » generic reactor antineutrino counting sensitivity, and capabilities in a plutonium disposition application example.« less

KWU's high conversion reactor concept - An economical evolution of modern pressurized water reactor technology toward improved uranium ore utilization

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

Markl, H.; Goetzmann, C.A.; Moldaschl, H.

The Kraftwerk Union AG high conversion reactor represents a quasi-standard PWR with fuel assemblies of more or less uniformly enriched fuel rods, arranged in a tight hexagonal array with a pitch-to-diameter ratio p/d approx. = 1.12. High fuel enrichment as well as a high conversion ratio of --0.9 will provide the potential for high burnup values up to 70 000 MWd/tonne and a low fissile material consumption. The overall objective of the actual RandD program is to have the technical feasibility, including that for licensibility, established by the early 1990s as a prerequisite for deciding whether to enter a demonstrationmore » plant program.« less

Diverse knowledges and competing interests: an essay on socio-technical problem-solving.

PubMed

di Norcia, Vincent

2002-01-01

Solving complex socio-technical problems, this paper claims, involves diverse knowledges (cognitive diversity), competing interests (social diversity), and pragmatism. To explain this view, this paper first explores two different cases: Canadian pulp and paper mill pollution and siting nuclear reactors in systematically sensitive areas of California. Solving such socio-technically complex problems involves cognitive diversity as well as social diversity and pragmatism. Cognitive diversity requires one to not only recognize relevant knowledges but also to assess their validity. Finally, it is suggested, integrating the resultant set of diverse relevant and valid knowledges determines the parameters of the solution space for the problem.

Index to Nuclear Safety: a technical progress review by chronology, permuted title, and author, Volume 18 (1) through Volume 22 (6)

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

Cottrell, W.B.; Passiakos, M.

This index to Nuclear Safety covers articles published in Nuclear Safety, Volume 18, Number 1 (January-February 1977) through Volume 22, Number 6 (November-December 1981). The index is divided into three section: a chronological list of articles (including abstracts), a permuted-title (KWIC) index, and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. Over 300 technical articles published in Nuclear Safety in the last 5 years are listed in this index.

76 FR 31381 - Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-31

...). The current revision issues a new Branch Technical Position (BTP) 8-8 on ``Onsite (Emergency Diesel Generators) and Offsite Power Sources Allowed Outage Time Extensions.'' (ADAMS Accession No. ML111180521...

A pilot level decision analysis of thermionic reactor development strategy for nuclear electric propulsion

NASA Technical Reports Server (NTRS)

Menke, M. M.; Judd, B. R.

1973-01-01

The development policy for thermionic reactors to provide electric propulsion and power for space exploration was analyzed to develop a logical procedure for selecting development alternatives that reflect the technical feasibility, JPL/NASA project objectives, and the economic environment of the project. The partial evolution of a decision model from the underlying philosophy of decision analysis to a deterministic pilot phase is presented, and the general manner in which this decision model can be employed to examine propulsion development alternatives is illustrated.

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

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

McConnell, Paul E.; Ross, Steven; Grey, Carissa Ann

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacificmore » Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.« less

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

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

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang

2006-10-01

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progressmore » toward element testing will be reviewed.« less

Preliminary Options Assessment of Versatile Irradiation Test Reactor

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

Sen, Ramazan Sonat

The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 xmore » 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.« less

Space station prototype Sabatier reactor design verification testing

NASA Technical Reports Server (NTRS)

Cusick, R. J.

1974-01-01

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

CRITICAL EXPERIMENT TANK (CET) REACTOR HAZARDS SUMMARY

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

Becar, N.J.; Kunze, J.F.; Pincock, G..D.

1961-03-31

The Critical Experiment Tank (CET) reactor assembly, the associated systems, and the Low Power Test Facility in which the reactor is to be operated are described. An evaluation and summary of the hazards associated with the operation of the CET reactor in the LPTF at the ldsho Test Station are also presented. (auth)

ETRCF, TRA654, INTERIOR. REACTOR OPERATED IN WATERFILLED TANK. CAMERA LOOKS ...

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

ETR-CF, TRA-654, INTERIOR. REACTOR OPERATED IN WATER-FILLED TANK. CAMERA LOOKS DOWN FROM ABOVE UPON LATER (NON-NUCLEAR) EXPERIMENTAL GEAR. INL NEGATIVE NO. HD24-1-1. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED ...

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

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED OUTSIDE OF MTR FOR EXPERIMENTS. THE AIRCRAFT NUCLEAR PROPULSION PROJECT DOMINATED THE USE OF THIS PART OF THE MTR. INL NEGATIVE NO. 7225. Unknown Photographer, 11/28/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Scaling Analysis Techniques to Establish Experimental Infrastructure for Component, Subsystem, and Integrated System Testing

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

Sabharwall, Piyush; O'Brien, James E.; McKellar, Michael G.

2015-03-01

Hybrid energy system research has the potential to expand the application for nuclear reactor technology beyond electricity. The purpose of this research is to reduce both technical and economic risks associated with energy systems of the future. Nuclear hybrid energy systems (NHES) mitigate the variability of renewable energy sources, provide opportunities to produce revenue from different product streams, and avoid capital inefficiencies by matching electrical output to demand by using excess generation capacity for other purposes when it is available. An essential step in the commercialization and deployment of this advanced technology is scaled testing to demonstrate integrated dynamic performancemore » of advanced systems and components when risks cannot be mitigated adequately by analysis or simulation. Further testing in a prototypical environment is needed for validation and higher confidence. This research supports the development of advanced nuclear reactor technology and NHES, and their adaptation to commercial industrial applications that will potentially advance U.S. energy security, economy, and reliability and further reduce carbon emissions. Experimental infrastructure development for testing and feasibility studies of coupled systems can similarly support other projects having similar developmental needs and can generate data required for validation of models in thermal energy storage and transport, energy, and conversion process development. Experiments performed in the Systems Integration Laboratory will acquire performance data, identify scalability issues, and quantify technology gaps and needs for various hybrid or other energy systems. This report discusses detailed scaling (component and integrated system) and heat transfer figures of merit that will establish the experimental infrastructure for component, subsystem, and integrated system testing to advance the technology readiness of components and systems to the level required for commercial application and demonstration under NHES.« less

Current status of nuclear engineering education

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

Palladino, N.J.

1975-09-01

The 65 colleges and universities offering undergraduate degrees in nuclear engineering and the 15 schools offering strong nuclear engineering options are, in general, doing a good job to meet the current spectrum of job opportunities. But, nuclear engineering programs are not producing enough graduates to meet growing demands. They currently receive little aid and support from their customers --industry and government--in the form of scholarships, grants, faculty research support, student thesis and project support, or student summer jobs. There is not enough interaction between industry and universities. Most nuclear engineering programs are geared too closely to the technology of themore » present family of reactors and too little to the future breeder reactors and controlled thermonuclear reactors. In addition, nuclear engineering programs attract too few women and members of minority ethnic groups. Further study of the reasons for this fact is needed so that effective corrective action can be taken. Faculty in nuclear engineering programs should assume greater initiative to provide attractive and objective nuclear energy electives for technical and nontechnical students in other disciplines to improve their technical understanding of the safety and environmental issues involved. More aggressive and persistent efforts must be made by nuclear engineering schools to obtain industry support and involvement in their programs. (auth)« less

Technical, Economical, and Microbiological Aspects of the Microaerobic Process on H2S Removal for Low Sulfate Concentration Wastewaters.

PubMed

Sousa, M R; Oliveira, C J S; Lopes, A C; Rodríguez, E R; Holanda, G B M; Landim, P G C; Firmino, P I M; Dos Santos, A B

2016-12-01

We studied the feasibility of the microaerobic process, in comparison with the traditional chemical absorption process (NaOH), on H 2 S removal in order to improve the biogas quality. The experiment consisted of two systems: R1, biogas from an anaerobic reactor was washed in a NaOH solution, and R2, headspace microaeration with atmospheric air in a former anaerobic reactor. The microaeration used for low sulfate concentration wastewater did not affect the anaerobic digestion, but even increased system stability. Methane production in the R2 was 14 % lower compared to R1, due to biogas dilution by the atmospheric air used. The presence of oxygen in the biogas reveals that not all the oxygen was consumed for sulfide oxidation in the liquid phase indicating mass transfer limitations. The reactor was able to rapidly recover its capacity on H 2 S removal after an operational failure. Bacterial and archaeal richness shifted due to changes in operational parameters, which match with the system functioning. Finally, the microaerobic system seems to be more advantageous for both technical and economical reasons, in which the payback of microaerobic process for H 2 S removal was 4.7 months.

Integration of Cleaner Production and Waste Water Treatment on Tofu Small Industry for Biogas Production using AnSBR Reactor

NASA Astrophysics Data System (ADS)

Rahayu, Suparni Setyowati; Budiyono; Purwanto

2018-02-01

A research on developing a system that integrates clean production and waste water treatment for biogas production in tofu small industry has been conducted. In this research, tofu waste water was turned into biogas using an AnSBR reactor. Mud from the sewage system serves as the inoculums. This research involved: (1) workshop; (2) supervising; (3) technical meeting; (4) network meeting, and (5) technical application. Implementation of clean production integrated with waste water treatment reduced the amount of waste water to be treated in a treatment plant. This means less cost for construction and operation of waste water treatment plants, as inherent limitations associated with such plants like lack of fund, limited area, and technological issues are inevitable. Implementation of clean production prior to waste water treatment reduces pollution figures down to certain levels that limitations in waste water treatment plants can be covered. Results show that biogas in 16 days HRT in an AnSBR reactor contains CH4(78.26 %) and CO2 (20.16 %). Meanwhile, treatments using a conventional bio-digester result in biogas with 72.16 % CH4 and 18.12 % CO2. Hence, biogas efficiency for the AnSBR system is 2.14 times greater than that of a conventional bio-digester.

Small reactor power systems for manned planetary surface bases

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

VVER Reactor Safety in Eastern Europe and Former Soviet Union

NASA Astrophysics Data System (ADS)

Papadopoulou, Demetra

2012-02-01

VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

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

HEDL FACILITIES CATALOG 400 AREA

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

MAYANCSIK BA

1987-03-01

The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

Tory II-A: a nuclear ramjet test reactor

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

Hadley, J.W.

Declassified 28 Nov 1973. The first test reactor in the Pluto program, leading to development of a nuclear ramjet engine, is called Tory II-A. While it is not an actual prototype engine, this reactor embodies a core design which is considered feasible for an engine, and operation of the reactor will provide a test of that core type as well as more generalized values in reactor design and testing. The design of Tory II-A and construction of the reactor and of its test facility are described. Operation of the Tory II-A core at a total power of 160 megawatts, withmore » 800 pounds of air per second passing through the core and emerging at a temperature of 2000 deg F, is the central objective of the test program. All other reactor and facility components exist to support operation of the core, and preliminary steps in the test program itself will be directed primarily toward ensuring attalnment of full-power operation and collection of meaningful data on core behavior during that operation. The core, 3 feet in diameter and 41/2 feet long, will be composed of bundled ceramic tubes whose central holes will provide continuous air passages from end to end of the reactor. These tubes are to be composed of a homogeneous mixture of UO/sub 2/ fuel and BeO moderator, compacted and sintered to achieve high strength and density. (30 references) (auth)« less

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

Mixed Oxide Fresh Fuel Package Auxiliary Equipment

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

Yapuncich, F.; Ross, A.; Clark, R.H.

2008-07-01

The United States Department of Energy's National Nuclear Security Administration (NNSA) is overseeing the construction the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) on the Savannah River Site. The new facility, being constructed by NNSA's contractor Shaw AREVA MOX Services, will fabricate fuel assemblies utilizing surplus plutonium as feedstock. The fuel will be used in designated commercial nuclear reactors. The MOX Fresh Fuel Package (MFFP), which has recently been licensed by the Nuclear Regulatory Commission (NRC) as a type B package (USA/9295/B(U)F-96), will be utilized to transport the fabricated fuel assemblies from the MFFF to the nuclear reactors. It wasmore » necessary to develop auxiliary equipment that would be able to efficiently handle the high precision fuel assemblies. Also, the physical constraints of the MFFF and the nuclear power plants require that the equipment be capable of loading and unloading the fuel assemblies both vertically and horizontally. The ability to reconfigure the load/unload evolution builds in a large degree of flexibility for the MFFP for the handling of many types of both fuel and non fuel payloads. The design and analysis met various technical specifications including dynamic and static seismic criteria. The fabrication was completed by three major fabrication facilities within the United States. The testing was conducted by Sandia National Laboratories. The unique design specifications and successful testing sequences will be discussed. (authors)« less

Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

NASA Astrophysics Data System (ADS)

Li, Jia; Zhang, Xiaokang; Gao, Fangfang; Pu, Yong

2016-02-01

China Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2 to ensure tritium self-sufficiency. A concept design for a water cooled ceramics breeding blanket (WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR. Based on this concept, a one-dimensional (1D) radial built breeding blanket was first designed, and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build. A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models, addressing neutron wall loading (NWL), tritium breeding ratio (TBR), fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components. The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

REACTOR SERVICE BUILDING, TRA635. CROWDED MOCKUP AREA. CAMERA FACES EAST. ...

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

REACTOR SERVICE BUILDING, TRA-635. CROWDED MOCK-UP AREA. CAMERA FACES EAST. PHOTOGRAPHER'S NOTE SAYS "PICTURE REQUESTED BY IDO IN SUPPORT OF FY '58 BUILDING PROJECTS." INL NEGATIVE NO. 56-3025. R.G. Larsen, Photographer, 9/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The Euratom Seventh Framework Programme FP7 (2007-2011)

NASA Astrophysics Data System (ADS)

Garbil, R.

2010-10-01

The objective of the Seventh Euratom Framework Program in the area of nuclear fission and radiation protection is to establish a sound scientific and technical basis to accelerate practical developments of nuclear energy related to resource efficiency, enhancing safety performance, cost-effectiveness and safer management of long-lived radioactive waste. Key cross-cutting topics such as the nuclear fuel cycle, actinide chemistry, risk analysis, safety assessment, even societal and governance issues are linked to the individual technical areas. Research need to explore new scientific and techno- logical opportunities and to respond in a flexible way to new policy needs that arise. The following activities are to be pursued. (a) Management of radioactive waste, research on partitioning and transmutation and/or other concepts aimed at reducing the amount and/or hazard of the waste for disposal; (b) Reactor systems research to underpin the con- tinued safe operation of all relevant types of existing reactor systems (including fuel cycle facilities), life-time extension, development of new advanced safety assessment methodologies and waste-management aspects of future reactor systems; (c) Radiation protection research in particular on the risks from low doses on medical uses and on the management of accidents; (d) Infrastructures and support given to the availability of, and cooperation between, research infrastructures necessary to maintain high standards of technical achievement, innovation and safety in the European nuclear sector and Research Area. (e) Human resources, mobility and training support to be provided for the retention and further development of scientific competence, human capacity through joint training activities in order to guarantee the availability of suitably qualified researchers, engineers and employees in the nuclear sector over the longer term.

Current Abstracts Nuclear Reactors and Technology

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

Bales, J.D.; Hicks, S.C.

1993-01-01

This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`smore » Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.« less

Nuclear Reactors and Technology

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

Cason, D.L.; Hicks, S.C.

1992-01-01

This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests inmore » NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.« less

Status and problems of fusion reactor development.

PubMed

Schumacher, U

2001-03-01

Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

ENGINEERING TEST REACTOR, TRA642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. ...

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

ENGINEERING TEST REACTOR, TRA-642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. CAMERA IS ON ROOF OF MTR BUILDING AND FACES DUE SOUTH. MTR SERVICE BUILDING, TRA-635, IN LOWER RIGHT CORNER. STEEL FRAMES SHOW BUILDINGS TO BE ATTACHED TO ETR BUILDING. HIGH-BAY SECTION IN CENTER IS REACTOR BUILDING. TWO-STORY CONTROL ROOM AND OFFICE BUILDING, TRA-647, IS BETWEEN IT AND MTR SERVICE BUILDING. STRUCTURE TO THE LEFT (WITH NO FRAMING YET) IS COMPRESSOR BUILDING, TRA-643, AND BEYOND IT WILL BE HEAT EXCHANGER BUILDING, TRA-644, GREAT SOUTHERN BUTTE ON HORIZON. INL NEGATIVE NO. 56-2382. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

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

An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

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

Pattrick Calderoni

2010-09-01

Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactormore » that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R&D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part of the same project [1]. However, this work focuses on two materials: the LiF-BeF2 eutectic (67 and 33 mol%, respectively, also known as flibe) as primary coolant and the LiF-NaF-KF eutectic (46.5, 11.5, and 52 mol%, respectively, also known as flinak) as secondary heat transport fluid. At first common issues are identified, involving the preparation and purification of the materials as well as the development of suitable diagnostics. Than issues specific to each material and its application are considered, with focus on the compatibility with structural materials and the extension of the existing properties database.« less

Batch Tests To Determine Activity Distribution and Kinetic Parameters for Acetate Utilization in Expanded-Bed Anaerobic Reactors

PubMed Central

Fox, Peter; Suidan, Makram T.

1990-01-01

Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distributed within the GAC reactors, whereas in the sand reactor a sludge blanket on top of the sand bed contained approximately 50% of the activity. The Monod half-velocity constant (Ks) for the acetate-utilizing methanogens in two expanded-bed GAC reactors was searched for by combining steady-state results with batch test data. All parameters necessary to develop a model with Monod kinetics were experimentally determined except for Ks. However, Ks was a function of the effluent 3-ethylphenol concentration, and batch test results demonstrated that maximum acetate utilization rates were not a function of the effluent 3-ethylphenol concentration. Addition of a competitive inhibition term into the Monod expression predicted the dependence of Ks on the effluent 3-ethylphenol concentration. A two-parameter search determined a Ks of 8.99 mg of acetate per liter and a Ki of 2.41 mg of 3-ethylphenol per liter. Model predictions were in agreement with experimental observations for all effluent 3-ethylphenol concentrations. Batch tests measured the activity for a specific substrate and determined the distribution of activity in the reactor. The use of steady-state data in conjunction with batch test results reduced the number of unknown kinetic parameters and thereby reduced the uncertainty in the results and the assumptions made. PMID:16348175

Batch tests to determine activity distribution and kinetic parameters for acetate utilization in expanded-bed anaerobic reactors.

PubMed

Fox, P; Suidan, M T

1990-04-01

Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distributed within the GAC reactors, whereas in the sand reactor a sludge blanket on top of the sand bed contained approximately 50% of the activity. The Monod half-velocity constant (K(s)) for the acetate-utilizing methanogens in two expanded-bed GAC reactors was searched for by combining steady-state results with batch test data. All parameters necessary to develop a model with Monod kinetics were experimentally determined except for K(s). However, K(s) was a function of the effluent 3-ethylphenol concentration, and batch test results demonstrated that maximum acetate utilization rates were not a function of the effluent 3-ethylphenol concentration. Addition of a competitive inhibition term into the Monod expression predicted the dependence of K(s) on the effluent 3-ethylphenol concentration. A two-parameter search determined a K(s) of 8.99 mg of acetate per liter and a K(i) of 2.41 mg of 3-ethylphenol per liter. Model predictions were in agreement with experimental observations for all effluent 3-ethylphenol concentrations. Batch tests measured the activity for a specific substrate and determined the distribution of activity in the reactor. The use of steady-state data in conjunction with batch test results reduced the number of unknown kinetic parameters and thereby reduced the uncertainty in the results and the assumptions made.

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…

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.

GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE ...

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

GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE OF REACTOR. INL NEGATIVE NO. 4000. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

NASA Astrophysics Data System (ADS)

Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos

2012-06-01

The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High Level Goals which is Business Goals.

Advanced reactor site feasibility evaluation

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

Haynes, Mark; Owens, Martin; Southworth, Finis

The report provides a description of the project, summarizes the methodology and results for each task of the project, and ends with the overall project conclusions and recommendations. In addition, the report contains a descriptive index of the technical reports generated during the course of the project.

Nuclear Technology Series. Course 7: Reactor Operations.

ERIC Educational Resources Information Center

Center for Occupational Research and Development, Inc., Waco, TX.

This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

Nuclear Technology Series. Course 8: Reactor Safety.

ERIC Educational Resources Information Center

Center for Occupational Research and Development, Inc., Waco, TX.

This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutians in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

Nuclear Technology Series. Course 9: Reactor Auxiliary Systems.

ERIC Educational Resources Information Center

Center for Occupational Research and Development, Inc., Waco, TX.

This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

Nuclear Technology Series. Course 12: Reactor Physics.

ERIC Educational Resources Information Center

Center for Occupational Research and Development, Inc., Waco, TX.

This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

Report Card on Nuclear Power

ERIC Educational Resources Information Center

Novick, Sheldon

1974-01-01

Problems facing the nuclear power industry include skyrocketing construction costs, technical failures, fuel scarcity, power plant safety, and the disposal of nuclear wastes. Possible solutions include: reductions in nuclear power plant construction, a complete moratorium on new plant construction, the construction of fast breeder reactors and the…

HYNOL PROCESS ENGINEERING: PROCESS CONFIGURATION, SITE PLAN, AND EQUIPMENT DESIGN

EPA Science Inventory

The report describes the design of the hydropyrolysis reactor system of the Hynol process. (NOTE: A bench scale methanol production facility is being constructed to demonstrate the technical feasibility of producing methanol from biomass using the Hynol process. The plant is bein...

ETR BUILDING, TRA642, INTERIOR. FIRST FLOOR. REACTOR IS IN CENTER ...

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

ETR BUILDING, TRA-642, INTERIOR. FIRST FLOOR. REACTOR IS IN CENTER OF VIEW. CAMERA FACES NORTHWEST. NOTE CRANE RAILS AND DANGLING ELECTRICAL CABLE AT UPPER PART OF VIEW FOR "MOFFETT 2 TON" CRANE. INL NEGATIVE NO. HD46-14-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

75 FR 11375 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-10

... Spent Fuel Storage/Reactor Decommissioning..... 2.7 0.2 0.2 Test and Research Reactors 0.2 0.0 0.0 Fuel... categories of licenses. The FY 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual...) Spent Fuel Storage/Reactor 122,000 143,000 Decommissioning Test and Research Reactors (Non-power 87,600...

ETR, TRA642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED ...

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

ETR, TRA-642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED WITHIN THE INNER METAL FORM. WHEN CONCRETE IS POURED OUTSIDE THIS FORM, CONDUIT HOLES WILL BE PRESERVE SPACE THROUGH HOLES. INL NEGATIVE NO. 56-1507. Jack L. Anderson, Photographer, 5/8/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR ...

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

REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR IS FUELED AS AN ETR MOCK-UP. LIGHTS DANGLE BELOW WATER LEVEL. CONTROL RODS AND OTHER APPARATUS DESCEND FROM ABOVE WATER LEVEL. INL NEGATIVE NO. 56-900. Jack L. Anderson, Photographer, 3/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTOR SERVICES BUILDING, TRA635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING ...

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

REACTOR SERVICES BUILDING, TRA-635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING AREA AND LABORATORY. CAMERA ON FIRST FLOOR FACING NORTH TOWARD MTR BUILDING. MOCK-UP AREA WAS TO THE RIGHT OF VIEW. INL NEGATIVE NO. HD46-10-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

ERIC Educational Resources Information Center

National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

Operation of the NETL Chemical Looping Reactor with Natural Gas and a Novel Copper-Iron Material

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

Straub, Douglas; Bayham, Samuel; Weber, Justin

The proposed Clean Power Plan requires CO 2 emission reductions of 30% by 2030 and further reductions are targeted by 2050. The current strategies to achieve the 30% reduction targets do not include options for coal. However, the 2016 Annual Energy Outlook suggests that coal will continue to provide more electricity than renewable sources for many regions of the country in 2035. Therefore, cost effective options to reduce greenhouse gas emissions from fossil fuel power plants are vital in order to achieve greenhouse gas reduction targets beyond 2030. As part of the U.S. Department of Energy’s Advanced Combustion Program, themore » National Energy Technology Laboratory’s Research and Innovation Center (NETL R&IC) is investigating the feasibility of a novel combustion concept in which the GHG emissions can be significantly reduced. This concept involves burning fuel and air without mixing these two reactants. If this concept is technically feasible, then CO 2 emissions can be significantly reduced at a much lower cost than more conventional approaches. This indirect combustion concept has been called Chemical Looping Combustion (CLC) because an intermediate material (i.e., a metal-oxide) is continuously cycled to oxidize the fuel. This CLC concept is the focus of this research and will be described in more detail in the following sections. The solid material that is used to transport oxygen is called an oxygen carrier material. The cost, durability, and performance of this material is a key issue for the CLC technology. Researchers at the NETL R&IC have developed an oxygen carrier material that consists of copper, iron, and alumina. This material has been tested extensively using lab scale instruments such as thermogravimetric analysis (TGA), scanning electron microscopy (SEM), mechanical attrition (ASTM D5757), and small fluidized bed reactor tests. This report will describe the results from a realistic, circulating, proof-of-concept test that was completed using NETL’s 50kW th circulating Chemical Looping Reactor (CLR) test facility.« less

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation

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

Gusev, S. I.; Karpov, V. N.; Kiselev, A. N.

2009-09-15

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Test development for the thermionic system evaluation test (TSET) project

NASA Astrophysics Data System (ADS)

Morris, D. Brent; Standley, Vaughn H.; Schuller, Michael J.

1992-01-01

The arrival of a Soviet TOPAZ-II space nuclear reactor affords the US space nuclear power (SNP) community the opportunity to study an assembled thermionic conversion power system. The TOPAZ-II will be studied via the Thermionic System Evaluation Test (TSET) Project. This paper is devoted to the discussion of TSET test development as related to the objectives contained in the TSET Project Plan (Standley et al. 1991). The objectives contained in the Project Plan are the foundation for scheduled TSET tests on TOPAZ-II and are derived from the needs of the Air Force Thermionic SNP program. Our ability to meet the objectives is bounded by unique constraints, such as procurement requirements, operational limitations, and necessary interaction between US and Soviet Scientists and engineers. The fulfillment of the test objectives involves a thorough methodology of test scheduling and data managment. The overall goals for the TSET program are gaining technical understanding of a thermionic SNP system and demonstrating the capabilities and limitations of such a system while assisting in the training of US scientist and engineers in preparation for US SNP system testing. Tests presently scheduled as part of TSET include setup, demonstration, and verification tests; normal and off-normal operating test, and system and component performance tests.

The Use of Case History Studies to Differentiate Potentially Infected from Potentially Noninfected Herds with Reactors to Brucella abortus Antigens

PubMed Central

Martin, S. W.; Gerrow, A. F.

1978-01-01

Data on farm characteristics and the results of the first herd test for brucellosis were collected for 74 reactor and 74 negative herds in Wellington County, Ontario. Each reactor herd was classified as either probably infected or probably not infected using the occurrence of abortions prior to the first herd test or during the testing period, the total number of cattle removed and/or the spread of reactors within the herd as criteria of infection. Statistical techniques were used to select variables which were good “discriminators” between probably infected and noninfected herds. In general, reactor herds were primarily dairy herds and were somewhat larger than negative herds. The presence of only single suspicious reactors on the first test appeared to be a good predictor of lack of infection with Brucella abortus. Among the 37 farms in this category the single reactor was removed from only eight farms and no evidence o fthe spread of infection was observed. The presence of one or more positive reactors on the first herd test appeared to be a good predictor of the presence of infection. Of the 24 farms in this category, evidence of the spread of infection was present in ten farms and seven of these ten farms were eventually depopulated. The brucella milk ring test appeared to be the most effective means of identifying infected herds under the conditions present in Wellington County. PMID:417777

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

DEVICE FOR TREATING MATERIALS

DOEpatents

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17

Test-hole construction in a reactor to facilitate inserting and removing test specimens from the reactor for irradiation therein is discussed. An elongated chamber extends from the outer face of the reactor shield into the reactor. A shield box, having an open end, is sealed to thc outer face of the reactor shield by its open end surrounding the outer end of the chamber. A removable door is provided in the side wall of the shield box for inscrtion and removal of test specimens. A means operable from thc exterior of the shield box is provided for transferring test specimens between the shield box and the irradiation position within the chamber and consists of an elongated rod having a specimen tray engaging member on its inner end, which may be manipulated by the operator.

In-reactor performance of LWR-type tritium target rods

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

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Progress on RERTR activities in Argentina

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

Balart, S.; Calzetta, O.; Cristini, P.

2008-07-15

Since last RERTR meeting, several tasks involving RERTR activities continued deploying in Argentina: through an agreement between CNEA and US-DoE final steps in the RA-6 reactor core conversion from HEU to LEU are taking place; by means of a return campaign of 42 US origin SNF in the frame of the US-SNF FRR program; an effective minimization of HEU inventory is close to be accomplished; development of a LEU dispersed U-Mo fuel prototype, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project is progressing; very high density monolithic U-Mo miniplates andmore » plates using MEU and LEU fuel with Zry-4 cladding were developed to be irradiated as a part of the RERTR program irradiation experiment; atomistic modeling prediction (BFS techniques and first principles) enabled to find some trends on the interaction phases; diffusion couples tests under X-ray synchrotron analysis allowed the characterization of several phases involving U-Mo(-Zr) / Al(-Si); finally CNEA continued spreading high quality LEU technology for fission RI production by means of agreements with different producers interested on HEU-LEU conversion. (author)« less

Development of ENDF/B-IV multigroup neutron cross-section libraries for the LEOPARD and LASER codes. Technical report on Phase 1

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

Jenquin, U.P.; Stewart, K.B.; Heeb, C.M.

1975-07-01

The principal aim of this neutron cross-section research is to provide the utility industry with a 'standard nuclear data base' that will perform satisfactorily when used for analysis of thermal power reactor systems. EPRI is coordinating its activities with those of the Cross Section Evaluation Working Group (CSEWG), responsible for the development of the Evaluated Nuclear Data File-B (ENDF/B) library, in order to improve the performance of the ENDF/B library in thermal reactors and other applications of interest to the utility industry. Battelle-Northwest (BNW) was commissioned to process the ENDF/B Version-4 data files into a group-constant form for use inmore » the LASER and LEOPARD neutronics codes. Performance information on the library should provide the necessary feedback for improving the next version of the library, and a consistent data base is expected to be useful in intercomparing the versions of the LASER and LEOPARD codes presently being used by different utility groups. This report describes the BNW multi-group libraries and the procedures followed in their preparation and testing. (GRA)« less

Production of aluminum-silicon alloy and ferrosilicon and commercial purity aluminum by the direct reduction process. First interim technical report, Phase D, January 1-March 31, 1981

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

Bruno, M.J.

1981-04-01

Operation of the bench AF-reactor on burden with all reducing carbon exterior to the ore pellet resulted in low metal alloy product yields and prematurely terminated runs, indicating the need for intimate contact between alumina and carbon to produce oxycarbide liquid prior to reaction with solid silicon carbide. Carbon solubility tests made on 60Al-40Si alloys at 2200/sup 0/C in graphite crucibles indicated continued reaction to form SiC for one hour. Efficiency of reduction to SiC ranged from 68 to 100%. The A-C two-electrode submerged arc reactor pilot, SAR-II, was successfully operated on both alumina-clay-coke and alumina-silicon carbide-coke (from the VSRmore » prereduction) burdens. Metal alloy was produced and tapped in each of four runs. The pilot crystallizer was operated to evalute the two-stage (stop and go) crystallization technique on obtaining high yields of Al in Al-Si eutectic, with a limit of 1.0% Fe and 0.1% Ti in the alloy product. 18 figures, 19 tables. (DLC)« less

Anaerobic wastewater treatment and membrane filtration: a one night stand or a sustainable relationship?

PubMed

Jeison, D; van Lier, J B

2008-01-01

Several anaerobic membrane bioreactors (AnMBR) were operated, under various conditions, applying different reactor configurations. Applicable fluxes were strongly determined by the physical properties of the sludge present in the reactors. Results show that particle size is a key determining factor for the attainable fluxes. Under thermophilic conditions, small sludge particle size was observed, resulting in low critical fluxes reaching 6-7 L/m2h for the submerged configuration and acidified substrate. In contrast, under mesophilic conditions critical fluxes of 20 L/m2h were obtained. The acidification level also showed a strong effect. Under thermophilic conditions, the presence of a significant fraction of non-acidified organic matter induced the growth of suspended acidogenic biomass that seriously affected the applicable fluxes, both in submerged and side-stream configurations. Under all conditions tested cake formation showed to be the limiting factor determining the applicable fluxes. Only low levels of irreversible fouling were observed. Due to technical and economical considerations, most interesting perspectives for the application of AnMBR are expected with the treatment of high-strength particulate wastewaters, and with extreme wastewaters characterised by high temperature, salinity, etc.

EBR-II and TREAT Digitization Project

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

Griffith, George W.; Rabiti, Cristian

2015-09-01

Digitizing the technical drawings for EBR-II and TREAT provides multiple benefits. Moving the scanned or hard copy drawings to modern 3-D CAD (Computer Aided Drawing) format saves data that could be lost over time. The 3-D drawings produce models that can interface with other drawings to make complex assemblies. The 3-D CAD format can also include detailed material properties and parametric coding that can tie critical dimensions together allowing easier modification. Creating the new files from the old drawings has found multiple inconsistencies that are being flagged or corrected improving understanding of the reactor(s).

Report on the activities of the Danish Atomic Energy Commission up to 31 March 1957

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

None

1958-01-15

Activities of the Danish Atomic Energy Commission from its establishment in 1955 through March, 1957, are reported. The technical and administrative organization of the Commission are outlined. Contracts were signed for the purchase of two reactors. The site for a reactor research establishment was acquired on the Risoe Peninsula near Roskilde. Land for agricultural experiments was acquired nearby. Buildings and facilities were nearing completion by 1957. Training programs for personnel were held. Areas of international cooperation in the peaceful use of atomic energy are outlined. A statement of expenditures is included. (C.H.)

Criticality Safety Evaluation of the LLNL Inherently Safe Subcritical Assembly (ISSA)

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

Percher, Catherine

2012-06-19

The LLNL Nuclear Criticality Safety Division has developed a training center to illustrate criticality safety and reactor physics concepts through hands-on experimental training. The experimental assembly, the Inherently Safe Subcritical Assembly (ISSA), uses surplus highly enriched research reactor fuel configured in a water tank. The training activities will be conducted by LLNL following the requirements of an Integration Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of LLNL instructors. This report provides the technical criticality safety basis for instructional operations with the ISSA experimental assembly.

Our Next Two Steps for Fukushima Daiichi Muon Tomography

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

Miyadera, Haruo

2012-04-11

After the vast disasters caused by the great earthquake and tsunami in eastern Japan, we proposed applying our Muon Tomography (MT) technique to help and improve the emergency situation at Fukushima Daiichi using cosmic-ray muons. A reactor-tomography team was formed at LANL which was supported by the Laboratory as a response to a request by the former Japanese Prime Minister, Naoto Kan. Our goal is to help the Japanese people and support remediation of the reactors. At LANL, we have carried out a proof-of-principle technical demonstration and simulation studies that established the feasibility of MT to image a reactor core.more » This proposal covers the next two critical steps for Fukushima Daiichi Muon Imaging: (1) undertake case study mock-up experiments of Fukushima Daiichi, and (2) system optimization. We requested funding to the US and Japanese government to assess damage of reactors at Fukushima Daiichi. The two steps will bring our project to the 'ready-to-go' level.« 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

A Plasma Reactor for the Synthesis of High-Temperature Materials: Electro Thermal, Processing and Service Life Characteristics

NASA Astrophysics Data System (ADS)

Galevskiy, G. V.; Rudneva, V. V.; Galevskiy, S. G.; Tomas, K. I.; Zubkov, M. S.

2016-08-01

The three-jet direct-flow plasma reactor with a channel diameter of 0.054 m was studied in terms of service life, thermal, technical, and functional capabilities. It was established that the near-optimal combination of thermal efficiency, required specific enthalpy of the plasma-forming gas and its mass flow rate is achieved at a reactor power of 150 kW. The bulk temperature of plasma flow over the rector of 12 gauges long varies within 5500÷3200 K and the wall temperature within 1900÷850 K, when a cylinder from zirconium dioxide of 0.005 m thick is used to thermally insulate the reactor. The specific electric power reaches a high of 1214 MW/m3. The rated service life of electrodes is 4700 hours for a copper anode and 111 hours for a tungsten cathode. The projected contamination of carbides and borides with elec-trode-erosion products doesn't exceed 0.0001% of copper and 0.00002% of tungsten.

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... involving a test and research reactor facility licensed under 10 CFR part 50 and any related inquiry...

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2013 CFR

2013-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... involving a test and research reactor facility licensed under 10 CFR part 50 and any related inquiry...

Nuclear energy acceptance and potential role to meet future energy demand. Which technical/scientific achievements are needed?

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

Schenkel, Roland

25 years after Chernobyl, the Fukushima disaster has changed the perspectives of nuclear power. The disaster has shed a negative light on the independence, reliability and rigor of the national nuclear regulator and plant operator and the usefulness of the international IAEA guidelines on nuclear safety. It has become clear that, in the light of the most severe earthquake in the history of Japan, the plants at Fukushima Daiichi were not adequately protected against tsunamis. Nuclear acceptance has suffered enormously and has changed the perspectives of nuclear energy dramatically in countries that have a very risk-sensitive population, Germany is anmore » example. The paper analyses the reactions in major countries and the expected impact on future deployment of reactors and on R and D activities. On the positive side, the disaster has demonstrated a remarkable robustness of most of the 14 reactors closest to the epicentre of the Tohoku Seaquake although not designed to an event of level 9.0. Public acceptance can only be regained with a rigorous and worldwide approach towards inherent reactor safety and design objectives that limit the impact of severe accidents to the plant itself (like many of the new Gen III reactors). A widespread release of radioactivity and the evacuation (temporary or permanent) of the population up to 30 km around a facility are simply not acceptable. Several countries have announced to request more stringent international standards for reactor safety. The IAEA should take this move forward and intensify and strengthen the different peer review mission schemes. The safety guidelines and peer reviews should in fact become legally binding for IAEA members. The paper gives examples of the new safety features developed over the last 20 years and which yield much safer reactors with lesser burden to the environment under severe accident conditions. The compatibility of these safety systems with the current concepts for fusion-fission hybrids, which have recently been proposed for energy production, is critically reviewed. There are major challenges remaining that are shortly outlined. Scientific/technical achievements that are required in the light of the Fukushima accident are highlighted.« less

Nuclear energy acceptance and potential role to meet future energy demand. Which technical/scientific achievements are needed?

NASA Astrophysics Data System (ADS)

Schenkel, Roland

2012-06-01

25 years after Chernobyl, the Fukushima disaster has changed the perspectives of nuclear power. The disaster has shed a negative light on the independence, reliability and rigor of the national nuclear regulator and plant operator and the usefulness of the international IAEA guidelines on nuclear safety. It has become clear that, in the light of the most severe earthquake in the history of Japan, the plants at Fukushima Daiichi were not adequately protected against tsunamis. Nuclear acceptance has suffered enormously and has changed the perspectives of nuclear energy dramatically in countries that have a very risk-sensitive population, Germany is an example. The paper analyses the reactions in major countries and the expected impact on future deployment of reactors and on R&D activities. On the positive side, the disaster has demonstrated a remarkable robustness of most of the 14 reactors closest to the epicentre of the Tohoku Seaquake although not designed to an event of level 9.0. Public acceptance can only be regained with a rigorous and worldwide approach towards inherent reactor safety and design objectives that limit the impact of severe accidents to the plant itself (like many of the new Gen III reactors). A widespread release of radioactivity and the evacuation (temporary or permanent) of the population up to 30 km around a facility are simply not acceptable. Several countries have announced to request more stringent international standards for reactor safety. The IAEA should take this move forward and intensify and strengthen the different peer review mission schemes. The safety guidelines and peer reviews should in fact become legally binding for IAEA members. The paper gives examples of the new safety features developed over the last 20 years and which yield much safer reactors with lesser burden to the environment under severe accident conditions. The compatibility of these safety systems with the current concepts for fusion-fission hybrids, which have recently been proposed for energy production, is critically reviewed. There are major challenges remaining that are shortly outlined. Scientific/technical achievements that are required in the light of the Fukushima accident are highlighted.

77 FR 1514 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-10

... Technical Specification (TS) 3.4.5, ``RCS Leakage Detection Instrumentation,'' consistent with the NRC... monitoring system are the only operable reactor coolant leakage detection monitoring systems. The modified... Coolant System] Leakage Instrumentation,'' Revision 3. The availability of this TS improvement was...

77 FR 18271 - Terrestrial Environmental Studies for Nuclear Power Stations

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-27

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0182] Terrestrial Environmental Studies for Nuclear Power... Environmental Studies for Nuclear Power Stations.'' This guide provides technical guidance that the NRC staff... nuclear power reactors. ADDRESSES: Please refer to Docket ID NRC-2011-0182 when contacting the NRC about...

Space station program phase B definition: Nuclear reactor-powered space station cost and schedules

NASA Technical Reports Server (NTRS)

1971-01-01

Tabulated data are presented on the costs, schedules, and technical characteristics for the space station phases C and D program. The work breakdown structure, schedule data, program ground rules, program costs, cost-estimating rationale, funding schedules, and supporting data are included.

Nuclear Education and Training Programs of Potential Interest to Utilities.

ERIC Educational Resources Information Center

Atomic Energy Commission, Washington, DC.

This compilation of education and training programs related to nuclear applications in electric power generation covers programs conducted by nuclear reactor vendors, public utilities, universities, technical institutes, and community colleges, which were available in December 1968. Several training-program consultant services are also included.…

Aluminum Carbothermic Technology

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

Bruno, Marshall J.

2005-03-31

This report documents the non-proprietary research and development conducted on the Aluminum Carbothermic Technology (ACT) project from contract inception on July 01, 2000 to termination on December 31, 2004. The objectives of the program were to demonstrate the technical and economic feasibility of a new carbothermic process for producing commercial grade aluminum, designated as the ''Advanced Reactor Process'' (ARP). The scope of the program ranged from fundamental research through small scale laboratory experiments (65 kW power input) to larger scale test modules at up to 1600 kW power input. The tasks included work on four components of the process, Stagesmore » 1 and 2 of the reactor, vapor recovery and metal alloy decarbonization; development of computer models; and economic analyses of capital and operating costs. Justification for developing a new, carbothermic route to aluminum production is defined by the potential benefits in reduced energy, lower costs and more favorable environmental characteristics than the conventional Hall-Heroult process presently used by the industry. The estimated metrics for these advantages include energy rates at approximately 10 kWh/kg Al (versus over 13 kWh/kg Al for Hall-Heroult), capital costs as low as $1250 per MTY (versus 4,000 per MTY for Hall-Heroult), operating cost reductions of over 10%, and up to 37% reduction in CO2 emissions for fossil-fuel power plants. Realization of these benefits would be critical to sustaining the US aluminum industries position as a global leader in primary aluminum production. One very attractive incentive for ARP is its perceived ability to cost effectively produce metal over a range of smelter sizes, not feasible for Hall-Heroult plants which must be large, 240,000 TPY or more, to be economical. Lower capacity stand alone carbothermic smelters could be utilized to supply molten metal at fabrication facilities similar to the mini-mill concept employed by the steel industry. Major accomplishments for the program include definition of the system thermo-chemistry, demonstration of reactor stage 1, development of reactor stage 2 critical components in a 500 kW module, experimental determination of the vapor recovery reactor fundamentals, detailed design and installation of an advanced stage 1/vapor recovery reactor, feasibility of efficient separation of Al-C metal alloy product, updated capital and operating cost estimates, and development of computer models for all steps of the Advanced Reactor Process.« 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

Validation of large-scale, monochromatic UV disinfection systems for drinking water using dyed microspheres.

PubMed

Blatchley, E R; Shen, C; Scheible, O K; Robinson, J P; Ragheb, K; Bergstrom, D E; Rokjer, D

2008-02-01

Dyed microspheres have been developed as a new method for validation of ultraviolet (UV) reactor systems. When properly applied, dyed microspheres allow measurement of the UV dose distribution delivered by a photochemical reactor for a given operating condition. Prior to this research, dyed microspheres had only been applied to a bench-scale UV reactor. The goal of this research was to extend the application of dyed microspheres to large-scale reactors. Dyed microsphere tests were conducted on two prototype large-scale UV reactors at the UV Validation and Research Center of New York (UV Center) in Johnstown, NY. All microsphere tests were conducted under conditions that had been used previously in biodosimetry experiments involving two challenge bacteriophage: MS2 and Qbeta. Numerical simulations based on computational fluid dynamics and irradiance field modeling were also performed for the same set of operating conditions used in the microspheres assays. Microsphere tests on the first reactor illustrated difficulties in sample collection and discrimination of microspheres against ambient particles. Changes in sample collection and work-up were implemented in tests conducted on the second reactor that allowed for improvements in microsphere capture and discrimination against the background. Under these conditions, estimates of the UV dose distribution from the microspheres assay were consistent with numerical simulations and the results of biodosimetry, using both challenge organisms. The combined application of dyed microspheres, biodosimetry, and numerical simulation offers the potential to provide a more in-depth description of reactor performance than any of these methods individually, or in combination. This approach also has the potential to substantially reduce uncertainties in reactor validation, thereby leading to better understanding of reactor performance, improvements in reactor design, and decreases in reactor capital and operating costs.

How does intensification influence the operational and environmental performance of photo-Fenton processes at acidic and circumneutral pH.

PubMed

Salazar, Luis Miguel; Grisales, Claudia Mildred; Garcia, Dorian Prato

2018-05-31

This study evaluates the technical, economical, and environmental impact of sodium persulfate (Na 2 S 2 O 8 ) as an enhancing agent in a photo-Fenton process within a solar-pond type reactor (SPR). Photo-Fenton (PF) and photo-Fenton intensified with the addition of persulfate (PFPS) processes decolorize 97% the azo dye direct blue 71 (DB71) and allow producing a highly biodegradable effluent. Intensification with persulfate allowed reducing treatment time in 33% (from 120 to 80 min) and the consumption of chemical auxiliaries needed for pH adjustment. Energy, reagents, and chemical auxiliaries are still and environmental hotspot for PF and PFPS; however, it is worth mentioning that their environmental footprint is lower than that observed for compound parabolic concentrator (CPC)-type reactors. A life-cycle assessment (LCA) confirms that H 2 O 2 , NaOH, and energy consumption are the variables with the highest impact from an environmental standpoint. The use of persulfate reduced the relative impact in 1.2 to 12% in 12 of the 18 environmental categories studied using the ReCiPe method. The PFPS process emits 1.23 kg CO 2 (CO 2 -Eqv/m 3 treated water). On the other hand, the PF process emits 1.28 kg CO 2 (CO 2 -Eqv/m 3 treated water). Process intensification, chemometric techniques, and the use of SPRs minimize the impact of some barriers (reagent and energy consumption, technical complexity of reactors, pressure drops, dirt on the reflecting surfaces, fragility of reactor materials), limiting the application of advanced oxidation systems at an industrial level, and decrease treatment cost as well as potential environmental impacts associated with energy and reagents consumption. Treatment costs for PF processes (US$0.78/m 3 ) and PFPS processes (US$0.63/m 3 ) were 20 times lower than those reported for photo-Fenton processes in CPC-type reactors.

Combined UHV/high-pressure catalysis setup for depth-resolved near-surface spectroscopic characterization and catalytic testing of model catalysts

NASA Astrophysics Data System (ADS)

Mayr, Lukas; Rameshan, Raffael; Klötzer, Bernhard; Penner, Simon; Rameshan, Christoph

2014-05-01

An ultra-high vacuum (UHV) setup for "real" and "inverse" model catalyst preparation, depth-resolved near-surface spectroscopic characterization, and quantification of catalytic activity and selectivity under technologically relevant conditions is described. Due to the all-quartz reactor attached directly to the UHV-chamber, transfer of the catalyst for in situ testing without intermediate contact to the ambient is possible. The design of the UHV-compatible re-circulating batch reactor setup allows the study of reaction kinetics under close to technically relevant catalytic conditions up to 1273 K without contact to metallic surfaces except those of the catalyst itself. With the attached differentially pumped exchangeable evaporators and the quartz-microbalance thickness monitoring equipment, a reproducible, versatile, and standardised sample preparation is possible. For three-dimensional near-surface sample characterization, the system is equipped with a hemispherical analyser for X-ray photoelectron spectroscopy (XPS), electron-beam or X-ray-excited Auger-electron spectroscopy, and low-energy ion scattering measurements. Due the dedicated geometry of the X-ray gun (54.7°, "magic angle") and the rotatable sample holder, depth analysis by angle-resolved XPS measurements can be performed. Thus, by the combination of characterisation methods with different information depths, a detailed three-dimensional picture of the electronic and geometric structure of the model catalyst can be obtained. To demonstrate the capability of the described system, comparative results for depth-resolved sample characterization and catalytic testing in methanol steam reforming on PdGa and PdZn near-surface intermetallic phases are shown.

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

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

Beatty, Randy L; Harrison, Thomas J

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

A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

NASA Astrophysics Data System (ADS)

Poston, David I.

2005-02-01

The reactor neutron spectrum is one of the fundamental design choices for any fission reactor, but the implications of using a moderated spectrum are vastly different for space reactors as opposed to terrestrial reactors. In addition, the pros and cons of neutron spectra are significantly different among many of the envisioned space power applications. This paper begins with a discussion of the neutronic differences between fast-spectrum and moderated space reactors. This is followed by a discussion of the pros and cons of fast-spectrum and moderated space reactors separated into three areas—technical risk, performance, and safety/safeguards. A mix of quantitative and qualitative arguments is presented, and some conclusions generally can be made regarding neutron spectrum and space power application. In most cases, a fast-spectrum system appears to be the better alternative (mostly because of simplicity and higher potential operating temperatures); however, in some cases, such as a low-power (<100-kWt) surface reactor, a moderated spectrum could provide a better approach. In all cases, the determination of which spectrum is preferred is a strong function of the metrics provided by the "customer"— i.e., if a certain level of performance is required, it could provide a different solution than if a certain level of safeguards is required (which in some cases could produce a null solution). The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

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

Tsai, H.; Gomes, I.C.; Smith, D.L.

1998-09-01

The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2011 CFR

2011-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... this part involving a test and research reactor facility licensed under 10 CFR part 50 and any related...

TREAT Reactor Control and Protection System

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

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

An Expert Elicitation of the Proliferation Resistance of Using Small Modular Reactors (SMR) for the Expansion of Civilian Nuclear Systems.

PubMed

Siegel, Jonas; Gilmore, Elisabeth A; Gallagher, Nancy; Fetter, Steve

2018-02-01

To facilitate the use of nuclear energy globally, small modular reactors (SMRs) may represent a viable alternative or complement to large reactor designs. One potential benefit is that SMRs could allow for more proliferation resistant designs, manufacturing arrangements, and fuel-cycle practices at widespread deployment. However, there is limited work evaluating the proliferation resistance of SMRs, and existing proliferation assessment approaches are not well suited for these novel arrangements. Here, we conduct an expert elicitation of the relative proliferation resistance of scenarios for future nuclear energy deployment driven by Generation III+ light-water reactors, fast reactors, or SMRs. Specifically, we construct the scenarios to investigate relevant technical and institutional features that are postulated to enhance the proliferation resistance of SMRs. The experts do not consistently judge the scenario with SMRs to have greater overall proliferation resistance than scenarios that rely on conventional nuclear energy generation options. Further, the experts disagreed on whether incorporating a long-lifetime sealed core into an SMR design would strengthen or weaken proliferation resistance. However, regardless of the type of reactor, the experts judged that proliferation resistance would be enhanced by improving international safeguards and operating several multinational fuel-cycle facilities rather than supporting many more national facilities. © 2017 Society for Risk Analysis.

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.

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

NONE

Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of researchmore » reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)« less

AGR-2 Irradiation Test Final As-Run Report, Rev 2

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

Collin, Blaise P.

2014-08-01

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samplesmore » for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO 2 fuel, while fast fluence values ranged from 1.94 to 3.47 x 10 25 n/m 2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53 x 10 25 n/m 2 (E >0.18 MeV) for UO 2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO 2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10 -6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2 x 10 -6. In the UO 2 capsule (Capsule 3), the R/B values during the first three cycles were below 10 -7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.« less

Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron Activated Shield Wall

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

Michael R. Kruzic

2007-09-16

Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility was used in the early to mid-1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles in the immediate area. Identified as Corrective Action Unit 115, the TCA facility was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the ''Federal Facilitymore » Agreement and Consent Order''. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously provided technical decisions are made by an experienced decision maker within the site conceptual site model, identified in the Data Quality Objective process. Facility closure involved a seven-step decommissioning strategy. Key lessons learned from the project included: (1) Targeted preliminary investigation activities provided a more solid technical approach, reduced surprises and scope creep, and made the working environment safer for the D&D worker. (2) Early identification of risks and uncertainties provided opportunities for risk management and mitigation planning to address challenges and unanticipated conditions. (3) Team reviews provided an excellent mechanism to consider all aspects of the task, integrated safety into activity performance, increase team unity and ''buy-in'' and promoted innovative and time saving ideas. (4) Development of CED protocols ensured safety and control. (5) The same proven D&D strategy is now being employed on the larger ''sister'' facility, Test Cell C.« less

The current state of the Russian reduced enrichment research reactors program

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

Aden, V.G.; Kartashov, E.F.; Lukichev, V.A.

1997-08-01

During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% frommore » RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.« less

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

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

NASA Technical Reports Server (NTRS)

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

2016-01-01

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

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

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

2016-01-01

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

An approach to model reactor core nodalization for deterministic safety analysis

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

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

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

Nuclear Technology Series. Course 6: Instrumentation and Control of Reactors and Plant Systems.

ERIC Educational Resources Information Center

Center for Occupational Research and Development, Inc., Waco, TX.

This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads

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

NONE

2013-07-01

The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

Silicon Chemical Vapor Deposition Process Using a Half-Inch Silicon Wafer for Minimal Manufacturing System

NASA Astrophysics Data System (ADS)

Li, Ning; Habuka, Hitoshi; Ikeda, Shin-ichi; Hara, Shiro

A chemical vapor deposition reactor for producing thin silicon films was designed and developed for achieving a new electronic device production system, the Minimal Manufacturing, using a half-inch wafer. This system requires a rapid process by a small footprint reactor. This was designed and verified by employing the technical issues, such as (i) vertical gas flow, (ii) thermal operation using a highly concentrated infrared flux, and (iii) reactor cleaning by chlorine trifluoride gas. The combination of (i) and (ii) could achieve a low heating power and a fast cooling designed by the heat balance of the small wafer placed at a position outside of the reflector. The cleaning process could be rapid by (iii). The heating step could be skipped because chlorine trifluoride gas was reactive at any temperature higher than room temperature.

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

Goldsmith, M.W.; Forbes, I.A.; Turnage, J.C.

The potential of new and future energy technologies is discussed, with information provided on availability, technical and economic feasibility, and limitations due to the form of the energy. Energy sources not presently in use (i.e., shale oil, garbage, geothermal, wind, tidal, breeder reactors, ocean thermal gradients, solar energy, and fusion) are expected to supply only 10 to 15% of the Nation's energy requirements in the year 2000. The following chapters are included: Energy Use and Supply; Extending Chemical Fuel Resources, which covers oil shale and tar sands, coal gasification and liquefaction, garbage, and biomass energy; Harnessing the Forces of Nature,more » which describes geothermal, tidal, hydro, wind, and solar energy; New Nuclear Technology (e.g., converter reactors, breeder reactors, fusion by magnetic confinement, and laser fusion); and Improving Energy Production Efficiency, with discussions on energy storage, MHD (magnetohydrodynamics), and combined cycles. (64 references) (BYB)« less

A liquid-metal filling system for pumped primary loop space reactors

NASA Astrophysics Data System (ADS)

Crandall, D. L.; Reed, W. C.

Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

10 CFR 171.3 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... holding an operating license for a power reactor, test reactor or research reactor issued under part 50 of... authorizes operation of a power reactor. The regulations in this part also apply to any person holding a...

Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

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

Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.

2011-11-14

For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may bemore » constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.« less

Advanced Instrumentation for Transient Reactor Testing

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

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

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

Chemical vapor deposition growth

NASA Technical Reports Server (NTRS)

Ruth, R. P.; Manasevit, H. M.; Kenty, J. L.; Moudy, L. A.; Simpson, W. I.; Yang, J. J.

1976-01-01

The chemical vapor deposition (CVD) method for the growth of Si sheet on inexpensive substrate materials is investigated. The objective is to develop CVD techniques for producing large areas of Si sheet on inexpensive substrate materials, with sheet properties suitable for fabricating solar cells meeting the technical goals of the Low Cost Silicon Solar Array Project. Specific areas covered include: (1) modification and test of existing CVD reactor system; (2) identification and/or development of suitable inexpensive substrate materials; (3) experimental investigation of CVD process parameters using various candidate substrate materials; (4) preparation of Si sheet samples for various special studies, including solar cell fabrication; (5) evaluation of the properties of the Si sheet material produced by the CVD process; and (6) fabrication and evaluation of experimental solar cell structures, using standard and near-standard processing techniques.

REACTOR SERVICE BUILDING, TRA635, CONTEXTUAL VIEW DURING CONSTRUCTION. CAMERA IS ...

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

REACTOR SERVICE BUILDING, TRA-635, CONTEXTUAL VIEW DURING CONSTRUCTION. CAMERA IS ATOP MTR BUILDING AND LOOKING SOUTHERLY. FOUNDATION AND DRAINS ARE UNDER CONSTRUCTION. THE BUILDING WILL BUTT AGAINST CHARGING FACE OF PLUG STORAGE BUILDING. HOT CELL BUILDING, TRA-632, IS UNDER CONSTRUCTION AT TOP CENTER OF VIEW. INL NEGATIVE NO. 8518. Unknown Photographer, 8/25/1953 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Reactor transient control in support of PFR/TREAT TUCOP experiments

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

Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

1984-01-01

Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

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

Soubies, B.; Henry, J.Y.; Le Meur, M.

1300 MWe pressurised water reactors (PWRs), like the 1400 MWe reactors, operate with microprocessor-based safety systems. This is particularly the case for the Digital Integrated Protection System (SPIN), which trips the reactor in an emergency and sets in action the safeguard functions. The softwares used in these systems must therefore be highly dependable in the execution of their functions. In the case of SPIN, three players are working at different levels to achieve this goal: the protection system manufacturer, Merlin Gerin; the designer of the nuclear steam supply system, Framatome; the operator of the nuclear power plants, Electricite de Francemore » (EDF), which is also responsible for the safety of its installations. Regulatory licenses are issued by the French safety authority, the Nuclear Installations Safety Directorate (French abbreviation DSIN), subsequent to a successful examination of the technical provisions adopted by the operator. This examination is carried out by the IPSN and the standing group on nuclear reactors. This communication sets out: the methods used by the manufacturer to develop SPIN software for the 1400 MWe PWRs (N4 series); the approach adopted by the IPSN to evaluate the safety software of the protection system for the N4 series of reactors.« less

Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

NASA Astrophysics Data System (ADS)

Passerini, Stefano

For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test the robustness of the conclusions presented in the MIT Fuel Cycle Study. These conclusions are found to still hold, even when considering alternative technologies and different sets of simulation assumptions. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. Optimization metrics of interest for different stakeholders in the fuel cycle (economics, fuel resource utilization, high level waste, transuranics/proliferation management, and environmental impact) are utilized for two different optimization techniques: a linear one and a stochastic one. Stakeholder elicitation provided sets of relative weights for the identified metrics appropriate to each stakeholder group, which were then successfully used to arrive at optimum fuel cycle configurations for recycling technologies. The stochastic optimization tool, based on a genetic algorithm, was used to identify non-inferior solutions according to Pareto's dominance approach to optimization. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

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.

HOT CELL BUILDING, TRA632. CONTEXTUAL VIEW ALONG WALLEYE AVENUE, CAMERA ...

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

HOT CELL BUILDING, TRA-632. CONTEXTUAL VIEW ALONG WALLEYE AVENUE, CAMERA FACING EASTERLY. HOT CELL BUILDING IS AT CENTER LEFT OF VIEW; THE LOW-BAY PROJECTION WITH LADDER IS THE TEST TRAIN ASSEMBLY FACILITY, ADDED IN 1968. MTR BUILDING IS IN LEFT OF VIEW. HIGH-BAY BUILDING AT RIGHT IS THE ENGINEERING TEST REACTOR BUILDING, TRA-642. INL NEGATIVE NO. HD46-32-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Affordable Development and Qualification Strategy for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Gerrish, Harold P., Jr.; Doughty, Glen E.; Bhattacharyya, Samit K.

2013-01-01

Nuclear Thermal Propulsion (NTP) is a concept which uses a nuclear reactor to heat a propellant to high temperatures without combustion and can achieve significantly greater specific impulse than chemical engines. NTP has been considered many times for human and cargo missions beyond low earth orbit. A lot of development and technical maturation of NTP components took place during the Rover/NERVA program of the 60's and early 70's. Other NTP programs and studies followed attempting to further mature the NTP concept and identify a champion customer willing to devote the funds and support the development schedule to a demonstration mission. Budgetary constraints require the use of an affordable development and qualification strategy that takes into account all the previous work performed on NTP to construct an existing database, and include lessons learned and past guidelines followed. Current guidelines and standards NASA uses for human rating chemical rocket engines is referenced. The long lead items for NTP development involve the fuel elements of the reactor and ground testing the engine system, subsystem, and components. Other considerations which greatly impact the development plans includes the National Space Policy, National Environmental Policy Act, Presidential Directive/National Security Council Memorandum #25 (Scientific or Technological Experiments with Possible Large-Scale Adverse Environmental Effects and Launch of Nuclear Systems into Space), and Safeguards and Security. Ground testing will utilize non-nuclear test capabilities to help down select components and subsystems before testing in a nuclear environment to save time and cost. Existing test facilities with minor modifications will be considered to the maximum extent practical. New facilities will be designed to meet minimum requirements. Engine and test facility requirements are based on the driving mission requirements with added factors of safety for better assurance and reliability. Emphasis will be placed on small engines, since the smaller the NTP engine, the easier it is to transport, assemble/disassemble, and filter the exhaust during tests. A new ground test concept using underground bore holes (modeled after the underground nuclear test program) to filter the NTP engine exhaust is being considered. The NTP engine system design, development, test, and evaluation plan includes many engine components and subsystems, which are very similar to those used in chemical engines, and can be developed in conjunction with them Other less mature NTP engine components and subsystems (e.g., reactor) will be thoroughly analyzed and tested to acceptable levels recommended by the referenced standards and guidelines. The affordable development strategy also considers a prototype flight test, as a final step in the development process. Preliminary development schedule estimates show that an aggressive development schedule (without much margin) will be required to be flight ready for a 2033 human mission to Mars.

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

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

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara

2005-02-06

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an earlymore » prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.« less

Environmental assessment of SP-100 ground engineering system test site: Hanford Site, Richland, Washington

NASA Astrophysics Data System (ADS)

1988-12-01

The US Department of Energy (DOE) proposes to modify an existing reactor containment building (decommissioned Plutonium Recycle Test Reactor (PRTR) 309 Building) to provide ground test capability for the prototype SP-100 reactor. The 309 Building (Figure 1.1) is located in the 300 Area on the Hanford Site in Washington State. The National Environmental Policy Act (NEPA) requires that Federal agencies assess the potential impacts that their actions may have on the environment. This Environmental Assessment describes the consideration given to environmental impacts during reactor concept and test site selection, examines the environmental effects of the DOE proposal to ground test the nuclear subsystem, describes alternatives to the proposed action, and examines radiological risks of potential SP-100 use in space.

An evaluation of alloys and coatings for use in automobile thermal reactors

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Oldrieve, R. E.

1974-01-01

Several candidate alloys and coatings were evaluated for use in automobile thermal reactors. Full-size reactors of the candidate materials were analyzed in cyclic engine dynamometer tests with peak temperature of 1900 F (1040 C). Two developmental ferritic iron alloys GE1541 and NASA-18T - exhibited the best overall performance lasting at least 60% of the life of the test engine. Four of the alloys evaluated warrant consideration for reactor use. They include GE1541, Armco 18 SR, NASA-18T, and Inconel 601. None of the commercial coating substrate combinations evaluated warrant consideration for reactor use.-

Evaluation of alloys and coatings for use in automobile thermal reactors

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Oldrieve, R. E.

1974-01-01

Several candidate alloys and coatings were evaluated for use in automobile thermal reactors. Full-size reactors of the candidate materials were evaluated in cyclic engine dynamometer tests with a peak temperature of 1040 C (1900 F). Two developmental ferritic-iron alloys, GE-1541 and NASA-18T, exhibited the best overall performance by lasting at least 60 percent of the life of test engine. Four of the alloys evaluated warrant consideration for reactor use. They are GE-1541, Armco 18 SR, NASA-18T, and Inconel 601. None of the commercial coating substrate combinations evaluated warrant consideration for reactor use.

Establishment of the roadmap for chlorination process development for zirconium recovery and recycle

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

Collins, E.D.; Del Cul, G.D.; Spencer, B.B.

Process development studies are being conducted to recover, purify, and reuse the zirconium (about 98.5% by mass) in used nuclear fuel (UNF) zirconium alloy cladding. Feasibility studies began in FY 2010 using empty cladding hulls that were left after fuel dissolution or after oxidation to a finely divided oxide powder (voloxidation). In FY 2012, two industrial teams (AREVA and Shaw-Westinghouse) were contracted by the Department of Energy Office of Nuclear Energy (NE) to provide technical assistance to the project. In FY 2013, the NE Fuel Cycle Research and Development Program requested development of a technology development roadmap to guide futuremore » work. The first step in the roadmap development was to assess the starting point, that is, the current state of the technology and the end goal. Based on previous test results, future work is to be focused on first using chlorine as the chlorinating agent and secondly on the use of a process design that utilizes a chlorination reactor and dual ZrCl{sub 4} product salt condensers. The likely need for a secondary purification step was recognized. Completion of feasibility testing required an experiment on the chemical decladding flowsheet option. This was done during April 2013. The roadmap for process development will continue through process chemistry optimization studies, the chlorinated reactor design configuration, product salt condensers, and the off-gas trapping of tritium or other volatile fission products from the off-gas stream. (authors)« less

The effect of catalyst length and downstream reactor distance on catalytic combustor performance

NASA Technical Reports Server (NTRS)

Anderson, D.

1980-01-01

A study was made to determine the effects on catalytic combustor performance which resulted from independently varying the length of a catalytic reactor and the length available for gas-phase reactions downstream of the catalyst. Monolithic combustion catalysts from three manufacturers were tested in a combustion test rig with no. 2 diesel fuel. Catalytic reactor lengths of 2.5 and 5.4 cm, and downstream gas-phase reaction distances of 7.3, 12.4, 17.5, and 22.5 cm were evaluated. Measurements of carbon monoxide, unburned hydrocarbons, nitrogen oxides, and pressure drop were made. The catalytic-reactor pressure drop was less than 1 percent of the upstream total pressure for all test configurations and test conditions. Nitrogen oxides and unburned hydrocarbons emissions were less than 0.25 g NO2/kg fuel and 0.6 g HC/kg fuel, respectively. The minimum operating temperature (defined as the adiabatic combustion temperature required to obtain carbon monoxide emissions below a reference level of 13.6 g CO/kg fuel) ranged from 1230 K to 1500 K for the various conditions and configurations tested. The minimum operating temperature decreased with increasing total (catalytic-reactor-plus-downstream-gas-phase-reactor-zone) residence time but was independent of the relative times spent in each region when the catalytic-reactor residence time was greater than or equal to 1.4 ms.

EAST FACE OF REACTOR BASE. COMING TOWARD CAMERA IS EXCAVATION ...

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

EAST FACE OF REACTOR BASE. COMING TOWARD CAMERA IS EXCAVATION FOR MTR CANAL. CAISSONS FLANK EACH SIDE. COUNTERFORT (SUPPORT PERPENDICULAR TO WHAT WILL BE THE LONG WALL OF THE CANAL) RESTS ATOP LEFT CAISSON. IN LOWER PART OF VIEW, DRILLERS PREPARE TRENCHES FOR SUPPORT BEAMS THAT WILL LIE BENEATH CANAL FLOOR. INL NEGATIVE NO. 739. Unknown Photographer, 10/6/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. ON GROUND FLOOR. THE 60TON ETR REACTOR VESSEL ...

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

ETR, TRA-642. ON GROUND FLOOR. THE 60-TON ETR REACTOR VESSEL IS DROPPED INTO PLACE OVER PIT. KAISER USED TWO MULTI-BLOCK DRUM PULLEYS WITH A COMBINED CAPACITY OF 100 TONS AND A 100-TON DRUM HOIST. THE VESSEL WAS 35 1/2 FEET LONG. INL NEGATIVE NO. 56-4149. R.G. Larsen, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

LOFT. Reactor arrives at containment building (TAN650), now being pushed ...

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

LOFT. Reactor arrives at containment building (TAN-650), now being pushed by locomotive. Camera facing northerly. Note "Hello Dolly" and "PWR MTA No. 1" (pressurized water reactor mobile test assembly) signs. Date: 1973. INEEL negative no. 73-3710 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE ...

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

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE OF ETR REACTOR, CAMERA FACING NORTH. CABINET CONTAINING "NUCLEAR INSTRUMENT SYSTEMS" IS RESTRICTED. INL NEGATIVE NO. HD46-18-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

FAST CHOPPER BUILDING, TRA665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. ...

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

FAST CHOPPER BUILDING, TRA-665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. MTR REACTOR SERVICES BUILDING,TRA-635, TO LEFT; MTR BUILDING TO RIGHT. CAMERA FACING WEST. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

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

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Nonthermal plasma technology for organic destruction

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

Heath, W.O.; Birmingham, J.G.

1995-06-01

Pacific Northwest Laboratory (PNL) is investigating the use of nonthermal, electrically driven plasmas for destroying organic contaminants near ambient temperatures and pressures. Three different plasma systems have been developed to treat organics in air, water, and soil. These systems are the Gas-Phase Corona Reactor (GPCR)III for treating air, the Liquid-Phase Corona Reactor for treating water, and In Situ Corona for treating soils. This presentation focuses on recent technical developments, commercial status, and project costs of OPCR as a cost-effective alternative to other air-purification technologies that are now in use to treat off-gases from site-remediation efforts as well as industrial emissions.

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

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee

The purpose of this collaborative Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, plant engineering, material compatibility studies, and coolant activation. The publications derived from workmore » on this project (since project inception) are listed in Appendix A.« less

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.

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

PubMed

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Development of LWR Fuels with Enhanced Accident Tolerance

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

Lahoda, Edward J.; Boylan, Frank A.

2015-10-30

Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15N and U 3Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3Si 2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pelletsmore » for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3Si 2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2, UN, and U 3Si 2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2. Cold spray application of either the amorphous steel or the Ti 2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.« less

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

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

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

2011-09-01

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

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

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

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

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

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely themore » total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.« less

Removal of anaerobic soluble microbial products in a biological activated carbon reactor.

PubMed

Dong, Xiaojing; Zhou, Weili; He, Shengbing

2013-09-01

The soluble microbial products (SMP) in the biological treatment effluent are generally of great amount and are poorly biodegradable. Focusing on the biodegradation of anaerobic SMP, the biological activated carbon (BAC) was introduced into the anaerobic system. The experiments were conducted in two identical lab-scale up-flow anaerobic sludge blanket (UASB) reactors. The high strength organics were degraded in the first UASB reactor (UASB1) and the second UASB (UASB2, i.e., BAC) functioned as a polishing step to remove SMP produced in UASB1. The results showed that 90% of the SMP could be removed before granular activated carbon was saturated. After the saturation, the SMP removal decreased to 60% on the average. Analysis of granular activated carbon adsorption revealed that the main role of SMP removal in BAC reactor was biodegradation. A strain of SMP-degrading bacteria, which was found highly similar to Klebsiella sp., was isolated, enriched and inoculated back to the BAC reactor. When the influent chemical oxygen demand (COD) was 10,000 mg/L and the organic loading rate achieved 10 kg COD/(m3 x day), the effluent from the BAC reactor could meet the discharge standard without further treatment. Anaerobic BAC reactor inoculated with the isolated Klebsiella was proved to be an effective, cheap and easy technical treatment approach for the removal of SMP in the treatment of easily-degradable wastewater with COD lower than 10,000 mg/L.

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2014 CFR

2014-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2012 CFR

2012-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2010 CFR

2010-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2013 CFR

2013-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility

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

Doerner, R.C.; Bauer, T.H.; Morman, J.A.

Prototypic oxide fuel was subjected to simulated, fast reactor severe accident conditions in a series of in-pile tests in the Transient Reactor Test Facility reactor. Seven experiments were performed on fresh and previously irradiated oxide fuel pins under transient overpower and transient undercooled. overpower accident conditions. For each of the tests, fuel motions were observed by the hodoscope. Hodoscope data are correlated with coolant flow, pressure, and temperature data recorded by the loop instrumentation. Data were analyzed from the onset of initial failure to a final mass distribution at the end of the test. In this paper results of thesemore » analyses are compared to pre- and posttest accident calculations and to posttest metallographic accident calculations and to posttest metallographic examinations and computed tomographic reconstructions from neutron radiographs.« less

EARLY ENTRANCE COPRODUCTION PLANT

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

David Storm; Govanon Nongbri; Steve Decanio

2004-01-12

The overall objective of this project is the three phase development of an Early Entrance Coproduction Plant (EECP) which uses petroleum coke to produce at least one product from at least two of the following three categories: (1) electric power (or heat), (2) fuels, and (3) chemicals using ChevronTexaco's proprietary gasification technology. The objective of Phase I is to determine the feasibility and define the concept for the EECP located at a specific site; develop a Research, Development, and Testing (RD&T) Plan to mitigate technical risks and barriers; and prepare a Preliminary Project Financing Plan. The objective of Phase IImore » is to implement the work as outlined in the Phase I RD&T Plan to enhance the development and commercial acceptance of coproduction technology. The objective of Phase III is to develop an engineering design package and a financing and testing plan for an EECP located at a specific site. The project's intended result is to provide the necessary technical, economic, and environmental information needed by industry to move the EECP forward to detailed design, construction, and operation. The partners in this project are Texaco Energy Systems LLC or TES (a subsidiary of ChevronTexaco), General Electric (GE), Praxair, and Kellogg Brown & Root (KBR) in addition to the U.S. Department of Energy (DOE). TES is providing gasification technology and Fischer-Tropsch (F-T) technology developed by Rentech, Inc., GE is providing combustion turbine technology, Praxair is providing air separation technology, and KBR is providing engineering. During Phase I, a design basis for the Fischer-Tropsch Synthesis section was developed based on limited experience with the specified feed gas and operating conditions. The objective of this Task in Phase II RD&T work was to confirm the performance of the F-T reactor at the set design conditions. Although much of the research, development, and testing work were done by TES outside of this project, several important issues were addressed in this phase of the project. They included Rejuvenation/Regeneration of the Fischer-Tropsch Catalyst, online Catalyst Withdrawal and Addition from the synthesis reactor, and the Fischer-Tropsch Design Basis Confirmation. In Phase III the results from these RD&T work will be incorporated in developing the engineering design package. This Topical Report documents the Phase II RD&T work that was completed for this task.« less

76 FR 64399 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-18

... Thursday, October 21, 2010 (74 FR 65038-65039). Thursday, November 3, 2011, Conference Room T2-B1, 11545... Evaluation of Diversity and Defense-In-Depth in Digital Computer-Based Instrumentation and Control Systems... staff regarding Branch Technical Position 7- 19, ``Guidance for the Evaluation of Diversity and Defense...

Dry coal feeder development program at Ingersoll-Rand Research, Incorporated. [for coal gasification systems

NASA Technical Reports Server (NTRS)

Mistry, D. K.; Chen, T. N.

1977-01-01

A dry coal screw feeder for feeding coal into coal gasification reactors operating at pressures up to 1500 psig is described. Results on the feeder under several different modes of operation are presented. In addition, three piston feeder concepts and their technical and economical merits are discussed.

10 CFR 52.157 - Contents of applications; technical information in final safety analysis report.

Code of Federal Regulations, 2010 CFR

2010-01-01

... analysis of the structures, systems, and components of the reactor to be manufactured, with emphasis upon... assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site... structures, systems, and components with the objective of assessing the risk to public health and safety...

REPORT OF THE UNITED STATES DELEGATION TO THE INTERNATIONAL CONFERENCE ON THE PEACEFUL USES OF ATOMIC ENERGY HELD BY THE UNITED NATIONS, AUGUST 8-20, 1955, GENEVA, SWITZERLAND WITH APPENDICES AND SELECTED DOCUMENTS. VOLUMES I AND II

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

Page, N.

The background for the conference is given. Plenary sessions papers are abstracted but discussions are quoted. Lists are included of technical sessions, U.S. papers, organizations, delegates, exhibits, with some pictures included, contents of the technical library and films available. Press releases are reported. Also included are two U.S. brochures prepared for the U.S. Exhibit, ''United States Research Reactor'' and ''Techical Exhibition of the United States of America.''

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

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

Wayne Moe

2012-08-01

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

Technical and economic assessment of different options for minor actinide transmutation: the French case

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

Chabert, C.; Coquelet-Pascal, C.; Saturnin, A.

Studies have been performed to assess the industrial perspectives of partitioning and transmutation of long-lived elements. These studies were carried out in tight connection with GEN-IV systems development. The results include the technical and economic evaluation of fuel cycle scenarios along with different options for optimizing the processes between the minor actinide transmutation in fast neutron reactors, their interim storage and geological disposal of ultimate waste. The results are analysed through several criteria (impacts on waste, on waste repository, on fuel cycle plants, on radiological exposure of workers, on costs and on industrial risks). These scenario evaluations take place inmore » the French context which considers the deployment of the first Sodium-cooled Fast Reactor (SFR) in 2040. 3 management options of minor actinides have been studied: no transmutation, transmutation in SFR and transmutation in an accelerator-driven system (ADS). Concerning economics the study shows that the cost overrun related to the transmutation process could vary between 5 to 9% in SFR and 26 % in the case of ADS.« less

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

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

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

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

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

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

A review of carbide fuel corrosion for nuclear thermal propulsion applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1993-10-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1994-07-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A nuclear driven metallic vapor MHD coupled with MPD thrusters

NASA Technical Reports Server (NTRS)

Anghaie, Samim; Kumar, Ratan

1991-01-01

Nuclear energy as a source of power for space missions, represents an enabling technology for advanced and ambitious space applications. Nuclear fuel in a gaseous or liquid form has been configured as a promising and practical candidate in this regard. The present study investigates and performs a feasibility analysis of an innovative concept for space power generation and propulsion. The system embodies a conceptual nuclear reactor with an MHD generator and coupled to MPD thrusters. The reactor utilizes liquid uranium in droplet form as fuel and superheated metallic vapor as the working fluid. This ultrahigh temperature vapor core reactor brings forward varied and challenging technical issues, and it has been addressed to in this paper. A parametric study of the conceived system has been performed in a qualitative and quantitative manner. Preliminary results show enough promise for further indepth analysis of this novel system.

Molybdenum-base cermet fuel development

NASA Astrophysics Data System (ADS)

Pilger, James P.; Gurwell, William E.; Moss, Ronald W.; White, George D.; Seifert, David A.

Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. The MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermte. This cermet is to have a high matrix density (greater than or equal to 95 percent) for high strength and high thermal conductance coupled with a high particle (UN) porosity (approximately 25 percent) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous Un microspheres become available. Process development was conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and the vacuum hot press consolidation techniques. This paper summarizes the status of these activities.

Computed tomography of radioactive objects and materials

NASA Astrophysics Data System (ADS)

Sawicka, B. D.; Murphy, R. V.; Tosello, G.; Reynolds, P. W.; Romaniszyn, T.

1990-12-01

Computed tomography (CT) has been performed on a number of radioactive objects and materials. Several unique technical problems are associated with CT of radioactive specimens. These include general safety considerations, techniques to reduce background-radiation effects on CT images and selection criteria for the CT source to permit object penetration and to reveal accurate values of material density. In the present paper, three groups of experiments will be described, for objects with low, medium and high levels of radioactivity. CT studies on radioactive specimens will be presented. They include the following: (1) examination of individual ceramic reactor-fuel (uranium dioxide) pellets, (2) examination of fuel samples from the Three Mile Island reactor, (3) examination of a CANDU (CANada Deuterium Uraniun: registered trademark) nuclear-fuel bundle which underwent a simulated loss-of-coolant accident resulting in high-temperature damage and (4) examination of a PWR nuclear-reactor fuel assembly.

MTR, TRA603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTYMETER CHOPPER ...

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

MTR, TRA-603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTY-METER CHOPPER HOUSE. COFFIN TURNING ROLLS. REMOVABLE PANEL OVER CANAL ON EAST SIDE. NEW PLUG STORAGE ACCESS. DOOR SCHEDULE INDICATES STEEL (FOR VAULT), WIRE MESH, AND HOLLOW METAL TYPES. STORAGE AND ISSUE ROOM. SAFETY SHOWERS. DOORWAY TO WING, TRA-604. BLAW-KNOX 3150-803-2, 7/1950. INL INDEX NO. 531-0603-00-098-100561, REV. 10. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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.

REACTOR SERVICE BUILDING, TRA635, INTERIOR. CAMERA FACES NORTHWEST TOWARDS INTERIOR ...

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

REACTOR SERVICE BUILDING, TRA-635, INTERIOR. CAMERA FACES NORTHWEST TOWARDS INTERIOR WALL ENCLOSING STORAGE AND OFFICE SPACE ALONG THE WEST SIDE. AT RIGHT EDGE IS DOOR TO MTR BUILDING. FROM RIGHT TO LEFT, SPACE WAS PLANNED FOR A LOCKER ROOM, MTR ISSUE ROOM, AND STORAGE AREAS AND RELATED OFFICES. NOTE SECOND "MEZZANINE" FLOOR ABOVE. INL NEGATIVE NO. 10227. Unknown Photographer, 3/23/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NASA-EPA automotive thermal reactor technology program

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

Preliminary Process Design of ITER ELM Coil Bracket Brazing

NASA Astrophysics Data System (ADS)

LI, Xiangbin; SHI, Yi

2015-03-01

With the technical requirement of the International Thermonuclear Experimental Reactor (ITER) project, the manufacture and assembly technology of the mid Edge Localized Modes (ELM) coil was developed by the Institute of Plasma Physics, Chinese Academy of Science (ASIPP). As the gap between the bracket and the Stainless Steel jacketed and Mineral Insulated Conductor (SSMIC) can be larger than 0.5 mm instead of 0.01 mm to 0.1 mm as in normal industrial cases, the process of mid ELM coil bracket brazing to the SSMICT becomes quiet challenging, from a technical viewpoint. This paper described the preliminary design of ELM coil bracket brazing to the SSMIC process, the optimal bracket brazing curve and the thermal simulation of the bracket furnace brazing method developed by ANSYS. BAg-6 foil (Bag50Cu34Zn16) plus BAg-1a paste (Bag45CuZnCd) solders were chosen as the brazing filler. By testing an SSMICT prototype, it is shown that the average gap between the bracket and the SSMIC could be controlled to 0.2-0.3 mm, and that there were few voids in the brazing surface. The results also verified that the preliminary design had a favorable heat conducting performance in the bracket.

75 FR 34219 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-16

....8 $6.3 $7.5 Spent Fuel Storage/Reactor Decommissioning..... -- -- 2.7 0.2 0.2 Test and Research... 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual Fees FY2009 Annual FY 2010... Decommissioning Test and Research Reactors (Non-power 87,600 81,700 Reactors) High Enriched Uranium Fuel Facility...

ETR, TRA642. ON GROUND FLOOR. WITH OUTER THERMAL RING IN ...

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

ETR, TRA-642. ON GROUND FLOOR. WITH OUTER THERMAL RING IN PLACE AND CONDUIT PRESERVED, HIGH-DENSITY CONCRETE IS PLACED BETWEEN THE THERMAL RING AND THE OUTER REACTOR FORM. INL NEGATIVE NO. 56-2400. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Reliability and safety of the electrical power supply complex of the Hanford production reactors

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

Robbins, F.D.

Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high ordermore » of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to need adjustment but most of all on those changes which will improve reactor safety. This report is the result of such a study.« less

Reference reactor module for NASA's lunar surface fission power system

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

Poston, David I; Kapernick, Richard J; Dixon, David D

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on themore » lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.« less

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

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

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

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

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

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

A K Wertsching

2012-09-01

As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Fluxmore » Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.« less

HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E

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

Susan Stacy; Hollie K. Gilbert

2005-02-01

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to housemore » the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for commercial power reactors. Other projects included NASA's Systems for Nuclear Auxiliary Power and storage of Three Mile Island meltdown debris. National missions for TAN in reactor research and safety research have expired; demolition of historic TAN buildings is underway.« less

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

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

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

2004-02-04

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

Effects of glucose on the performance of enhanced biological phosphorus removal activated sludge enriched with acetate.

PubMed

Gebremariam, Seyoum Yami; Beutel, Marc W; Christian, David; Hess, Thomas F

2012-10-01

The effects of glucose on enhanced biological phosphorus removal (EBPR) activated sludge enriched with acetate was investigated using sequencing batch reactors. A glucose/acetate mixture was serially added to the test reactor in ratios of 25/75%, 50/50%, and 75/25% and the EBPR activity was compared to the control reactor fed with 100% acetate. P removal increased at a statistically significant level to a near-complete in the test reactor when the mixture increased to 50/50%. However, EBPR deteriorated when the glucose/acetate mixture increased to 75/25% in the test reactor and when the control reactor abruptly switched to 100% glucose. These results, in contrast to the EBPR conventional wisdom, suggest that the addition of glucose at moderate levels in wastewaters does not impede and may enhance EBPR, and that glucose waste products should be explored as an economical sustainable alternative when COD enhancement of EBPR is needed. Copyright © 2012 Elsevier Ltd. All rights reserved.

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.

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

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

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas releasemore » and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.« less

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Innovations for In-Pile Measurements in the Framework of the CEA-SCK•CEN Joint Instrumentation Laboratory

NASA Astrophysics Data System (ADS)

Villard, Jean-Francois; Schyns, Marc

2010-12-01

Optimizing the life cycle of nuclear systems under safety constraints requires high-performance experimental programs to reduce uncertainties on margins and limits. In addition to improvement in modeling and simulation, innovation in instrumentation is crucial for analytical and integral experiments conducted in research reactors. The quality of nuclear research programs relies obviously on an excellent knowledge of their experimental environment which constantly calls for better online determination of neutron and gamma flux. But the combination of continuously increasing scientific requirements and new experimental domains -brought for example by Generation IV programsnecessitates also major innovations for in-pile measurements of temperature, dimensions, pressure or chemical analysis in innovative mediums. At the same time, the recent arising of a European platform around the building of the Jules Horowitz Reactor offers new opportunities for research institutes and organizations to pool their resources in order to face these technical challenges. In this situation, CEA (French Nuclear Energy Commission) and SCK'CEN (Belgian Nuclear Research Centre) have combined their efforts and now share common developments through a Joint Instrumentation Laboratory. Significant progresses have thus been obtained recently in the field of in-pile measurements, on one hand by improvement of existing measurement methods, and on the other hand by introduction in research reactors of original measurement techniques. This paper highlights the state-of-the-art and the main requirements regarding in-pile measurements, particularly for the needs of current and future irradiation programs performed in material testing reactors. Some of the main on-going developments performed in the framework of the Joint Instrumentation Laboratory are also described, such as: - a unique fast neutron flux measurement system using fission chambers with 242Pu deposit and a specific online data processing, - an optical system designed to perform in-pile dimensional measurements of material samples under irradiation, - an acoustical instrumentation allowing the online characterization of fission gas release in Pressurized Water Reactor fuel rods. For each example, the obtained results, expected impacts and development status are detailed.

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

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

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

Commercial-scale demonstration of the Liquid Phase Methanol (LPMEOH{trademark}) process. Technical progress report number 11, January 1--March 31, 1997

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

NONE

1997-06-11

During this quarter, the third draft of the Topical Report on Process Economics Studies was issued for review. A recommendation to continue with design verification testing on the coproduction of methanol and dimethyl ether (DME) was made. A liquid phase dimethyl ether (LPDME) catalyst system with reasonable long-term activity and stability is being developed, and a decision to proceed with a proof-of-concept test run at the LaPorte Alternative Fuels Development Unit (AFDU) is pending the release of a memo from Air Products on the catalyst targets and corresponding economics for a commercially successful LPDME catalyst. The off-site product-use test planmore » is to be updated in June of 1997. During this quarter, Air Products and Acurex Environmental Corporation continued developing the listing of product-use test participants who are involved in fuel cell, transportation, and stationary power plant applications. Start-up activities (Task 3.1) began during the reporting period, and coal-derived synthesis gas (syngas) was introduced to the demonstration unit. The recycle compressor was tested successfully on syngas at line pressure of 700 psig, and the reactor loop reached 220 C for carbonyl burnout. Iron carbonyl in the balanced gas feed remained below the 10 ppbv detection limit for all samples but one. Within the reactor loop, iron carbonyl levels peaked out near 200 ppbv after about 40 hours on-stream, before decreasing to between 10--20 ppbv at 160 hours on -stream. Nickel carbonyl measurements reached a peak of about 60 ppbv, and decreased at all sampling locations to below the 10 ppbv detection limit by 70 hours on-stream. Catalyst activation of the nine 2,250 lb batches required for the initial catalyst charge began and concluded. All batches met or slightly exceeded the theoretical maximum uptake of 2.82 SCF of reducing gas/lb catalyst.« less

46 CFR 160.066-13 - Technical tests.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 46 Shipping 6 2010-10-01 2010-10-01 false Technical tests. 160.066-13 Section 160.066-13 Shipping....066-13 Technical tests. (a) The following conditions apply to technical tests as described in this... signals are protected by sealed packaging, then the conditioning for the technical tests must be conducted...

46 CFR 160.066-13 - Technical tests.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 46 Shipping 6 2014-10-01 2014-10-01 false Technical tests. 160.066-13 Section 160.066-13 Shipping....066-13 Technical tests. (a) The following conditions apply to technical tests as described in this... signals are protected by sealed packaging, then the conditioning for the technical tests must be conducted...

REVIEW OF PROPOSED METHODOLOGY FOR A RISK- INFORMED RELAXATION TO ASME SECTION XI APPENDIX G

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

Dickson, Terry L; Kirk, Mark

2010-01-01

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned normal reactor startup (heat-up) and shut-down (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are now recognized by the technical community as being conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, the nuclearmore » industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to provide a relaxation to the current regulations which will provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials, while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) have recently performed an independent review of the industry proposed methodology. The NRC / ORNL review consisted of performing probabilistic fracture mechanics (PFM) analyses for a matrix of cool-down and heat-up rates, permutated over various reactor geometries and characteristics, each at multiple levels of embrittlement, including 60 effective full power years (EFPY) and beyond, for various postulated flaw characterizations. The objective of this review is to quantify the risk of a reactor vessel experiencing non-ductile fracture, and possible subsequent failure, over a wide range of normal transient conditions, when the maximum allowable thermal-hydraulic boundary conditions, derived from both the current ASME code and the industry proposed methodology, are imposed on the inner surface of the reactor vessel. This paper discusses the results of the NRC/ORNL review of the industry proposal including the matrices of PFM analyses, results, insights, and conclusions derived from these analyses.« less

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

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

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

NEUTRONIC REACTOR SHIELDING

DOEpatents

Borst, L.B.

1961-07-11

A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

Reactor performance and microbial community dynamics during anaerobic co-digestion of municipal wastewater sludge with restaurant grease waste at steady state and overloading stages.

PubMed

Razaviarani, Vahid; Buchanan, Ian D

2014-11-01

Linkage between reactor performance and microbial community dynamics was investigated during mesophilic anaerobic co-digestion of restaurant grease waste (GTW) with municipal wastewater sludge (MWS) using 10L completely mixed reactors and a 20day SRT. Test reactors received a mixture of GTW and MWS while control reactors received only MWS. Addition of GTW to the test reactors enhanced the biogas production and methane yield by up to 65% and 120%, respectively. Pyrosequencing revealed that Methanosaeta and Methanomicrobium were the dominant acetoclastic and hydrogenotrophic methanogen genera, respectively, during stable reactor operation. The number of Methanosarcina and Methanomicrobium sequences increased and that of Methanosaeta declined when the proportion of GTW in the feed was increased to cause an overload condition. Under this overload condition, the pH, alkalinity and methane production decreased and VFA concentrations increased dramatically. Candidatus cloacamonas, affiliated within phylum Spirochaetes, were the dominant bacterial genus at all reactor loadings. Copyright © 2014 Elsevier Ltd. All rights reserved.

ETRCF, TRA654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE ...

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

ETR-CF, TRA-654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE HOOKS. ELECTRICAL EQUIPMENT IS PART OF PAST EXPERIMENT. DOOR AT LEFT EDGE OF VIEW LEADS TO REACTOR SERVICE BUILDING, TRA-635. INL NEGATIVE NO. HD24-1-2. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Comparison of Models for Spacer Grid Pressure Loss in Nuclear Fuel Bundles for One and Two-Phase Flows

NASA Astrophysics Data System (ADS)

Maskal, Alan B.

Spacer grids maintain the structural integrity of the fuel rods within fuel bundles of nuclear power plants. They can also improve flow characteristics within the nuclear reactor core. However, spacer grids add reactor coolant pressure losses, which require estimation and engineering into the design. Several mathematical models and computer codes were developed over decades to predict spacer grid pressure loss. Most models use generalized characteristics, measured by older, less precise equipment. The study of OECD/US-NRC BWR Full-Size Fine Mesh Bundle Tests (BFBT) provides updated and detailed experimental single and two-phase results, using technically advanced flow measurements for a wide range of boundary conditions. This thesis compares the predictions from the mathematical models to the BFBT experimental data by utilizing statistical formulae for accuracy and precision. This thesis also analyzes the effects of BFBT flow characteristics on spacer grids. No single model has been identified as valid for all flow conditions. However, some models' predictions perform better than others within a range of flow conditions, based on the accuracy and precision of the models' predictions. This study also demonstrates that pressure and flow quality have a significant effect on two-phase flow spacer grid models' biases.

Multi-MW Closed Cycle MHD Nuclear Space Power Via Nonequilibrium He/Xe Working Plasma

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Harada, Nobuhiro

2011-01-01

Prospects for a low specific mass multi-megawatt nuclear space power plant were examined assuming closed cycle coupling of a high-temperature fission reactor with magnetohydrodynamic (MHD) energy conversion and utilization of a nonequilibrium helium/xenon frozen inert plasma (FIP). Critical evaluation of performance attributes and specific mass characteristics was based on a comprehensive systems analysis assuming a reactor operating temperature of 1800 K for a range of subsystem mass properties. Total plant efficiency was expected to be 55.2% including plasma pre-ionization power, and the effects of compressor stage number, regenerator efficiency and radiation cooler temperature on plant efficiency were assessed. Optimal specific mass characteristics were found to be dependent on overall power plant scale with 3 kg/kWe being potentially achievable at a net electrical power output of 1-MWe. This figure drops to less than 2 kg/kWe when power output exceeds 3 MWe. Key technical issues include identification of effective methods for non-equilibrium pre-ionization and achievement of frozen inert plasma conditions within the MHD generator channel. A three-phase research and development strategy is proposed encompassing Phase-I Proof of Principle Experiments, a Phase-II Subscale Power Generation Experiment, and a Phase-III Closed-Loop Prototypical Laboratory Demonstration Test.

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

Cappiello, M.; Hobbins, R.; Penny, K.

As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As partmore » of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.« less

Nuclear code case development of printed-circuit heat exchangers with thermal and mechanical performance testing

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

Aakre, Shaun R.; Jentz, Ian W.; Anderson, Mark H.

The U.S. Department of Energy has agreed to fund a three-year integrated research project to close technical gaps involved with compact heat exchangers to be used in nuclear applications. This paper introduces the goals of the project, the research institutions, and industrial partners working in collaboration to develop a draft Boiler and Pressure Vessel Code Case for this technology. Heat exchanger testing, as well as non-destructive and destructive evaluation, will be performed by researchers across the country to understand the performance of compact heat exchangers. Testing will be performed using coolants and conditions proposed for Gen IV Reactor designs. Preliminarymore » observations of the mechanical failure mechanisms of the heat exchangers using destructive and non-destructive methods is presented. Unit-cell finite element models assembled to help predict the mechanical behavior of these high-temperature components are discussed as well. Performance testing methodology is laid out in this paper along with preliminary modeling results, an introduction to x-ray and neutron inspection techniques, and results from a recent pressurization test of a printed-circuit heat exchanger. The operational and quality assurance knowledge gained from these models and validation tests will be useful to developers of supercritical CO 2 systems, which commonly employ printed-circuit heat exchangers.« 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